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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Cesium Batch Contact Equilibrium Testing of Crystalline Silicotitanate (CST) Sieve Cuts in SRS Average Simulant and Examination of CST Samples Aged in Caustic and Simulant in Support of Tank Closure Cesium Removal 1A (TCCR-1A)

Batch contact testing to determine cesium equilibrium loading on Crystalline Silicotitanate ion exchange media in Savannah River Site Average Waste Simulant at 25 °C indicated that smaller diameter particles isolated by sieving pretreated CST media may load slightly higher (<10%) amounts of cesium, though the differences are within analytical uncertainty. In addition, ion exchange media sub-samples stored in 2-4 M NaOH and caustic simulant solutions for ~2.5 years were examined by optical microscopy and the 3 M NaOH sample was also analyzed to determine whether changes in the particle size distribution occurred during storage. No visual indications of particle attrition or agglomeration were observed for any sample. Particle size analysis indicated that a slight decrease occurred in the average particle diameter following contact with 3 M NaOH (541 µm average diameter versus 566 µm for the pretreated CST prior to contact). A small increase (from 0 to <0.5 wt. %) in the number of particles ranging from 271 and 322 µm was also observed for the CST sample contacted with 3 M NaOH relative to a sample of the original pretreated material. However, this small change could be due to sub-sampling differences or analytical uncertainty. It does not appear that small particles are formed to a significant degree during CST caustic contact or that small particles which do form (presumably from attrition of larger particles during pretreatment) load significantly more cesium than the bulk material. Minimal other negative consequences were observed associated with CST extended caustic or simulant contact, except for the tendency for more concentrated salt solutions to form some salt crystals which deposit on the media over time.

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DOE OSTI2025

An Application of Molecular Recognition for the Efficient Removal of Cesium from Hanford Nuclear Waste by Modular Solvent Extraction

In this work, experimental results leading to flowsheet design are presented showing how a calixarene-crown ether based solvent-extraction process can meet the challenge of cesium removal from nuclear tank wastes stored at the US Department of Energy Hanford site. Cleanup of legacy Cold War nuclear waste stored in underground tanks represents one of the greatest environmental challenges facing the US Department of Energy in terms of risk, cost, and effectiveness of applicable science and technology. Planning for the cleanup at the Hanford Site calls for the removal of the radioactive fission product 137Cs from its alkaline salt waste, including the use of modular processes that can be deployed near the tank farms. To meet the resulting need for extremely high selectivity, the Next-Generation Caustic-Side Solvent Extraction (NG-CSSX) process employing a calix[4]arene-crown ether in modified kerosene has been adapted to remove sub-millimolar cesium in competition with molar sodium and potassium in a high-nitrate alkaline matrix. Potassium loading in the solvent was determined in extraction, scrubbing, and stripping, leading to an empirical model closely approximating cesium distribution ratios for a variety of Hanford waste types. Process chemistry has been developed based on this molecular-recognition approach, focusing on the competitive effect of potassium loading and the mitigating process modifications needed, including extending the scrub section. The result is a modular flowsheet design that can achieve cesium decontamination factors well in excess of 15,000 even for the worst-case Hanford waste.

Williams, Neil [ORNL] (ORCID:000000023159226X)↗
DOE OSTI2021

Cesium Exchange onto Crystalline Silicotitanate from Blended Hanford Tank Wastes

The Tank Side Cesium Removal (TSCR) system was developed to filter and remove cesium (Cs and 137 Cs) from Hanford tank waste supernate in preparation for vitrification. The Cs removal will be conducted with crystalline silicotitanate (CST) ion exchange media. Under the planned waste-processing strategy, the tank waste supernate will be queued for TSCR processing in tank 241-AP-107 (AP-107). Once AP-107 tank waste volume is sufficiently depleted, the waste supernate from tank 241-AP-105 (AP-105, the holding tank before transfer to AP-107) will be transferred to tank AP-107. Supernate from another tank will be transferred to the holding tank, AP-105, for eventual transfer to tank AP-107. These supernate streams will undergo blending in tanks AP-107 and AP-105; the volume blend ratios will be driven by how much the tank waste supernate volumes are depleted before the next tank waste is added. The consequence of tank waste blending on Cs uptake by CST was of interest and was tested via batch contacts; results are reported herein.

