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DOE OSTI2022

Characterization of Infrequent Samples from the Concentration, Storage, and Transfer Facility: H-Area Diversion Box 7 (HDB-7) Sump Sample: January 2022 Samples

Savannah River Remediation Engineering (SRR-E) requested that the Savannah River National Laboratory (SRNL) analyze the Concentration, Storage, and Transfer Facility (CSTF) samples from the following Tank Farm areas: the sump encasement, catch tank, drain cell, and waste tank annulus. In general, these CSTF samples will be analyzed on an infrequent basis and analyses will include detection for total beta/gamma, total alpha activities, density, free hydroxide, and pH measurements. This report presents characterization results for the H-Area diversion box 7 sump (HDB-7) 07 January 2022 and 18 January 2022 samples. The samples were clear and colorless and free from any solids. The results are measurements for total gamma, total alpha, total beta, density, free hydroxide, and pH.

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DOE OSTI2022

Characterization of Infrequent Samples from the Concentration, Storage, and Transfer Facility: H-Area Diversion Box 1 (HDB-1) Sump Sample: January 25, 2022 Sample

Savannah River Remediation Engineering (SRR-E) requested that the Savannah River National Laboratory (SRNL) analyze the Concentration, Storage, and Transfer Facility (CSTF) samples from the following Tank Farm areas: the sump encasement, catch tank, drain cell, and waste tank annulus. In general, these CSTF samples will be analyzed on an infrequent basis and analyses will include detection for total beta/gamma, total alpha activities, density, free hydroxide, and pH measurements. This report presents characterization results for the H-Area diversion box 1 sump (HDB-1) January 2022 sample. The sample was slightly brown with a small amount of suspended fine particulates visually estimated to be less than 1% by volume. The results are measurements for total gamma, total alpha, total beta, density, free hydroxide, and pH.

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DOE OSTI2022

Characterization of Infrequent Samples from the Concentration, Storage, and Transfer Facility: Leak Detection Box (LDB) Drain Cell Sample: February 09, 2022 Sample

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DOE OSTI2022

Characterization of Infrequent Samples from the Concentration, Storage, and Transfer Facility: Leak Detection Box (LDB) Drain Cell Sample: August 04, 2022 Sample

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DOE OSTI2023

Characterization of Infrequent Samples from The Concentration, Storage, And Transfer Facility: Leak Detection Box (LDB) Drain Cell Sample: February 26, 2023, Sample

Savannah River Mission Completion Engineering (SRMC-E) requested that the Savannah River National Laboratory (SRNL) analyze the Concentration, Storage, and Transfer Facility (CSTF) samples from the following Tank Farm areas: the sump encasement, catch tank, drain cell, and waste tank annulus. In general, these CSTF samples will be analyzed on an infrequent basis and analyses will include detection for total beta/gamma activities, total alpha activity, free hydroxide, and pH measurements. This report presents characterization results for the leak detection box (LDB) February 26, 2023, drain cell sample. The sample was clear and colorless with no visible particulates. The results are measurements for total gamma, total alpha, total beta, free hydroxide, pH, and density. These analyses were performed in triplicate. A summary of the average analytical results for the LDB sample includes the following. The directly measured pH for the LDB February 26, 2023 "as-received" drain cell sample range was 7.12-7.16, and the free hydroxide concentration was <0.02 M. The density of the "as-received" drain cell sample determined at 25 °C was 1.02±0.00 g/mL. The total alpha activity for the LDB February 26, 2023, sample is reported as a less than value (Upper Limit) because of possible spectral interferences. Thus, the total alpha activity averaged <3.52E+02 dpm/mL. This value is less than 4.83E+03 dpm/mL, which is the procedural limit for non-waste determination. The total beta activity in the LDB February 26, 2023, drain cell sample is above the instrument detection limits and average 9.76E+03±5.94E+02 dpm/mL. The total beta activity in the LDB February 26, 2023, drain cell sample is above he instrument detection limits and averaged 7.76E+03±5.94E+02 dpm/mL. The average measured cesium-137 activity (dominant beta emitter) in the LDB February 26, 2023, drain cell sample is 7.76E+03±2.35E+02 dpm/mL. The corresponding Ba-137m (dominant gamma emitter) activity, calculated as 94.6% of the Cs-137 values, is 7.34E+03±2.23E+02 dpm/mL. The total empirical activity of the beta and gamma emitting (represented by the sum of total beta and Ba-137m activities) averaged 1.71E+04±8.18E+02 dpm/mL. This value is less than 8.69E+05 dpm/mL, which is the procedural limit for non-waste determination.

