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At least 235 records · Page 13

Investigation of Mercury Compound Partitioning Through the Solvent Extraction System

Savannah River National Laboratory (SRNL) has been requested by Savannah River Mission Completion (SRMC)/ Salt Waste Processing Facility (SWPF) personnel to investigate the potential routes for titanium and mercury accumulations within the SWPF Flowsheet. This report documents work performed to examine how various forms of mercury can migrate through the multiple flowsheets of a solvent extraction system. This entailed 29 single stage distribution tests, a multi-stage Extraction-Scrub-Strip (ESS) test, and a detailed speciation analysis of several salt batch feed samples.

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Solid-State Recrystallization Pathways of Sodium Aluminate Hydroxy Hydrates

Crystallization of Al-bearing solid phases from highly alkaline Na 2 O:Al 2 O 3 :H 2 O solutions commonly necessitates an Al 3+ coordination change from tetrahedral to octahedral, but intermediate coordination states are often difficult to isolate. Here, a similar Al 3+ coordination change process is examined during the solid-state recrystallization of sodium oxohydroxoaluminate hydrate (MSA) to nonasodium bis(hexahydroxy aluminate) trihydroxide hexahydrate (NSA) at ambient temperature. While the MSA structure contains solely polymerized tetrahedral Al 3+ , the NSA structure is a molecular aluminate salt solely based upon monomeric octahedral Al 3+ . Spontaneous recrystallization of MSA and excess sodium hydroxide hydrate into NSA over 3 days of reaction time was clearly evident in X-ray diffractograms and in Raman spectra. In situ 27 Al multiple quantum MAS NMR spectroscopy showed no evidence of intermediate aluminates, suggesting that intermediates, such as pentacoordinate Al 3+ , are short-lived and require more highly time resolved spectroscopies to detect. Such research is advancing upon detailed mechanistic understanding of Al 3+ coordination change mechanisms in these highly alkaline systems, with relevance to aluminum refining, corrosion sciences, and nuclear waste processing.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Evaluation of the Potential for Precipitation of Solids during Storage of Non-Aluminum SNF Solutions

Non-aluminum clad spent nuclear fuels (NASNF) stored in the L-Area basin will be dissolved in H-Canyon using the 6.3D electrolytic dissolver. The solutions will be stored in either the hot or warm canyon until the preparation of a sludge batch for the Defense Waste Processing Facility. Spent nuclear fuel solutions could be stored for 1-2 years before transfer to the H-Area Tank Farm depending on the interval between sludge batches. The solution level in the storage tanks will be maintained; therefore, precipitation of solids due to evaporation is not an issue. However, the precipitation of solids from completely dissolved SNF due to solution instabilities has been observed during intermediate storage of solutions generating hydrated oxides.The presence of fissile material in these solids is generally associated with zirconium molybdate, which is known to act as a host lattice for Pu and can carry the actinides upon precipitation. The formation of zirconium molybdate solids which carry fissile material is a potential concern for the storage of NASNF solutions. To address this concern, the Savannah River National Laboratory performed a literature review to identify knowledge gaps which may require experimental work to determine if the formation of solids is a concern during storage of these solutions. Based on the literature review, the precipitation of zirconium molybdate solids from the Campaign 1 NASNF solutions during intermediatestorage is expected. This conclusion is supported by the identification of zirconium molybdate solids found on the H-Canyon 6.1D Dissolver MK-12 insert spacer. The formation of the zirconium molybdate solids is attributed to hydrolysis and radiolytic processes in the nitric acid solution. As the molybdate solids form, U and Pu can substitute for Zr in the crystal lattice resulting in co-precipitation. Generally, the Pu substitutes directly into the crystal lattice during precipitation while the U associated with the molybdate solids more likely absorbs from the solution. The U in the NASNF solutions is present as uranyl nitrate, a 2+ cation which will not substitute as easily into the molybdate crystal lattice for the Zr 4+ ion.

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SRNL Sludge Batch 10 Qualification SRAT and SME Off-Gas Results

Savannah River National Laboratory (SRNL) completed a small-scale demonstration of the Defense Waste Processing Facility (DWPF) Chemical Process Cell (CPC) utilizing the nitric-glycolic acid (NGA) flowsheet to support Sludge Batch 10 (SB10) qualification. The demonstration utilized a Tank 51 slurry sample washed by SRNL (with added H-canyon material). The purpose of this document is to report the observed off-gas results from the demonstration. With the NGA flowsheet, DWPF has a CPC hydrogen generation limit of 2.4×10 -2 lb/h. The peak observed rate was nearly 90 times less than that limit during SRNL testing.