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DOE OSTI2022

Ion Exchange Processing of AP-105 Hanford Tank Waste through Crystalline Silicotitanate in a Staged 2- then 3-Column System (Rev.1)

The Tank Side Cesium Removal (TSCR) system, under development by Washington River Protection Solutions LLC (WRPS), will send initial low-activity Hanford waste tank supernate feeds to the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Facility. In addition to entrained solids removal from the supernate, the primary goal of TSCR is to remove cesium-137 ( 137 Cs) by ion exchange, allowing contact handling of the liquid effluent product at the WTP. Crystalline silicotitanate (CST) ion exchange media, manufactured by Honeywell UOP, LLC (product IONSIVTM R9140-B), was selected as the ion exchange media at TSCR. CST is a non-elutable inorganic material that has demonstrated robust chemical, physical, and radiation tolerance while maintaining functionality. However, exchange kinetics of Cs onto CST is slow, resulting in low utilization of the CST Cs load capacity before unacceptable Cs breakthrough. Two process flow designs have been tested, as follows. 1. Lead-lag column processing: The lead column was removed after the lag column effluent reached the waste acceptance criteria (WAC) limit, the lag column was moved into the lead position, and a new lag column was installed. This format used ~52% Cs load capacity on the lead column. 2. Lead-lag-polish column processing: The processing was stopped when the polish column effluent reached the WAC limit. This format resulted in 81% Cs load capacity on the lead column. Testing with diluted feed from Hanford tank AP-105 (AP-105DF) incorporated a nuanced change to the lead-lag-polish column system where the polish column was inserted when the lag column effluent reached WAC limit. A 10.9-L volume of AP-105DF (diluted to 5.6 M Na) was processed through the Direct Feed Test Platform system, established at Pacific Northwest National Laboratory to support small-scale waste qualification efforts. The columns consisted of 10-mL CST beds (CST Lot 2002009604, sieved to screen out >30-mesh particles) placed in 1.5-cm-inner-diameter columns. Feed was processed at 1.83 bed volumes (BV) per hour; the flowrate, in terms of contact time with the CST bed, matched the expected flowrate at TSCR. The <30-mesh CST sieve cut was expected to provide appropriate performance scaling to a full-height column. The installation of the polish column later in processing (after processing 523 BVs) did not appear to fundamentally change the utilization of the lead column for Cs exchange nor did it extend the total feed processing volume when compared to the previous test with AP-107 feed. Table ES.1 and Figure ES.1 summarize the measured AP-105DF Cs load performance.

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DOE OSTI2020

Preliminary Determination of the Impact of Alkaline Earth Metals on Crystalline Silicotitanate

Savannah River Remediation (SRR) recently deployed the Tank Closure Cesium Removal (TCCR) system which utilizes an ion exchange process to remove 137 Cs from tank waste supernate. In TCCR, salt solution is filtered and then passed through ion exchange columns containing crystalline silicotitanate (CST) media, commercially known as UOP IONSIV™ R9120-B, which is highly selective for removing cesium. The current process focuses on dissolving salt in Savannah River Site (SRS) Tank 10H, followed by processing in the TCCR skid located near the tank. Prior to processing each batch of waste generated in Tank 10H, the projected maximum cesium loading on the CST is determined using an in-tank batch contact equilibrium test. This involves placing a small amount (~0.1 g) of CST in a “teabag” device and submerging it in the tank waste for a period of 10 days. Results from those earlier teabag tests had indicated much lower cesium loading than predicted by the equilibrium sorption model (ZAM). There are numerous possible explanations for this result; however, based on the observed loadings of Ca (and other metals) on the CST from the teabag tests, it was hypothesized that competition from these other ions may be limiting the cesium loadings.

