Search NASA⌕ Search

SEARCH · Search NASA

Results for “WASTE PROCESSING”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 523 records · Page 29

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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 Here, the 1980s research is re-examined along with previously unpublished information and, in light of recent published work, a plausible mechanism is proposed to explain these phenomena.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↗

Results of Hg speciation testing on DWPF SMECT-4, SMECT-6, and RCT-2 samples

The Savannah River National Laboratory (SRNL) was tasked with preparing and shipping samples for Hg speciation by Eurofins Frontier Global Sciences, Inc. in Seattle, WA on behalf of the Savannah River Remediation (SRR) Mercury Task Team.i,ii The fifteenth shipment of samples was designated to include Defense Waste Processing Facility (DWPF) Slurry Mix Evaporator Condensate Tank (SMECT) samples from Sludge Receipt and Adjustment Tank (SRAT) Batch 738 and a Recycle Condensate Tank (RCT) sample from SRAT Batch 736. The DWPF sample designations for the three samples analyzed are provided in Table 1. The Batch 738 ‘Baseline’ SMECT sample was taken prior to Precipitate Reactor Feed Tank (PRFT) addition and concentration and therefore, precedes the SMECT-5 sample reported previously. iii The Batch 738 ‘End of SRAT Cycle’ SMECT sample was taken at the conclusion of SRAT operations for this batch (PRFT addition/concentration, acid additions, initial concentration, MCU addition, and steam stripping). Batch 738 experienced a sludge slurry carryover event, which introduced sludge solids to the SMECT that were particularly evident in the SMECT-5 sample, but less evident in the ‘End of SRAT Cycle’ SMECT-6 sample. The Batch 736 ‘After SME’ RCT sample was taken after completion of SMECT transfers at the end of the SME cycle.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Nitric-glycolic flowsheet evaluation with the slurry-fed melt rate furnace

The Savannah River National Laboratory (SRNL) was tasked to support validation of the Defense Waste Processing Facility (DWPF) melter offgas flammability model for the nitric-glycolic (NG) flowsheet. The work supports Deliverable 4 of the DWPF & Saltstone Facility Engineering Technical Task Request (TTR)1 and is supplemental to the Cold Cap Evaluation Furnace (CEF) testing conducted in 2014.2 The Slurry-fed Melt Rate Furnace (SMRF) was selected for the supplemental testing as it requires significantly less resources than the CEF and could provide a tool for more rapid analysis of melter feeds in the future. The SMRF platform has been used previously to evaluate melt rate behavior of DWPF glasses, but was modified to accommodate analysis of the offgas stream. Additionally, the Melt Rate Furnace (MRF) and Quartz Melt Rate Furnace (QMRF) were utilized for evaluations. MRF data was used exclusively for melt behavior observations and REDuction/OXidation (REDOX) prediction comparisons and will be briefly discussed in conjunction with its support of the SMRF testing. The QMRF was operated similarly to the SMRF for the same TTR task, but will be discussed in a separate future report. The overall objectives of the SMRF testing were to; 1) Evaluate the efficacy of the SMRF as a platform for steady state melter testing with continuous feeding and offgas analysis; and 2) Generate supplemental melter offgas flammability data to support the melter offgas flammability modelling effort for DWPF implementation of the NG flowsheet.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DWPF Melter Off-Gas Chemistry Model for the Nitric-Glycolic Acid Flowsheet