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M-Star ® Software Test and Verification

Savannah River Mission Completion (SRMC) currently manages the risk for retained hydrogen in the Defense Waste Processing Facility (DWPF) vessels by implementing a Retained Hydrogen Program. The current program relies on Sludge Batch (SB) 8 Gas Chromatograph data and on conservative assumptions concerning gas release and retention to determine allowable vessel Quiescent time (Q-time). The authors identified M-Star ® CFD as a software that could simulate processes such as fluid flow, heat transfer, species transport, chemical reactions, particle transport, and retained hydrogen gas release. Preliminary simulation results suggest that more realistic assumptions on gas retention and release may be feasible for the DWPF retained hydrogen program. Because of the desire to use the M-Star ® software to perform analyses that support nuclear safety, SRMC has requested Savannah River National Laboratory (SRNL) to upgrade the software classification level of M-Star ® CFD from level D to level A to perform analyses that support nuclear safety.

08 HYDROGEN↗

Ion Chromatography (IC) Round Robin Analyses of Low Glycolate Concentrations in Recycle Collection Tank (RCT) Post Permanganate Treatment Simulant

This work is a demonstration of Ion Chromatography (IC) analysis of low concentrations of glycolate in chemical simulant designed to mimic the matrix in the Recycle Collection Tank (RCT) at the Defense Waste Processing Facility (DWPF) after sodium permanganate oxidation treatment. The IC method was previously developed [1] and this report covers the results of round robin testing with three analytical laboratories located at the Savannah River Site (SRS). The laboratories are termed the Sensing & Metrology (S&M) laboratory at the Savannah River National Laboratory (SRNL), the Processing Science Analytical Laboratory (PSAL) at SRNL, and the DWPF laboratory at SRS. Each laboratory received four samples: (1) 200 mL of 21.3 mg/L glycolate in RCT post permanganate strike sulfite quenched simulant, (2) 200 mL of 38.0 mg/L glycolate in RCT post permanganate strike sulfite quenched simulant, (3) 200 mL of 54.9 mg/L glycolate in RCT post permanganate strike sulfite quenched simulant, and (4) 600 mL of RCT post permanganate strike sulfite quenched simulant to use for matrix matched blanks.

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Uranium Solubility and Supernate Enrichment Testing for Accelerated Basin De-Inventory Discards to Concentration, Storage, and Transfer Facilities

After future Accelerated Basin De-inventory (ABD) enriched uranium discards into sludge batches (SB), portions of supernate decants during SB preparation will be blended into salt batches (StB) and the resulting feeds must meet the nuclear safety requirements for the Salt Waste Processing Facility (SWPF). During previous SB10 sampling and testing, which involved H-Canyon material containing enriched uranium being mixed with Tank 51 sludge shortly after the H-Canyon stream was neutralized, it was identified that the uranium isotopic enrichment in the supernate deviated from the uranium isotopic enrichment in the slurry. The higher uranium isotopic enrichment in the supernate introduced the risk of challenging the feed requirements of SWPF. The ABD material added to SB11 was isotopically diluted with depleted uranium, mitigating any downstream impacts. However, H-Canyon desires to eliminate or minimize future depleted uranium additions in order to meet the mission schedule.

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Probing the pH Effect on Boehmite Particles in Water Using Vacuum Ultraviolet Single-Photon Ionization Mass Spectrometry

Boehmite has been widely used in theoretical research and industry, especially for hazardous material processing. For the liquid-phase treating process, the interfacial properties of boehmite are believed to be affected by pH conditions, which change its physicochemical behavior. However, molecular-level detection on cluster ions is challenging when using bulk approaches. Herein we employ in situ vacuum ultraviolet single-photon ionization mass spectrometry (VUV SPI-MS) coupled with a vacuum-compatible microreactor system for analysis at the liquid–vacuum interface (SALVI) to investigate the solute molecular composition of boehmite under different pH conditions for the first time. The mass spectral results show that more complex clustering of solute molecules exists at the solid–liquid (s–l) interface than conventionally perceived in a “simple” aqueous solution. Besides solute ions, such as boehmite molecules and fragments, the composition and appearance energies of these newly discovered solvated cluster ions are determined by VUV SPI-MS in different pH solutions. We offer new results for the pH-dependent effect of boehmite and provide insights into a more detailed solvation mechanism at the s–l interface. By comparing the key products under different pH conditions, fundamental understanding of boehmite dissolution is revealed to assist the engineering design of waste processing and storage solutions.