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DOE OSTI2021

Temperature Effect of Cesium Exchange onto Crystalline Silicotitanate in AP-107 and AP-105 Hanford Tank Wastes and Two Simulants

Washington River Protection Solutions, LLC (WRPS) is charged with the development of the Tank Side Cesium Removal (TSCR) system to process Hanford tank waste supernates in preparation for vitrification. In addition to a filtration step, TSCR will remove cesium (Cs) using ion exchange columns filled with crystalline silicotitanate (CST) ion exchange media. CST is produced by Honeywell UOP, LLC. The documented safety analysis (DSA) developed for the TSCR system limits a single column loading to 141,600 Ci 137 Cs. Given a 137 Cs isotopic mass fraction of 20% and the planned CST bed size of 596 L (157.5 gal) in a TSCR column, this equates to 0.10 mmole Cs per g CST (Cs distribution coefficient, K d , 1400 mL/g). Factors that influence Cs uptake by CST include (but are not limited to) (1) CST production (lot-to-lot variations), (2) contact temperature, (3) contact duration, (4) competitors in the tank waste feed, (5) anionic composition of the tank waste feed, and (6) the 137 Cs isotopic mass fraction (differs slightly among tank wastes and decreases with time).

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DOE OSTI2020

Assessment of the Corrosion Risks to a Double Shell Tank During the Selective Removal of a Cesium-rich Supernate Layer - 20227

The Direct Feed Low Activity Waste (DF-LAW) activity will provide the initial feed to the Waste Treatment and Immobilization Plant at the Hanford Tank Farm facility. Feed preparation for DF-LAW will be carried out in multiple double shell waste tanks (DST) utilizing the Tank Side Cesium Removal (TSCR) process. Tank AP-106 will be an integral part of the TSCR process and hence key to the DFLAW activity. Prior to being utilized for this purpose, the Cs-137 concentration levels in the AP-106 supernate must be reduced significantly. The AP-106 re-purposing process manages normal waste tank addition procedures such that waste acceptance criteria objectives for Cs-137 removal are achieved. As this process was being developed, several concerns regarding tank integrity during and after repurposing of Tank AP-106, and the associated tanks (i.e., AP-102, AN-101, and AP-108), were addressed. The potential corrosion degradation mechanisms identified based on process knowledge included: general and pitting corrosion, caustic stress corrosion cracking, concentration cell corrosion, and microbiologically induced corrosion (MIC). A systematic approach was undertaken to assess the risk of corrosion during the AP-106 re-purposing process. The assessment considered the initial and present condition of the tank, the service history of the tank, the environmental conditions created in each tank during each stage of the process, process timing, and corrosion testing and modeling that approximated tank conditions. Based on this information, recommendations were provided to minimize the exposure of the tank to corrosive conditions. These recommendations included time limits for each stage of the process, staging of corrosion inhibitor solutions should process delays occur, and waste sampling with potential chemistry adjustments at defined steps in the process. Additionally, recommendations for post-process supernate chemistry adjustments and tank inspections were put forth. The greatest risk for this process occurs during the manipulation of large quantities of Columbia River water (i.e., greater than 450,000 gallons) within AP-106, the associated tanks and the supporting infrastructure (i.e., transfer line piping). This risk becomes more pronounced if there is a process upset and the tank walls remain exposed to a corrosive condition longer than intended. As long as the river water is blended into the supernate in a timely fashion, and the waste is adjusted appropriately with corrosion inhibitors, corrosion may be minimized and the final condition of the supernate in AP-106 and the associated tanks is expected to be in compliance with corrosion chemistry requirements. Contingency plans will be made to minimize the impact of a process upset. (authors)