The Defense Waste Processing Facility (DWPF) melter off-gas flammability model for the current nitricformic acid (NFA) flowsheet was revised in Revision 0 to reflect the upcoming reductant change from formic to glycolic acid. The necessary data for the model revision was obtained during Phase 2 of the Coldcap Evaluation Furnace (CEF) test, which was completed on March 21, 2014 after 24 days of round-theclock feeding and pouring, including a total downtime of ~20 hours due to equipment failures. Specifically, the CEF was run to generate steady state melter off-gas data during the first 10 days using two nitricglycolic acid (NGA) flowsheet feeds prepared at 100% and 125% acid stoichiometry from the same Sludge Batch 6 (SB6) simulant used during the Phase 1 CEF test in 2013. Each feed was fed under both bubbled and non-bubbled conditions, while maintaining the CEF vapor space temperature constant for at least two hours approximately at 700, 600, 500, 400, 350, and <300 °C for a total of 26 steady state runs. During the next 8 days, the CEF was run at the nominal vapor space temperature of 700 °C to generate the off-gas surge data using the 100% acid-stoichiometry feed under bubbled and non-bubbled conditions. The remaining days were spent on producing additional data using the 100% acid-stoichiometry feed spiked with excess antifoam as well as re-running some of the earlier steady state runs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Submersible Blend Pump Mixing Evaluation

Savannah River Remediation (SRR) requested Savannah River National Laboratory (SRNL) to perform an engineering analysis to investigate the mixing behavior in feed blend tanks for the Salt Waste Processing Facility (SWPF). The baseline for mixing the feed blend tanks is a U 0 D of 5.1 ft 2 /s, a pump nozzle elevation of 165.86 inches above the sludge layer and 174 inches above the tank bottom, a liquid level 174.28 inches above the pump nozzle, and a blend time of 10.63 hours. The current Salt Batch Plan identifies Tanks 21, 26, 41, and 42 as the designated blend tanks. This supplemental study only addresses Type IIIA tanks (i.e., Tanks 26, 41, and 42). Using the Tank 41 current configuration as the baseline, the analysis considers the impact on blending if: 1) the distance between the pump discharge nozzle and the liquid surface increases above the baseline, 2) the distance between the pump discharge nozzle and the liquid surface decreases below the baseline, 3) the distance between the pump discharge nozzle and the sludge layer at the bottom of the tank increases above the baseline, or 4) the distance between the pump discharge nozzle and the sludge layer at the bottom of the tank decreases below the baseline, each while maintaining the minimum U 0 D of 5.1 ft 2 /s. Per subsequent direction from SRR, the minimum distance between the pump discharge nozzle and the sludge will be 165.86 inches. Hence, distances less than 165.86 inches between the pump discharge nozzle and the sludge were not included in the analysis. In addition, SRR requested SRNL to evaluate the impact of the abandoned in place, failed slurry pumps in Tank 42 on liquid blending.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Tank 40 final sludge batch 9 chemical and fissile radionuclide characterization results

A sample of Sludge Batch (SB) 9 was pulled from Tank 40 in order to obtain radionuclide inventory analyses necessary for compliance with the Waste Acceptance Product Specifications (WAPS)i. The SB9 WAPS sample was also analyzed for chemical composition, including noble metals, and fissile constituents, and these results are reported here. These analyses along with the WAPS radionuclide analyses will help define the composition of the sludge in Tank 40 that is fed to the Defense Waste Processing Facility (DWPF) as SB9. At the Savannah River National Laboratory (SRNL), the 3-L Tank 40 SB9 sample was transferred from the shipping container into a 4-L high density polyethylene bottle and solids were allowed to settle. Supernate was then siphoned off and circulated through the shipping container to complete the transfer of the sample. Following thorough mixing of the 3-L sample, a 547 g sub-sample was removed. This sub-sample was then utilized for all subsequent slurry sample preparations. Eight separate aliquots of the slurry were digested, four with HNO3/HCl (aqua regiaii) in sealed Teflon® vessels and four with NaOH/Na2O2 (alkali or peroxide fusioniii) using Zr crucibles. Three Analytical Reference Glass – 1iv (ARG-1) standards were digested along with a blank for each preparation. Each aqua regia digestion and blank was diluted to 1:100 with deionized water and submitted to Analytical Development (AD) for inductively coupled plasma – atomic emission spectroscopy (ICP-AES) analysis, inductively coupled plasma – mass spectrometry (ICP-MS) analysis, atomic absorption spectroscopy (AA) for As and Se, and cold vapor atomic absorption spectroscopy (CV-AA) for Hg. Equivalent dilutions of the alkali fusion digestions and blank were submitted to AD for ICP-AES analysis. Tank 40 SB9 supernate was collected from a mixed slurry sample in the SRNL Shielded Cells and submitted to AD for ICP-AES, ion chromatography (IC), total base/free OH-/other base, total inorganic carbon/total organic carbon (TIC/TOC) analyses. Weighted dilutions of slurry were submitted for IC, TIC/TOC, and total base/free OH-/other base analyses. Activities for U-233, U-235, and Pu-239 were determined from the ICP-MS data for the aqua regia digestions of the SB9 WAPS slurry using the specific activity of each isotope. The Pu-241 value was determined from a Pu-238/-241 method developed by SRNL AD and previously described.v