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DOE OSTI2023

Characterization of Infrequent Samples from the Concentration, Storage, and Transfer Facility: Leak Detection Box (LDB) Drain Cell Sample: May 15, 2023, Sample

Savannah River Mission Completion Engineering (SRMC-E) requested that the Savannah River National Laboratory (SRNL) analyze the Concentration, Storage, and Transfer Facility (CSTF) samples from the following Tank Farm areas: the sump encasement, catch tank, drain cell, and waste tank annulus. In general, these CSTF samples will be analyzed on an infrequent basis and analyses will include detection for total beta/gamma activities, total alpha activity, free hydroxide, and pH measurement. SRMC-E corresponded with SRNL to obtain the final results through memo SRNL-L3120-2023-00013. This report presents characterization from results for the leak detection box (LDB) May 15, 2023, drain cell sample. The sample was clear and colorless with no visible particulates. The results are measurements for total gamma, total alpha, total beta, free hydroxide, pH, and density. These analyses were performed in triplicate. A summary of the average analytical results for the LDB sample includes the following. The measured pH for the LDB May 15, 2023 "as-received" drain cell sample range was 6.60-6.89, and the free hydroxide concentration as expected, based upon the pH, was <0.02 M. The average measured density of the "as-received" drain cell sample determined at 22 °C was 1.022±0.001 g/mL. the average measured total alpha activity for the LDB May 15, 2023, sample was 3.04E+02±1.13E+01. This value is less than 4.83E+03 dpm/mL, which is the procedural limit for non-waste determination. The total empirical activity of the beta and gamma emitting (represented by by the sum of total beta and B-137 m activities) averaged to 1.83+04±1.07E+03 dpm/mL. This value is less than 8.69E+05 dpm/mL, which is the procedural limit for non-waste determination.

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NASA NTRS2018

Preventing Premature Death in the M&S Lifecycle: Lessons Learned from Resurrection and Modernization of a Space System Contamination Model

Models and simulations (M&S) are often developed to meet specific needs and unique requirements for a particular situation. Once the M&S is implemented for a specific case and questions are answered, the M&S may go dormant until a similar need arises again at a later time, perhaps months to years later. Possible modification of the M&S may be required, and issues may arise if the M&S is not well documented, captured, or available. This can severely limit the useful life of the M&S and hinder future development or enhancements. This situation occurred with an M&S tool that had been developed to determine the impact to space system performance due to the presence of molecular contaminant films accumulating on key spacecraft surfaces. The challenges and issues encountered when resurrecting, executing, and modernizing the tool will be presented as a case study. To stay ahead of tomorrow’s challenges, resources to create M&S tools must be utilized efficiently. Lessons learned from this case study will aid M&S developers and users in planning for proper maintenance, transfer, and capture of key M&S tools and knowledge to avoid increased cost, increased development time, and wasted resources for projects relying on M&S.

Seasly, Elaine↗

DOE OSTI2019

Performance Metric for Cementitious Waste Form Inventory Release in the Integrated Disposal Facility

This report describes work supporting Phase 1 of the 2018 Department of Energy (DOE) Office of River Protection’s (ORP) Grand Challenge recipient proposal “Cementitious Immobilization of Treated LAW and Ancillary Benefits to Risk Reduction in Cross Site Transfer”. Phase 1 was to establish a performance metric for grout leachability against which laboratory leach data (as observed diffusivities) can be assed, followed by a comparison of historical test data against this metric. Historically, advancements in the cementitious (grout) waste form performance have been made over the last 40 years, with the majority of progress achieved within the past 8 years as part of the Washington River Protection Solutions, LLC testing program. This report presents a summary of that effort relevant to the Hanford Site.

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DOE OSTI2019

Review Guidelines for Biosphere Safety Assessment of Low-Level Waste Disposal Repository in Taiwan - 19606