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DOE OSTI2020

TCCR Operational Summary and Optimization for Tank 9 Processing - 20314

Savannah River Remediation (SRR) manages and operates the liquid waste facilities at Savannah River Site (SRS) for the Department of Energy (DOE). Stored liquid waste is a complex mixture of insoluble solids (sludge) and soluble salts in an alkaline solution. SRR has deployed the Tank Closure Cesium Removal (TCCR) system, a tank-side ion exchange process, to remove radioactive cesium from salt waste and enable onsite disposal of the resulting decontaminated salt solution as low-level waste at the Saltstone facilities. The TCCR system consists of two prefilters, four ion exchange (IX) columns, one resin trap, and a ventilation system. The IX process uses a form of inorganic crystalline silicotitanate (CST), which has a high affinity for cesium and other alkali metals, strontium, and actinides. This process is currently deployed utilizing salt feed from Tank 10, with future plans to dissolve solid salt in Tank 9 and transfer the salt solution to Tank 10 for processing through TCCR. The feed for TCCR must be created from salt-cake in Tank 10 through a dissolution process. Once enough salt has been dissolved, a qualification process is entered. This process characterizes the feed and ensures the cesium loading on the columns will not cause boiling of waste within the columns during or after processing. Once the batch has been qualified, salt waste is fed to the TCCR system through a transfer pump in the center of the tank. The waste is filtered through a set of two shielded, dead-end prefilters that prevent solids buildup in the columns. The filtered salt solution then travels to the shielded IX columns, which can be operated individually or in series, where the cesium is sorbed on the CST media. The decontaminated salt solution (DSS) then travels through a resin trap and out of the module to Tank 11. TCCR has successfully processed approximately 795,000 L of Tank 10H radioactive salt waste over two batches to date. There has not yet been a system induced shutdown. The prefilters performed as expected with only minor degradation in recovery of differential pressure after a backflush sequence. The time between backflushes decreased as each batch reached the end of processing. The hydraulics in the IXCs mostly performed as expected at all flow rates, except for one IXC that will be further investigated during Batch 3 processing. The TCCR system has shown some opportunities for more efficient processing during the length of the demonstration so far. For future processing of material from Tank 9H through Tank 10H and the TCCR unit, TCCR 1A will implement changes to the prefilters and the IXCs. The prefilters will have an increased surface area and a new filter media in an effort to increase time between filter swaps and improve backwashing cleaning capability. The IXCs will have a reduced diameter to allow for increased heat transfer out of the column and increased loading of Cs-137. Additionally, a new form of CST with an increased kinetic performance is being investigated for use during TCCR 1A operation. (authors)

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DOE OSTI2020

Characterization and CST Batch Contact Equilibrium Testing of Modified Tank 9H Process Supernate Samples in Support of TCCR

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with crystalline silicotitanate (CST) media to process radioactive waste solutions for the removal of ¹³⁷Cs. TCCR currently focuses on dissolving Savannah River Site (SRS) radioactive tank waste (primarily sodium saltcake solids) within Tank 10H followed by at-tank ion exchange column treatment. Two supernate batches from Tank 10H have been processed through the TCCR unit and processing of a third batch is expected soon. After processing of this supernate batch, plans are to replace the CST columns and process dissolved salt solution from Tank 9H through Tank 10H and then through the new CST columns installed in the TCCR unit. The new columns are expected to contain either a media similar to an archived CST batch (IE-911) or the R9120-B CST media used in the current TCCR columns (the two materials are fundamentally the same; just different specifications, product names, and preconditioning steps). Samples of Tank 9H dissolved saltcake were received at the Savannah River National Laboratory (SRNL) and characterized. The Tank 9H supernate contained a high sodium concentration (~9.6 M Na⁺) and will require dilution to near 6 M [Na⁺] prior to processing through the TCCR unit. The cesium concentration in Tank 9H is currently significantly higher than was observed with Tank 10H. Three dilutions of the Tank 9H supernate were conducted to mimic possible dilutions that could be conducted in the tank farm prior to TCCR processing using inhibited water, sodium hydroxide, and sodium nitrate solutions. Dilution #1 was prepared by diluting the Tank 9H supernate by a factor of 1.6 with inhibited water, Dilution #2 was prepared by diluting the Tank 9H supernate by a factor of 3.7 with a mixed sodium hydroxide/sodium nitrate diluent, and Dilution #3 was prepared by also diluting by a factor of 3.7 but with sodium hydroxide only. All three dilutions have similar Na⁺ concentrations (~6 M), but Dilutions #2 and #3 contain significantly less cesium. Major and key minor components of the diluted Tank 9H supernate samples are provided in Table ES-1. Batch contact equilibrium tests were conducted with the Tank 9H dilutions and the two different CST media batches being considered for use in the new TCCR columns. Results are summarized in Table ES-2 and compared to ZAM model predictions. The highest cesium distribution coefficient and percent removal were observed with Dilution #3. IE-911 CST (an archived CST media batch) was more effective at removing cesium than the more recently prepared R9120-B CST media, though both media samples removed >88% of the cesium and the differences may not be statistically different considering the overall uncertainty. Based on the results, maximum cesium loadings were calculated for each CST media type and Tank 9H dilution. In general, maximum cesium loadings from dilutions of this supernate batch are quite high (approaching 0.1 mmol total Cs⁺/g CST for Dilution #1). The highest calculated maximum ¹³⁷Cs loading for the Tank 9H dilutions using the ZAM model with input of the tank compositions and batch contact results was 207 Ci/kg CST (Dilution #1 with IE-911 CST). In all cases, higher maximum cesium loading values were predicted for IE-911 CST versus R9120-B, although the differences varied considerably between the dilutions. The maximum loading value for IE-911 CST with Tank 9H Dilution #1 was only 7% higher than the maximum loading for R9120-B. The maximum loading value for IE-911 CST with Dilution #3 was 24% higher than the maximum loading for R9120-B. Dilution #2 was intermediate between these values. Note that these maximum loading values are the theoretical calculated values, and actual operating conditions will cause differences. A CST binder dilution (correction) factor near 0.7 (relative to pure powder CST) was required for each batch contact test with IE-911 engineered CST using the three Tank 9H dilutions. Correction factors calculated for R9120-B CST ranged from 0.56 to 0.66 for the three dilutions. Following the batch contact tests, the CST samples were isolated from the Tank 9H solution, washed, dried, and digested in acid following established procedures. The analysis results are provided in Table ES-3. Total cesium loading values determined by CST digestion were similar to the calculated loading values based on solution analysis (Table ES-2) for all samples. In addition, the CST was observed to load calcium (R9120-B sample only), iron, strontium, lead, uranium, and plutonium after contact with the waste supernate, as has been observed previously.