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Assessment of Cold Spraying Mild-Steel Powder onto Mild-Steel Substrate for Development of Waste Tank Repair System - 20515

Infrastructure for nuclear waste storage and processing (e.g., waste tanks, transfer pipelines, evaporators) at the U.S. Department of Energy (DOE) Hanford Site has been in service well beyond its original design life. Past corrosion control methods at Hanford for waste tanks have not been fully successful, as evidenced by leaks forming in tank AY-102, the first double-shell tank (DST) constructed, and significant degradation detected in the primary tanks of AP-102 and AY-101. Failure of infrastructure components can pose a significant risk to the environment, worker safety, and project schedules. Therefore, remotely deployable methods for repairing and extending the service life of the existing tank waste infrastructure are needed. Cold spray has been identified as a potential technique for performing in situ repairs and material buildup for life extension of waste storage and processing infrastructure. Cold-spray repair methods and technology are well-established and are used extensively by industry and the military for making repairs using dissimilar materials. Pacific Northwest National Laboratory is evaluating the process of making repairs using mild-steel powder applied to a mild-steel substrate to minimize the galvanic corrosion potential. This paper presents the initial results for a feasibility ('go/no-go') assessment as to whether low-density deposits with sound metallurgic bonds, to build-back steel plate material, can be achieved using existing commercial cold-spray equipment. An assessment of steel-on-steel deposit is necessary to determine if cold spray is a viable candidate for DST carbon steel plate restoration/repair and thus DST service life extension. Powder properties and preparation requirements have been investigated and operating conditions and parameters for depositing a cold-spray buildup of mild steel on a mild-steel substrate have been determined. Soundness of metallurgic bonding as a function of the operating parameters exercised and results from applying cold spray to lay a deposit over simulated small-diameter flaws in 1/2-inch mild-steel plate are presented. Application of in situ cold-spray repair to the DOE complex has the potential to reduce cost and schedule impacts associated with component failures and the need to procure and construct replacement infrastructure. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Assessing I 2 S-LWR economic competitiveness using systematic differential capital cost evaluation methodology