Conducting reviews of radiological dose assessments is a crucial step toward the approval of long-term radiological safety requirements for low-level radioactive waste (LLW) disposal facilities. However, the concentration of radionuclides released from the near field into the biosphere and the radiation exposure levels to an individual can influence the performance acceptability of the LLW disposal facility and safety of the disposal environment. Therefore, all LLW must be managed and disposed of such that it minimizes risk to individuals and the environment. A dose of 0.25 mSv/year is the limit for individuals expected to receive the highest level of exposure to residual radioactivity. In Taiwan, a review for license applications has been proposed for LLW disposal facilities. These review guidelines are based on the Standard Review Plan (NUREG-1200) and are amended according to the environmental conditions in Taiwan. These guidelines can help reviewers to perform safety assessment reviews of applications for the construction and operation of LLW disposal facilities. The review guidelines for biosphere safety assessment include radiological dose assessment, description of waste characteristics, radionuclide transport characteristics, radiation dose under normal and unusual conditions, radionuclide release into the biosphere, and interpretation of evaluation results in various transfer mechanisms. These guidelines can also serve as references for regulatory departments conducting safety assessments of radiological doses in LLW disposal facilities in Taiwan. (authors)

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DOE OSTI2020

Preliminary Criticality and Radiation Shielding Analysis for the Storage and Transfer of MARVEL Reactor Spent Nuclear Fuel

This report documents the results of preliminary nuclear criticality and radiation shielding assessments during transfer and dry storage of irradiated Microreactor Applications Research Validation and Evaluation (MARVEL) reactor fuel at Idaho National Laboratory (INL). The assessments focus on transfer casks and storage canisters that are currently in use at INL, which may be compatible with the irradiated MARVEL reactor fuel. The criticality assessments were performed using the radiation transport code MCNP6 with 37 MARVEL reactor fuel elements in various configurations and scenarios. All transfer and storage configurations under dry conditions were below the assumed criticality safety limit of 0.93. Some storage and transfer configurations under wet conditions exceeded the criticality safety limit. This suggests that the appropriate administrative and engineering controls, in addition to reducing the number of MARVEL reactor fuel elements per transfer cask or storage canister, can be expected to ensure criticality safety under all scenarios. The radiation shielding assessments were performed by generating conservative neutron and photon source terms using the ORIGEN module in the SCALE suite of codes. These source spectra were used to estimate the dose equivalent rates using the radiation transport code MCNP6, both on contact and 1 m away from the fuel and transfer casks. The maximum estimated dose equivalent rate of 37 unshielded MARVEL reactor fuel elements on contact is approximately 42000 R/hr. The maximum estimated dose equivalent rates on contact to the ATR transfer cask, HFEF-5 transfer cask, and high load charger were approximately 233 mR/hr, 171 mR/hr, and 201 mR/hr, respectively. This suggests that with the appropriate administrative and engineering controls, all three transfer casks analyzed can be expected to provide sufficient radiation shielding to workers during transfer of irradiated MARVEL reactor fuel. These calculations are performed to support the planning and strategy for the MARVEL project and will demonstrate the technical viability of the different configurations discussed and help identify where engineered or administrative controls may be necessary. A complete criticality safety analysis and radiation shielding analysis, including validation and contingency and accident analysis must be completed by licensed and authorized personnel before any transfer or storage of MARVEL reactor nuclear fuel.

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NASA NTRS2013

Spacecraft Crew Cabin Condensation Control

A report discusses a new technique to prevent condensation on the cabin walls of manned spacecraft exposed to the cold environment of space, as such condensation could lead to free water in the cabin. This could facilitate the growth of mold and bacteria, and could lead to oxidation and weakening of the cabin wall. This condensation control technique employs a passive method that uses spacecraft waste heat as the primary wallheating mechanism. A network of heat pipes is bonded to the crew cabin pressure vessel, as well as the pipes to each other, in order to provide for efficient heat transfer to the cabin walls and from one heat pipe to another. When properly sized, the heat-pipe network can maintain the crew cabin walls at a nearly uniform temperature. It can also accept and distribute spacecraft waste heat to maintain the pressure vessel above dew point.

Carrillo, Laurie Y.↗

DOE OSTI2022

Analysis of DWPF Sludge Batch 6 (Macrobatch 7) Pour Stream Glass Samples

The Defense Waste Processing Facility (DWPF) began processing Sludge Batch 6 (SB6), also referred to as Macrobatch 7 (MB7), in June 2010. SB6 is a blend of the heel of Tank 40 from Sludge Batch 5 (SB5), H-Canyon Np transfers and SB6 that was transferred to Tank 40 from Tank 51. SB6 was processed using Frit 418. During processing of each sludge batch, the DWPF is required to take at least one glass sample to meet the objectives of the Glass Product Control Program and to complete the necessary Production Records so that the final glass product may be disposed of at a Federal Repository. Four pour stream glass samples and two Melter Feed Tank (MFT) slurry samples were collected while processing SB6. The samples were transferred to the Savannah River National Laboratory (SRNL) where they were analyzed.