SALVI↗

Gandolinium poison solubility testing for the downstream impacts from accelerated basin de-inventory

The Accelerated Basin De-inventory (ABD) Program at the Savannah River Site (SRS) is designed to accelerate the de-inventory of L-Basin and accelerate the Spent Nuclear Fuel (SNF) Disposition mission. Spent fuel will be dissolved in H-Canyon without recovery of uranium. The dissolver solutions will be temporarily stored, pH-adjusted to excess hydroxide (which will facilitate precipitation of metal oxides/hydroxides), transferred to the Concentration, Storage, and Transfer Facility (CSTF), and subsequently immobilized in the Defense Waste Processing Facility (DWPF) during planned sludge batch campaigns. ABD accelerates basin closure, significantly reduces programmatic risk, and greatly reduces the lifecycle budget requirements for the site by eliminating the need for a SNF drying and packaging capability. The ABD approach represents a significant change to the clean-up approach for the SRS. However, the increased fissile loading in sludge batches, due to the dissolver solutions, requires investigation to ensure fissile limits are efficiently and safely managed; higher fissile loadings in the glass are projected to be two to three times higher than the current fissile concentration limit of 897 g/m 3 and will be addressed in a future report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Results of Long-Term Analyses of the Salt (Macro) Batch 5 Tank 41H Qualification Sample (Resample)

Savannah River National Laboratory (SRNL) analyzed samples from Tank 41H in support of qualification of Salt Waste Processing Facility (SWPF) Salt Batch 5. This document reports the long-term results of the analyses of samples of Tank 41H. Analysis of the Tank 41H Salt Batch 5 sample indicates that the material does not display any unusual characteristics or observations, such as floating solids, the presence of large amounts of solids, or unusual colors. This memo satisfies Deliverable 3 of the Technical Task Request (TTR). After a data review it was found that the original results reported in the revision 0 of this document were actually from Salt Batch 4 (Tank 21H). A new set of samples were sent for analysis to satisfy Deliverable 3 of the TTR. The results from the original analyses have been replaced with the recent analyses of the actual Salt Batch 5 sample.

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SB10 Frit Recommendation, and Evaluations of the Glass Variability Study and Cs-137 Concentrations in Strip Effluent Based on May 2021 Projections

The Defense Waste Processing Facility (DWPF) is currently preparing to initiate processing of Sludge Batch 10 (SB10), which is comprised of material from Tanks 11H, 13H, 15H, and 26F, Alternate Feed Stock-2 (AFS-2) and Sodium Reactor Experiment (SRE) material from H-Canyon. In support of SB10 qualification, frit development using 2020 Tank 40 blend projections and experimental work for the glass variability study were previously conducted. Frit 473 and Frit 209 were identified as candidate frits and both were included in the development of the variability study test matrix; however, a final frit recommendation was postponed until more information could be determined about the composition of SB10 after washing. In May 2021, Savannah River Remediation (SRR) reprojected SB10 based on the analytical results from the Tank 51 qualification sample that was washed in the Savannah River National Laboratory (SRNL) Shielded Cells Facility.

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Mercury Testing with Sludge Batch 10 Tank 40 Simulant

Savannah River Mission Completion (SRMC) requested that researchers at Savannah River National Laboratory (SRNL) perform testing designed to examine why the Defense Waste Processing Facility (DWPF) is not collecting elemental mercury in the Mercury Water Wash Tank (MWWT). In order for DWPF to recover mercury, mercuric oxide must first be reduced to elemental mercury. The elemental mercury must then be steam stripped, condense, and coalesce in the Mercury Water Wash Tank (MWWT) during chemical processing in the Sludge Receipt and Adjustment Tank (SRAT). The efficiency of these steps was investigated in a series of laboratory scale SRAT experiments under the nitric-glycolic and nitric-formic flowsheets utilizing Momentive Y-17112 and Antifoam 747. Mercury speciation in the Slurry Mix Evaporator Condensate Tank (SMECT) and condensate streams was also examined. The key conclusions from these experiments are as follows: Mercury II Oxide may not be fully reduced to elemental mercury during acid addition at 93°C. Higher temperatures, i.e., boiling may be necessary to fully reduce Mercury II Oxide. The highest percent mercury recovery (71 %) in the MWWT was observed in the MS-NGA-17112 experiment (nitric-glycolic acid flowsheet with Momentive Y-17112), which is how DWPF is currently operating the SRAT.

