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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.

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At least 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
NASA NTRS2005

Effects of Bedrock Landsliding on Cosmogenically Determined Erosion Rates

The successful quantification of long-term erosion rates underpins our understanding of landscape. formation, the topographic evolution of mountain ranges, and the mass balance within active orogens. The measurement of in situ-produced cosmogenic radionuclides (CRNs) in fluvial and alluvial sediments is perhaps the method with the greatest ability to provide such long-term erosion rates. In active orogens, however, deep-seated bedrock landsliding is an important erosional process, the effect of which on CRN-derived erosion rates is largely unquantified. We present a numerical simulation of cosmogenic nuclide production and distribution in landslide-dominated catchments to address the effect of bedrock landsliding on cosmogenic erosion rates in actively eroding landscapes. Results of the simulation indicate that the temporal stability of erosion rates determined from CRN concentrations in sediment decreases with increased ratios of landsliding to sediment detachment rates within a given catchment area, and that larger catchment areas must be sampled with increased frequency of landsliding in order to accurately evaluate long-term erosion rates. In addition, results of this simulation suggest that sediment sampling for CRNs is the appropriate method for determining long-term erosion rates in regions dominated by mass-wasting processes, while bedrock surface sampling for CRNs is generally an ineffective means of determining long-term erosion rates. Response times of CRN concentrations to changes in erosion rate indicate that climatically driven cycles of erosion may be detected relatively quickly after such changes occur, but that complete equilibration of CRN concentrations to new erosional conditions may take tens of thousands of years. Simulation results of CRN erosion rates are compared with a new, rich dataset of CRN concentrations from the Nepalese Himalaya, supporting conclusions drawn from the simulation.

Niemi, Nathan↗
NASA NTRS2000

System Design Techniques for Reducing the Power Requirements of Advanced life Support Systems

The high power requirement associated with overall operation of regenerative life support systems is a critical Z:p technological challenge. Optimization of individual processors alone will not be sufficient to produce an optimized system. System studies must be used in order to improve the overall efficiency of life support systems. Current research efforts at NASA Ames Research Center are aimed at developing approaches for reducing system power and energy usage in advanced life support systems. System energy integration and energy reuse techniques are being applied to advanced life support, in addition to advanced control methods for efficient distribution of power and thermal resources. An overview of current results of this work will be presented. The development of integrated system designs that reuse waste heat from sources such as crop lighting and solid waste processing systems will reduce overall power and cooling requirements. Using an energy integration technique known as Pinch analysis, system heat exchange designs are being developed that match hot and cold streams according to specific design principles. For various designs, the potential savings for power, heating and cooling are being identified and quantified. The use of state-of-the-art control methods for distribution of resources, such as system cooling water or electrical power, will also reduce overall power and cooling requirements. Control algorithms are being developed which dynamically adjust the use of system resources by the various subsystems and components in order to achieve an overall goal, such as smoothing of power usage and/or heat rejection profiles, while maintaining adequate reserves of food, water, oxygen, and other consumables, and preventing excessive build-up of waste materials. Reductions in the peak loading of the power and thermal systems will lead to lower overall requirements. Computer simulation models are being used to test various control system designs.

Finn, Cory↗
DOE OSTI2020

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↗
DOE OSTI2024

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Comparison of Exhibited Wear in Tefzel?, Kynar?, and UHMWPE Valve Seats Exposed to Cyclic Operations with Abrasive Slurry - 17519