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DOE OSTI2012

Critical Drivers for Safety Culture: Examining Department of Energy and U.S. Army Operational Experiences - 12382

Evaluating operational incidents can provide a window into the drivers most critical to establishing and maintaining a strong safety culture, thereby minimizing the potential project risk associated with safety incidents. By examining U.S. Department of Energy (DOE) versus U.S. Army drivers in terms of regulatory and contract requirements, programs implemented to address the requirements, and example case studies of operational events, a view of the elements most critical to making a positive influence on safety culture is presented. Four case studies are used in this evaluation; two from DOE and two from U.S. Army experiences. Although the standards guiding operations at these facilities are different, there are many similarities in the level of hazards, as well as the causes and the potential consequences of the events presented. Two of the incidents examined, one from a DOE operation and the other from a U.S. Army facility, resulted in workers receiving chemical burns. The remaining two incidents are similar in that significant conduct of operations failures occurred resulting in high-level radioactive waste (in the case of the DOE facility) or chemical agent (in the case of the Army facility) being transferred outside of engineering controls. A review of the investigation reports for all four events indicates the primary causes to be failures in work planning leading to ineffective hazard evaluation and control, lack of procedure adherence, and most importantly, lack of management oversight to effectively reinforce expectations for safe work planning and execution. DOE and Army safety programs are similar, and although there are some differences in contractual requirements, the expectations for safe performance are essentially the same. This analysis concludes that instilling a positive safety culture comes down to management leadership and engagement to (1) cultivate an environment that values a questioning attitude and (2) continually reinforce expectations for the appropriate level of rigor in work planning and procedure adherence. A review of the root causes and key contributing causes to the events indicate: - Three of the four root cause analyses cite lack of management engagement (oversight, involvement, ability to recognize issues, etc.) as a root cause to the events. - Two of the four root cause analyses cite work planning failures as a root cause to the events and all cause analyses reflect work planning failures as contributing factors to the events. - All events with the exception of the Tuba City plant shutdown indicate procedure noncompliance as a key contributor; in the case of Tuba City the procedure issues were primarily related to a lack of procedures, or a lack of sufficiently detailed procedures. - All events included discussion or suggestion of a lack of a questioning attitude, either on the part of management/supervision, work planners, or workers. This analysis suggests that the most critical drivers to safety culture are: - Management engagement, - Effective work planning and procedures, and - Procedure adherence with a questioning attitude to ensure procedural problems are identified and fixed. In high-hazard operational environments the importance of robust work planning processes and procedure adherence cannot be overstated. However, having the processes by themselves is not enough. Management must actively engage in expectation setting and ensure work planning that meets expectations for hazard analysis and control, develop a culture that encourages incident reporting and a questioning attitude, and routinely observe work performance to reinforce expectations for adherence to procedures/work control documents. In conclusion, the most critical driver to achieving a workforce culture that supports safe and effective project performance can be summarized as follows: 'Management engagement to continually reinforce expectations for work planning processes and procedure adherence in an environment that cultivates a questioning attitude'. (authors)

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DOE OSTI2012

Preservation of Records, Knowledge and Memory Across Generations. An emerging Multidisciplinary Work Area and an NEA Project - 12218

Disposal in engineered facilities built in stable, deep geological formations is the reference means for permanently isolating long-lived radioactive waste from the human biosphere. This management method is designed to be intrinsically safe and final, i.e. not dependent on human presence and intervention in order to fulfil its safety goal. There is however no intention to forgo, at any time, knowledge and awareness either of the repository or of the waste that it contains. The preservation of Records, Knowledge and Memory (RK and M) is seen as an integral part of radioactive waste management, supporting lengthy and complex socio-technical processes across pre-operational, operational and post-operational lifetimes. Long-term preservation of RK and M is an emerging multidisciplinary work area in which much learning is expected over the coming years. Novel methods are being sought that are least vulnerable to both natural degradation and to changes in socio-economic conditions. Progress has been made in individual countries, but there is a need to internationalise the thinking, compare approaches, investigate potential solutions and share decisions. This is the task of the NEA RK and M project. A major outcome of the project will be a 'menu-driven document' that will allow people to identify the main elements of a strategic action plan for RK and M preservation. In sum, the preservation of RK and M is a unprecedented task in which technical, scientific and social information is interwoven and needs to be developed and preserved across generations and across specialist boundaries. Important studies have been undertaken in the past decades to explore a variety of approaches to preserving RK and M across different timescales, including archives and markers. The work of the past in this area is useful, but innovative thinking is also needed. Seen from today's perspective, very little work is available on for example the contextualization of data for later use; on the systematic identification of mechanisms for RK and M transfer; on implementing a culture of RK and M-keeping in organisations; and on creating cultural links between the waste disposals and the siting communities. Moreover, international cooperation is recognised as being crucial in providing shared means and meanings for memory transmission over longer timescales. International cooperation has also been identified as a catalyst to ensure that a wide range of approaches and experiences is considered, thus potentially reducing uncertainty related to variations in approach. Overall, multiple approaches, requiring active and less active care, need to be considered from the start of the radioactive waste management programme and refined in the course of time. The RK and M project members want to further investigate and deliver support to fulfil this task. (authors)