One of the major factors impeding nuclear power plant (NPP) construction is their economics. In this study, a systematic differential economics evaluation approach was developed through the use of the Code of Accounts guidelines to assess the costs of nuclear power plants. The methodology, entirely based on publicly available data, may serve as a template to evaluate direct costs for reactors of any size and design at any stage of developments. In particular, this approach was used to assess the costs of the Integral Inherently Safe Light Water Reactor (I 2 S-LWR). The I 2 S-LWR is a design concept of a large (~1000 MWe) light water reactor. One of its key design features promoting inherent safety is implementation of an integral primary circuit configuration, which however necessitates a compact design of the core and primary circuit components. Through the methodology here presented, a representative loop PWR design was taken as a reference and the differential cost was estimated for each individual account based on the design difference, or similarity. Cost scaling techniques were applied to the accounts representing systems that differ from the ones of the reference PWR. Cost estimating techniques were used to evaluate cost of innovative components that are not part of standard PWR designs. By evaluating the cost difference of the I 2 S-LWR from the standard PWR, rather than the absolute cost, the uncertainty in the estimate is reduced. A similar approach was used by ORNL to estimate the cost of a Fluoride-salt High-temperature Reactor (FHR). A traditional four-loop PWR plant with a core thermal power of 3417 MWth (1144 MWe) was selected as the reference. Costs for that plant were prepared in 1978 by the Department of Energy (DOE) Energy Economics Data Base (EEDB), averaging actual cost incurred in the construction of several nuclear power plants (NPP), itemized with a great level of detail according to the Code of Accounts. This best estimate costs are denoted PWR12-BE. For each account, the cost of equipment, site labor and site material is provided. Industry experts at Westinghouse Electric Company performed a “sanity check” of the cost items, adjusting the cost of several items to match the current market and supply chain data. The detailed cost assessment of I 2 S-LWR was performed, systematically analyzing cost for each account, and applying the differential economics approach. First, Relative importance of each account, i.e., its contribution to the total cost was established, to help focus analysis on the most significant contributors. Moreover, the accounts describing components that are different than that of the PWR12-BE were identified. The integral configuration of the reactor has important implications on the cost of the reactor plant equipment (accounts 22x). Turbine generator equipment (Accounts 23x) is not believed to be much different than that of the reference design. I 2 S-LWR structures (Accounts 21x) mainly differ from that of a standard LWR as several buildings (containment building, shield building, annex building, waste processing building, fuel storage building) are integrated into a single building (nuclear island). Yardwork has a higher cost for the I 2 S-LWR, as the NI is partially below grade. On the other hand, due to its compact Nuclear Island footprint, I 2 S-LWR facilitates (and includes in its reference version) the use of seismic isolators, which contribute to reducing the direct cost. The total capital investment cost (TCIC), on the $/kW basis, of I 2 S-LWR is 5.84% lower than that of PWR12-BE, in spite of the lower power output of I 2 S-LWR. However, for the Western US (0.7 g), benefits of the seismic isolation are more pronounced, and the I 2 S-LWR total capital investment cost is 13.02% lower than that of PWR12-BE. If the I 2 S-LWR is compared to a PWR10-BE (traditional loop design, but scaled to the same power level as I 2 S-LWR), the savings are even higher, in the range 11.12%–17.89%. The analysis indicates that I 2 S-LWR has potential to offer an economically attractive design, with TCIC lower than that of a nuclear power plant based on a traditional loop PWR design. In other words, I 2 S-LWR design offers significantly enhanced safety, while at the same time improving economics. The differential economics approach developed in this paper can be used in identifying changes in cost of key components in order to improve the economics of a nuclear reactor design. The method can also help compare the economics of advanced Generation IV reactors and innovative water-cooled SMR (Small Modular Reactors) with respect to standard LWR technologies. In summary, the differential economics approach and can be used as a tool capable of helping stakeholder decisions.

Economic↗

The relative viscosity of NaNO 3 and NaNO 2 aqueous solutions

In aqueous solution, both nitrate and nitrite are planar, monovalent, and have the same elements but different sizes and charge densities. Comparing the viscosity of NaNO 2 and NaNO 3 aqueous solutions provides an opportunity to determine the relative importance of anion size versus strength of anion interaction with water. The viscosity of aqueous NaNO 2 and NaNO 3 were measured over a temperature and concentration range relevant to nuclear waste processing. The viscosity of NaNO 2 solutions was consistently larger than NaNO 3 under all conditions, even though nitrate is larger than nitrite. This was interpreted in terms of quantum mechanical charge field molecular dynamics calculations that indicate that nitrite forms more and stronger hydrogen bonds with water per oxygen atom than nitrate. Furthermore, these hydrogen bonds inhibit rotational motion required for fluid flow, thus increasing the nitrite solution viscosity relative to that of an equivalent nitrate solution.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Municipal wastewater affects adipose deposition in male mice and increases 3T3-L1 cell differentiation