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DOE OSTI2021

Characterization of Infrequent Samples from the Concentration, Storage, and Transfer Facility: F-Area Pump Pit 3 (FPP-3) Sump Sample: December 2021 Sample

In December 2021, SRR-E sent an ~185 mL sample identified as FPP-3 from a F-Area pump pit sump to SRNL for analysis. The sample was pale yellow in color and contained a small amount of suspended solids visually estimated to be less than 1% by volume. SRNL analysis indicated that the sample contained 4.4 E+05 dpm/mL Cs-137, 6.2 E+05 dpm/mL total beta activity, 9.3 E+02 dpm/mL beta activity following cesium removal, and below detectable levels of alpha activity following cesium removal. In addition, the sample contained 0.015 M free OH- and 947 μg total organic carbon/mL.

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DOE OSTI2023

Options for the Removal of Cesium and Strontium from Acidic Uranyl Sulfate Solutions from Molybdenum-99 Production: Part II The Effect of Sodium Carbonate Adjustment to pH 9 on Strontium Removal

A small laboratory-scale column process aimed at removing fission products from prototypical acidic sulfate solution wastes from a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (0.4 wt % 235 U), plutonium ( 239 Pu), neptunium ( 237 Np), non-radioactive surrogates of common fission products (including Cs and Sr) and two promising absorbents – AMP-PAN and CST R9120-B. The goal of this work was to demonstrate the removal and concentration of high specific activity isotopes which are expected to dominate the classification of the low-level waste from the production of 99 Mo. Both AMP-PAN and CST R9120-B are attractive absorbents for the removal of 137 Cs from an acidic sulfate waste stream. They do not provide a solution for the removal of 90 Sr from the acidic solutions. CST was shown to be useful at removing both Cs and Sr after adjustment of the acidic sulfate solution with sodium carbonate to pH 9. Limited removal of transuranic species and almost no removal of uranium was observed for CST R9120B absorbent. This could be important due to the relatively large amount of uranium expected to be present in such waste streams. The capacity of these absorbents for 137 Cs will likely be dependent on the Cs concentration in the solutions processed but loading in excess of 0.7 mg Cs per gram of CST absorbent while loading 0.14 mg Sr per gram of CST absorbent appears to be achievable from pH 9 carbonate solution.

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DOE OSTI2021

Characterization of Infrequent Samples from the Concentration, Storage, and Transfer Facility: H-Area Diversion Box 7 (HDB-7) Sump Sample: December 2021 Sample

In December 2021, SRR-E sent an ~250 mL sample identified as HDB-7 from an H-Area diversion box sump to SRNL for analysis. The sample was clear and colorless and free from any solids. SRNL analysis indicated that the sample contained 2.97E+05 dpm/mL Cs-137, 3.74E+05 dpm/mL total beta activity, 1.04E+03 dpm/mL beta activity following cesium removal, and below detectable levels of alpha activity and alpha activity following cesium removal. In addition, the pH of the sample was 7.32, free hydroxide concentration was 2.09E-07 M, and the density was 0.991 g/mL.