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DOE OSTI2012

Analysis of DWPF Sludge Batch 7a (Macrobatch 8) Pour Stream Samples

The Defense Waste Processing Facility (DWPF) began processing Sludge Batch 7a (SB7a), also referred to as Macrobatch 8 (MB8), in June 2011. SB7a is a blend of the heel of Tank 40 from Sludge Batch 6 (SB6) and the SB7a material that was transferred to Tank 40 from Tank 51. SB7a was processed using Frit 418. During processing of each sludge batch, the DWPF is required to take at least one glass sample to meet the objectives of the Glass Product Control Program (GPCP), which is governed by the DWPF Waste Compliance Plan, and to complete the necessary Production Records so that the final glass product may be disposed of at a Federal Repository. Three pour stream glass samples and two Melter Feed Tank (MFT) slurry samples were collected while processing SB7a. These additional samples were taken during SB7a to understand the impact of antifoam and the melter bubblers on glass redox chemistry. The samples were transferred to the Savannah River National Laboratory (SRNL) where they were analyzed. The following conclusions were drawn from the analytical results provided in this report: (1) The sum of oxides for the official SB7a pour stream glass is within the Product Composition Control System (PCCS) limits (95-105 wt%). (2) The average calculated Waste Dilution Factor (WDF) for SB7a is 2.3. In general, the measured radionuclide content of the official SB7a pour stream glass is in good agreement with the calculated values from the Tank 40 dried sludge results from the SB7a Waste Acceptance Program Specification (WAPS) sample. (3) As in previous pour stream samples, ruthenium and rhodium inclusions were detected by Scanning Electron Microscopy-Electron Dispersive Spectroscopy (SEM-EDS) in the official SB7a pour stream sample. (4) The Product Consistency Test (PCT) results indicate that the official SB7a pour stream glass meets the waste acceptance criteria for durability with a normalized boron release of 0.64 g/L, which is an order of magnitude less than the Environmental Assessment (EA) glass. (5) The measured density of the SB7a pour stream glass was 2.7 g/cm 3 . (6) The Fe 2+ ΣFe ratios of the SB7a pour stream samples were in the range of 0.04-0.13, while the MFT sample glasses prepared by SRNL were in the range of 0.02-0.04.

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DOE OSTI2019

Modification of an Agitated Tube Reactor for use in the Reprocessing of Spent Nuclear Fuel

Aims: Convert an Agitated Tube Reactor (ATR) into a counter-current solvent extraction device. Assess effect of liquid fill level on liquid dispersion. The agitated tube reactor (ATR) is an array of connected tubes. The ATR laterally shakes in a sinusoidal path, moving the agitator bar to create mixing. Originally designed by AM Technology as a cocurrent reactor, the ATR has many features that are potentially useful in the nuclear industry: Tubes are connected by non-moving seals; Reduction in seal damage and reduced risk of expensive repairs in radioactive environments; Minimal axial dispersion within a tube; Plug flow; Narrower residence time distributions that traditional technology; Multiple tubes allow for multiple reactions in series; Converting to a counter-current design allows the ATR to be used for solvent extraction; Reprocessing or decommissioning. A 180 ml PERSPEX ATR tube with no liquid flow through was constructed and filled with a 1:1 solvent:aqueous ratio of kerosene and water. The tube was filled such that the total liquid filled 25 %, 50 %, 75 % or 100 % of the tube. A Photron Fastcam SA5, positioned at the side and the end of the tube was set to a frame rate of 5400 fps with a resolution of 1024 x 1024 and 6000 fps with a resolution of 1024 x 896 respectively. Droplets were recorded after 2 minutes of shaking to ensure good mixing. Droplet sizes are used to determine optimum fill level in the ATR. The shaking frequency follows a sinusoidal path and was set to 5 Hz. The images shown in the results section were taken at the peak of the sinusoidal motion. To convert to a counter-current design the two liquids need to be separated at the end of each tube. Liquids have been allowed to separate under gravity separation. Inclusion of baffles to reduce remixing by sloshing. Wall with height corresponding to 25 % fill volume allows kerosene to overflow into separate outlet. Increase in liquid fill level decreases the dispersion of the two liquid phases. Higher dispersion of droplets decreases surface area and therefore increases mass transfer in solvent extraction. 25 % should provide the greatest mass transfer though more work needs to completed to confirm this.