Trace concentration of EDs (endocrine disrupting compounds) in water bodies caused by wastewater treatment plant effluents is a recognized problem for the health of aquatic organisms and their potential to affect human health. In this paper we show that continuous exposure of male mice from early development to the adult life (140 days) to unrestricted drinking of wastewater collected from a municipal sewage treatment plant, is associated with an increased adipose deposition and weight gain during adulthood because of altered body homeostasis. In parallel, bisphenol A (BPA) at the administration dose of 5 μg/kg/body weight, shows an increasing effect on total body weight and fat mass. In vitro, a solid phase extract (SPE) of the wastewater (eTW), caused stimulation of 3T3-L1 adipocyte differentiation at dilutions of 0.4 and 1 % in the final culture medium which contained a concentration of BPA of 40 nM and 90 nM respectively. Pure BPA also promoted adipocytes differentiation at the concentration of 50 and 80 μM. BPA effect in 3T3-L1 cells was associated to the specific activation of the estrogen receptor alpha (ERα) in undifferentiated cells and the estrogen receptor beta (ERβ) in differentiated cells. BPA also activated the Peroxisome Proliferator Activated Receptor gamma (PPARγ) upregulating a minimal 3XPPARE luciferase reporter and the PPARγ-target promoter of the aP2 gene in adipose cells, while it was not effective in preadipocytes. The pure estrogen receptor agonist diethylstilbestrol (DES) played an opposite action to that of BPA inhibiting PPARγ activity in adipocytes, preventing cell differentiation, activating ERα in preadipocytes and inhibiting ERα and ERβ regulation in adipocytes. The results of this work show that the drinking of chemically-contaminated wastewater promotes fat deposition in male mice and that EDs present in sewage are likely responsible for this effect through a nuclear receptor-mediated mechanism. - Highlights: • Sewage watewaters (TW) sampled from a Municipal Treatment Sewage Plant are biologically active in mice and in vitro. • In male mice TW promotes an increase in adipose mass deposition following exposure from foetal to adult life. A similar effect is produced by BPA. • The expression of key genes of the adipose metabolism is increased in epidydimal fat both by TW and BPA. • An extract from the TW which concentrated the contained chemicals by a factor of 1000, is able to stimulate 3T3-L1 adipocyte differentiation. • The major chemicals contained in the TW, bisphenol A and nonylphenol, are able to promote adipocyte differentiation at the low environmentally relevant doses as are found in the TW.

60 APPLIED LIFE SCIENCES↗

Impact of different catalysis supported by oyster shells on the pyrolysis of tyre wastes in a single and a double fixed bed reactor

Highlights: • An innovative way in the disposal of tyre wastes with energy recovery by pyrolysis is proposed. • Performance of catalysts is improved by spreading them over layers of oyster shells. • One or two levels of supported catalytic beds may favor the quantity and/or the calorific value of the pyrolysis by-products. - Abstract: The treatment and disposal of tyres from vehicles has long been of considerable environmental importance. Studies have been undertaken to reduce their environmental impact. In this study, an alternative gas was produced from automobile tyre wastes by the means of a controlled pyrolysis. To do so, a novel catalytic system was designed with the aim of increasing the rate of conversion and improving the quality of the pyrolysis products. This work aimed also to reduce the severity of the overall reactions, by using powder catalysts (MgO, Al{sub 2}O{sub 3}, CaCO{sub 3}, and zeolite ZSM-5) uniformly distributed on two layers of oyster shells (OS) particles. The catalyst/tyres mass ratio was kept for all the tests at 1/30. The pyrolysis reactor was maintained at 500 °C and the influence of each catalyst and of the number of shell beds (0, 1 or 2), on the yield and composition of the derived products, was examined. The gas yields could contribute by 1.2% of total consumption in Tunisia. Furthermore, some combinations could upgrade the derived gas and made it possible to use it as such or with the minimum of post-treatment. It was found that, with the use of supported catalyst, the gas produced is 45% greater compared to classical thermal pyrolysis. The Heating value of the produced gas was also improved by the use of supported catalysts; it was found 16% greater with the use of Al{sub 2}O{sub 3}/OS compared to non-catalytic pyrolysis. When compared to the gas obtained from only one catalytic supported bed, the sulfur content was reduced by 80% with the use of CaCO{sub 3}/OS on two catalytic beds.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Computational Investigation of Technetium(IV) Incorporation into Inverse Spinels: Magnetite (Fe 3 O 4 ) and Trevorite (NiFe 2 O 4 )