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Results of the Short-Term Analyses of the Salt (Macro) Batch 5 Tank 41H Qualification Sample (Resample)

Savannah River National Laboratory (SRNL) analyzed samples from Tank 41H in November 2021 in support of qualification of Salt Waste Processing Facility (SWPF) Salt Batch 5. This document reports the short-term results of the analyses of samples of Tank 41H. Analysis of the Tank 41H Salt Batch 5 sample indicates that the material does not display any unusual characteristics or observations, such as floating solids, the presence of large amounts of solids, or unusual colors. Further sample results will be reported in a future document. This report satisfies part of Deliverable 3 of the Technical Task Request (TTR). After a data review it was found that the original results reported in the revision 0 of this document were actually from Salt Batch 4 (Tank 21H). A new set of samples was sent for analysis to satisfy deliverable 3 of the TTR. The results from the original analyses have been replaced with the recent analyses of the actual Salt Batch 5 sample.

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Results of the Short-Term Analyses of the Salt (Macro) Batch 5 Tank 41H Qualification Sample (Resample)

Savannah River National Laboratory (SRNL) analyzed samples from Tank 41H in November 2021 in support of qualification of Salt Waste Processing Facility (SWPF) Salt Batch 5. This document reports the short-term results of the analyses of samples of Tank 41H. Analysis of the Tank 41H Salt Batch 5 sample indicates that the material does not display any unusual characteristics or observations, such as floating solids, the presence of large amounts of solids, or unusual colors. Further sample results will be reported in a future document. This report satisfies part of Deliverable 3 of the Technical Task Request (TTR). After a data review it was found that the original results reported in the revision 0 of this document were actually from Salt Batch 4 (Tank 21H). A new set of samples was sent for analysis to satisfy deliverable 3 of the TTR. The results from the original analyses have been replaced with the recent analyses of the actual Salt Batch 5 sample.

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On silica's roles in controlling americium migration in contaminated sediments

Two studies in the early 1980s described the leaching behavior of americium (Am) disposed as part of acidic high-salt processing wastes from the Hanford Site’s Plutonium Finishing Plant to nearby ground sediments. Here these batch leach experiments showed that the Am concentrations followed a linear log [Am] versus pH relationship with a slope of –1. Column leach experiments in the second study, however, did not follow this relationship and only ~30% of the americium desorbed even after extensive column leaching. Here, the 1980s research is re-examined along with previously unpublished information and, in light of recent published work, a plausible mechanism is proposed to explain these phenomena. Amorphous silica in the contaminated sediments is postulated to be the substrate responsible for both the exchangeable Am available for leaching and the retained low-leachable Am made evident in the column leach experiments. The exchangeable Am 3+ in the contaminated sediment leach experiments behaves with pH dependence similar to that observed for uptake onto amorphous silica of sodium (Na + ), calcium (Ca 2+ ), barium (Ba 2+ ), cadmium (Cd 2+ ), uranyl (UO 2 2+ ), ferric (Fe 3+ ), chromic (Cr 3+ ), cupric (Cu 2+ ), plumbous (Pb 2+ ), uranium(IV) (U 4+ ), plutonium(IV) (Pu 4+ ), zirconium (Zr 4+ ), analogue lanthanide (gadolinium, Gd 3+ , europium, Eu 3+ , and lutetium, Lu 3+ ) and curium (Cm 3+ ) ions as well as other studies with Am 3+ . Correspondingly, the residual low-leachable Am 3+ revealed in the column leach experiments is attributed to incorporation of Am 3+ within amorphous silica by dynamic Am 3+ sorption and silica precipitation processes.

58 GEOSCIENCES↗

Data Qualification Report: SRNL Glass Composition-Properties (ComPro) Database

The Savannah River National Laboratory Glass Composition-Properties (ComPro) database is an extensive database containing pertinent composition and durability data to support the accelerated clean-up mission at the Defense Waste Processing Facility. The activities described in this data qualification report were performed to support the information contained in the database. There were two objectives of the original data qualification process. The first objective was to review supporting documentation to determine if DOE/RW-0333P Quality Assurance Requirements and Description had been implemented during the original work. If the DOE/RW-0333P Quality Assurance Requirements and Description had not been directly implemented during the original work, the second objective was to determine if the controls that were used were adequate to meet the intent of the DOE/RW-0333P Quality Assurance Requirements and Description. The results of these two objectives and the activities performed to support these decisions are described in this document. An assessment of each dataset was made to determine if the data were RW-0333P Compliant, RW-0333P Equivalent or Non-RW-0333P Compliant. The original data qualification was performed in accordance with E7, Conduct of Engineering Manual, Procedure 3.70, Revision 4, Qualification of Data. The specific method that was used was Equivalent Controls as described in E7, 3.70. Revision 2 of this document adds supporting information for the RW-0333P Compliant datasets added to Revision 3 of the database.