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DOE OSTI2020

Development of a Testbed for Pipeline Flushing - 20425

Pipelines which carry high-level waste within the DOE complex should be properly flushed in the cases where stationary or moving beds of solid sediment occur, or lines are prone to hydrogen gas buildup. Present guidelines establish a minimum for the amount of water (flush volume) and flush velocity values used for post-transfer flushing operations to achieve satisfactory cleaning of pipelines. However, further studies are needed to find optimal operating velocity modes/values in order to minimizing the flush volume and consequent downstream waste. These efforts are significantly helpful to DOE waste remediation sites by preserving tank storage, preventing additional waste processing, and minimizing dilution and changes in waste chemistry. An experimental test loop was recently developed for study of non-Newtonian slurry flushing at the Florida International University (FIU). This loop was designed to create sediment beds of various materials and bed heights and investigate parameters that effect the efficiency of flushing operations. The objective is to find flush velocity values/modes which lead to satisfactory cleaning of transport lines with a minimum amount of water usage. A 165 ft experimental test loop made of 3-inch carbon steel pipes was constructed and used in initial testing campaigns. This pipe loop was equipped with real-time monitoring instruments to record variation of pressure, density, mass flow rate, and temperature during testing. Post-flush in-situ no-flow evaluations were performed using ultrasonic, endoscopy, and visual inspection. Additional post-flush evaluations were conducted using filtration and density monitoring of residuals within a designated circulations loop. These provisions enable comparison of pipeline cleanness before the start and after termination of flushing operations for efficiency evaluation. This paper presents efforts associated with flushing of kaolin-water mixtures at different concentrations and with various flush volume and flush velocity values/modes. In initial testing, efforts focused on creating repeatable sediment beds inside the pipeline in both fully-flooded (no pre-flush drainage) and gravity-drained (pre-flush drained) systems. Bed characterization (i.e., measurement of sediment height and solids concentration) was conducted to ensure about consistency of initial conditions between tests. Preliminary results for flushing study of 10% vol. kaolin-water mixtures starting from a fully-flooded initial condition showed that flushing with 1-line volume (65 gallons) could remove up to 72 percent of solids from the pipeline. (authors)

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DOE OSTI2025

FY24 Task 4: Studies of Phosphate and Fluoride Solubility for Dissolution of High Phosphate Tank Waste

Knowledge gaps have been identified in phosphate solubility in almost every single- and multi-component system, and particularly for aluminate in phosphate/hydroxide, where uncertainties in predicted values of aluminum solubility are greater than 50%. Solubility data of multicomponent, aqueous electrolytes containing sodium hydroxide, sodium fluoride, sodium phosphate, sodium nitrite, sodium nitrate, and dissolved gibbsite are also sparse. For example, for solutions of (i) sodium nitrate and sodium phosphate, data are only available at 30 and 50 °C, and for (ii) dissolved gibbsite in sodium hydroxide and sodium phosphate, only two data points are available at 20 and 40 °C. There is no data available on mixtures of sodium hydroxide, sodium phosphate, sodium fluoride, and dissolved gibbsite. These knowledge gaps were identified in a technical review of waste solubility data and the impact of dilution on solution stabilities and will be addressed in this work to predict aluminum hydroxide and sodium phosphate solubility in multicomponent electrolytes upon dilution, and upon variation of temperature. Results will provide the technical basis to develop accurate models for (i) gibbsite solubility and mass transfer of aluminum between solid and liquid forms following the sluicing of sludge and saltcake with water; and (ii) further blending of these suspensions with bismuth from bismuth phosphate waste, and zirconium and uranium left from the fuel decladding. This work will be essential to developing a disposition path for retrieval solutions from bismuth phosphate wastes. This effort would also support sludge washing to further reduce phosphate concentration if that process were to be added back to the flowsheet in the future.

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