Iron oxides and oxyhydroxides play an important role in minimizing the mobility of redox-sensitive elements in engineered and natural environments. For the radionuclide technetium-99 (Tc), these phases hold promise as primary hosts for increasing Tc loading into glass waste form matrices, or as secondary sinks during the long-term storage of nuclear materials. Recent experiments show that the inverse spinel, magnetite [Fe(II)Fe(III)2O4], can incorporate Tc(IV) into its octahedral sub-lattice, and in that same class of materials, trevorite [Ni(II)Fe(III)2O4] is also being investigated for its ability to host Tc(IV). However, questions remain regarding the most energetically favorable charge-compensation mechanism for Tc(IV) incorporation in each structure, which will affect Tc behavior under changing waste processing or storage conditions. Here, quantum-mechanical methods were used to evaluate incorporation energies and optimized lattice bonding environments for three different, charge-balanced Tc(IV) incorporation mechanisms in magnetite and trevorite. In both cases, the removal of two octahedral Fe(II) or Ni(II) ions upon the addition of Tc(IV) to an octahedral site is the most stable mechanism, relative to the creation of octahedral Fe(III) defects or increasing octahedral Fe(II) content. Following hydration-energy corrections, Tc(IV) incorporation into magnetite is energetically favorable while an energy barrier exists for trevorite.

Environmental Molecular Sciences Laboratory↗

Coordination Chemistry of a Strongly-Donating Hydroxylamine with Early Actinides: An Investigation of Redox Properties and Electronic Structure

Separations of f-block elements are a critical aspect of nuclear waste processing. Redox-based separations offer promise, but challenges remain in stabilizing and differentiating actinides in high oxidation states. The investigation of new ligand types that provide thermodynamic stabilization to high-valent actinides is essential for expanding their fundamental chemistry and to elaborate new separation techniques and storage methods. We report herein the preparation and characterization of Th and U complexes of the pyridyl-hydroxylamine ligand, N-tert-butyl-N-(pyridin-2-yl)hydroxylamine (pyNO–). Electrochemical studies performed on the homoleptic complexes [M(pyNO) 4 ] (M = Th, U) revealed significant stabilization of the U complex upon one-electron oxidation. The salt [U(pyNO) 4 ] + was isolated by chemical oxidation of [U(pyNO) 4 ]; spectroscopic and computational data support assignment as a U V cation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Stabilization of an Unprecedented Hexanuclear Secondary Building Unit in a Thorium-Based Metal–Organic Framework

The crystal structures of thorium clusters are important for understanding the formation and transformation mechanisms of actinide species in solution, which can in turns benefit nuclear waste processing and management. However, stabilizing thorium clusters in aqueous solution is quite challenging because of their fast olation and oxolation reactions. Here, we report a thorium-based meta-organic framework, NU-905, with the formula [Th 6 (μ 3 -O) 2 (HCOO) 4 (H 2 O) 6 ] [TCPP = tetrakis(4-carboxyphenyl)porphyrin], synthesized by a solvothermal reaction in N,N-dimethylformamide and water at 120 °C. NU-905 contains a hexanuclear secondary building unit (SBU), [Th 6 (μ 3 -O) 2 (HCOO) 4 (H 2 O) 6 ], which has never been reported previously. The SBUs are capped and bridged by the tetratopic linker TCPP to form a three-dimensional network with scu topology. The activated NU-905 exhibits permanent porosity and shows high catalytic activity for the selective photooxidation of a mustard gas simulant.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