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Next Generation Solvent Vapor Pressure Testing

Previous work by Savannah River National Laboratory (SRNL) indicated that the actual Next Generation Solvent vapor pressure would be higher than the Original Caustic Side Solvent Extraction (CSSX) solvent vapor pressure, but below a bounding vapor pressure using Raoult’s Law. Since solvent vapor pressure has a significant impact on Composite Lower Flammability Limits (CLFL) and attendant accident analyses, obtaining an additional margin from the bounding NGS vapor pressure would be valuable. In order to quantify how much margin might be gained from the NGS bounding solvent, SRNL researchers were requested to perform vapor pressure testing with the Next Generation Solvent (NGS) by Savannah River Mission Completion (SRMC). The vapor pressure curve for the NGS formulation set to be deployed at the Salt Waste Processing Facility (SWPF) has been determined by SRNL up to 55°C (131°F) using headspace Gas Chromatography (GC). It was expected that NGS would have a higher vapor pressure than the Original CSSX solvent; however, experimental results indicate a lower vapor pressure. At this time, it is uncertain if this difference is due to the 7x increase in concentration of the large calixarene in the solvent (0.007M BOBCalix in Original CSSX solvent vs. 0.05M MaxCalix in NGS) or to minor batch-to-batch variations in Isopar-L constituents. The NGS and the Original CSSX solvent vapor pressure data were fitted to the Antoine Equation. The Antoine equation gives a more accurate representation of the expected vapor pressure of the solvents outside of the temperature range tested. The Antoine equation fitting for NGS is given below: $P=10^{6.364⁻\frac{1788.4}{T+219.3}}$. Where, p is the vapor pressure (partial pressure) of NGS in mmHg and T is the temperature in °C. It is recommended that SRMC either continue using the more conservative Isopar-L vapor pressure calculations at SWPF with the equation developed for the Original CSSX solvent, or the equation presented above for the NGS solvent. Additionally, it is recommended to study the variability in Isopar-L vapor pressure between lots.

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Literature Review on Next Generation Solvent Isopar ® L Vapor Pressure Curve and the Partitioning of its Modifier and Extractant

The Next Generation Solvent (NGS) is set to replace the Original Caustic Side Solvent Extractant (CSSX) at the Salt Waste Processing Facility (SWPF). The Savannah River National Laboratory (SRNL) was requested by Savannah River Mission Completion (SRMC), formerly Savannah River Remediation (SRR), to perform a literature review on the following topics to address flammability concerns with the current solvent: Isopar ® L vapor pressure curve for NGS, partitioning ratio for the extractant MaxCalix and the modifier Cs-7SB, and high cesium concentration impacts on NGS radiolysis and potential solvent degradation rates in high cesium concentrations. The following conclusions and recommendations are made based on previous experimental work and literature: (1) Current SWPF flammable gas generation calculations use an Isopar ® L vapor pressure curve based on experimental testing with the Original CSSX solvent. No such testing to date has been performed with NGS. It is suggested that the decrease in Cs-7SB concentration for NGS compared to the Original CSSX solvent would lead to a slightly higher vapor pressure at all temperatures in SWPF vessels. A bounding NGS vapor pressure curve has been provided; it is recommended to see if these values would challenge current flammability controls and to perform testing if needed.(2) The partitioning ratio for Cs-7SB is known in the Original CSSX solvent with dilute nitric acid and caustic solutions. No tests could be found for the partitioning of Cs-7SB to dilute boric acid solutions; however, a similar partitioning ratio is expected. Due to the lipophilic alkyl chains on MaxCalix, it is expected to be even less soluble than BOBCalixC6 in the aqueous phase and should not be considered a significant contributor to the f organic term. Additionally, the reaction rate of N,N’,N’’-Tris(3,7-dimethyloctyl)guanidine (TiDG) or its degradation products with a hydrogen radical should be estimated/determined if they are found to be significant contributors to the f organic term. (3) NGS is expected to see much higher Cs concentrations at SWPF in comparison to its use at the Modular CSSX Unit (MCU). These higher Cs concentrations could influence radiolytic degradation rates of the solvent. NGS appears to be fairly stable to radiolytic degradation based on previous testing and its use at MCU. However, there has not been radiolytic flammable gas generation testing with NGS to date. There is a risk that the continued use of G-values obtained for flammable gases produced from the irradiation of the Original CSSX solvent is not bounding for NGS, but this is considered a very low risk due to the similarities in the composition of the solvents, as well as the conservatisms in the experimental design of the Original CSSX testing.

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