Pacific Northwest National Laboratory (PNNL) conducted abrasive-wear testing of ball valves used for double valve isolation in nuclear waste process lines. The valve seat materials consisted of Kynar{sup R}, Tefzel{sup R}, and ultra-high-molecular-weight polyethylene (UHMWPE). Cycle testing was conducted for 1500 operating cycles for the Kynar{sup R} and Tefzel{sup R} valve seats and 5500 cycles for the UHMWPE valve seats. A valve operational cycle consisted of slurry flow, slurry gravity drain, flush water flow, and flush water gravity drain. Both two-way and three-way valves were evaluated, and valves were configured in simulated manifolds so the impact of serving various operational functions (e.g., slurry, flush, or drain control) could be evaluated as some valves are flushed prior to being cycled and others are not. Testing was conducted in PNNL's Multi-Phase Transport Evaluation Loop Facility using a performance-based simulant developed by PNNL specifically for this test program to emulate abrasive slurries for Hanford Tank Farms Waste Transfer System during single-shell tank retrieval and waste feed delivery to the Hanford Tank Waste Treatment and Immobilization Plant. Pre- and post-test valve tear-down evaluations were conducted to observe changes in the condition of internal components for the various seat materials. The various forms of physical wear observed for the different seat materials are presented. The wear results can be used to assess seat materials and valve configurations for application to waste transfer operations, and the observed wear mechanisms can provide insight into assessing valve service life. This information will be used to identify and select additional valve designs that should be evaluated for certification for future use. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Analysis of the Sludge Batch 8 (Macrobatch 10) DWPF Pour Stream Glass Sample

To meet the objectives of the Defense Waste Processing Facility (DWPF) Glass Product Control Program, DWPF and Saltstone Facility Engineering (D&S-FE) requested that Savannah River National Laboratory (SRNL) analyze a radioactive glass pour stream (PS) sample collected during the processing of Sludge Batch 8 (SB8), also known as Macrobatch 10 (MB10). The analyses requested by D&S-FE (listed as tasks one through five in the Technical Task Request (TTR) included determination of the chemical and radionuclide composition, performance of the Product Consistency Test (PCT), and measurement of glass density and REDOX state. The tasks associated with accomplishing these analyses were controlled under a Task Technical and Quality Assurance Plan (TTQAP) and Analytical Study Plan. This report provides results from the requested analyses and satisfies deliverable number five of the TTR.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

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↗
DOE OSTI2021

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Thermolytic Hydrogen Generation Testing of Tank 22 Material

Savannah River National Laboratory performed hydrogen generation rate (HGR) testing using Tank 22 actual-waste sample material. The main objective of the testing was to determine the thermolysis HGR for Tank 22 material with and without added glycolate. An important use of the data from actual-waste sample testing is to validate equations for the Savannah River Site Concentration, Storage, and Transfer Facility (CSTF) HGR being developed through simulated-waste testing. Not exclusive to providing data for model validation, the desired outcome of this testing was to demonstrate a negligible thermolytic HGR in Tank 22 from the contribution of glycolate from the future Defense Waste Processing Facility (DWPF) Nitric-Glycolic Acid (NGA) flowsheet to Tank 22. Tank 22 is the pathway for glycolate in the DWPF recycle stream to enter the CSTF. Testing with added glycolate used the estimated future concentration of glycolate based on DWPF recycle with limited Sludge Receipt and Adjustment Tank and Slurry Mix Evaporator (SME) material carry-over but without glycolate mitigation in the Recycle Collection Tank. Minor objectives include the determination of the impact of the glycolate source used in testing and the inclusion of sludge solids in testing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

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.

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

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↗
DOE OSTI2021

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↗
DOE OSTI2018

Low Temperature Aluminum Dissolution (LTAD) Real Waste Testing of the November 2017 Tank 51 Slurry Sample

Low Temperature Aluminum Dissolution (LTAD) will be performed on Tank 51 sludge prior to sending Sludge Batch 10 to the Defense Waste Processing Facility (DWPF), such that a sizable fraction of the insoluble aluminum will be dissolved and removed. During LTAD, the free hydroxide concentration and temperature of the sludge slurry are both raised, resulting in significantly increased aluminum solubility, which facilitates aluminum removal via decanting of the liquid phase. This Research and Development task will provide input information regarding the expected gas generation rates under the bounding temperature and free hydroxide conditions of the upcoming Tank 51 LTAD evolution. It will also provide information on the expected fraction of insoluble aluminum that will dissolve as a function of temperature and time, as well as the expected chemical and phys ical properties of the Tank 51 sludge slurry, prior to and following the LTAD operations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

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↗