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DOE OSTI2022

Options for the Removal of Cesium and Strontium from Acidic Uranyl Sulfate Solutions from Molybdenum-99 Production

A small laboratory-scale column process aimed at removing fission products from prototypical acidic sulfate solution wastes from a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (0.4 wt % 235 U), plutonium ( 239 Pu), neptunium ( 237 Np), non-radioactive surrogates of common fission products (including Cs and Sr) and two promising absorbents AMP-PAN and CST R91290-B. The goal of this work was to demonstrate the removal and concentration of high specific activity isotopes which are expected to dominate the classification of the low-level waste from the production of 99 Mo. Both AMP-PAN and CST R9120-B are attractive absorbents for the removal of 137 Cs from an acidic sulfate waste stream. They do not provide a solution for the removal of 90 Sr from the acidic solutions, but CST should be investigated at a higher pH. Limited removal of transuranic species and almost no removal of uranium was observed in this testing. This could be important due to the relatively large amount of uranium expected to be present in such waste streams. The capacity of these absorbents for 137 Cs will likely be dependent on the Cs concentration in the solutions processed but loading in the range of 0.2 to 5 mg Cs per gram of absorbent appears to be achievable.

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DOE OSTI2021

High Solids Performance Testing in a Scaled TSCR System

The Tank Side Cesium Removal (TSCR) project is a technology demonstration that will pretreat Hanford tank waste supernatant in support of the Direct Feed Low-Activity Waste (DFLAW) mission. The TSCR system employs two key separation technologies: dead-end filtration (DEF) and ion exchange (IX) using crystalline silicotitanate (CST) media. DEF will be used to remove undissolved solids from tank waste to protect the functionality of the IX columns and the IX system will remove Cs-137 from tank waste. The separation technologies (DEF and IX) used in TSCR are technically mature and have also been successfully deployed at the Savannah River site in a similar facility known as the Tank Closure Cesium Removal (TCCR) system. While testing with simulants and real waste has been successfully performed under conditions expected during the initial operation of TSCR, test data is absent for assessing off normal high solids loading that may be in the TSCR waste feed. Normal TSCR treatment operations are expected to handle wastes with solids content on the order of 200 ppm, and off normal solids loading could be much larger than the nominal level. The testing program described in this report was conducted to understand the consequence of operating the TSCR system at elevated solids loadings up to the high-solids limit of 15,000 ppm [i.e., 1.5-wt%] identified in the TSCR design basis. Although the system is not required to make throughput above the nominal solids loading, the testing was intended to provide important information related to potential off normal operations. At off normal levels near the high-solids limit, there are potential implications for TSCR performance in the areas of throughput, DEF pressure drop, filter backflush frequency, and IX column pressure drop. In addition, intrusion of solids into the IX column was postulated to impact the Cs-137 loading behavior by promoting channeling or flow maldistribution in the column; since the magnitude of the postulated effect was unknown, assessing it was also of interest. The testing was performed using representative waste simulants and a prototypic, integrated TSCR system designed and assembled specifically to conduct the high solids performance assessment. Overall, the scaled TSCR testing demonstrated that full-scale unit operations can succeed in fulfilling their processing objectives in the presence of solids up to 3,000 ppm, but there are potential performance challenges to filtration operations at solids loadings as low as ~500 ppm. The severity of the challenge is likely to be dependent on the type and size distribution of solids, of which the current testing only examined a single type and size distribution. To provide some flexibility for future full-scale operations, the results of the testing suggest two possible risk reduction strategies that can be implemented without any changes in TSCR design or configuration. One option would be to enact an administrative limit on the solids loading to protect TSCR from feeds that are likely to require a high DEF swap frequency. Another option is to permit operation of the DEFs at differential pressures greater than 2 psid before swapping filters. The selection of a higher differential pressure target is not anticipated to adversely impact DEF backflushing efficacy and would reduce both swap frequency and the amount of waste sent to AP-108.

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