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At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Compendium of Phase-I Mini-SHINE Experiments

Argonne National Laboratory is assisting SHINE Medical Technologies in their efforts to develop the technology to become a domestic Mo-99 producer using low-enriched uranium (LEU). Mini-SHINE experiments are being performed with the high-current electron linear accelerator (linac) at Argonne. The target solution is a 90-150 g-U/L LEU uranyl sulfate at pH 1. In Phase 1, the convertor was tantalum with a maximum beam power on the convertor of 10 kW, and the target solution was limited to 5 L. This configuration generated a peak fission power density of 0.05 W/mL. Nine experiments were performed between February and October 2015. Results are reported and discussed for each experiment regarding the off-gas analysis system, the sampling and Mo-recovery operation, and the Mo-product concentration and purification system. In Phase 2, the convertor will be depleted uranium; beam power will increase to 20 kW; and the solution volume will be 18 L. This configuration will generate a fission power density of up to 1 W/mL.

43 PARTICLE ACCELERATORS↗
DOE OSTI2017

DFLAW Glass and Feed Qualifications to Support WTP Start-Up and Flow-Sheet Development (Final Report, Rev. 0)

In each test, a distinct feed composition was used to determine the effects of feed composition changes on the retention of technetium and other species as detailed in the Test Plan for this work. Key operating parameters such as the glass temperature (nominally 1150°C) were held constant to investigate the effect of each new feed composition on the retention of technetium. The processing rate was maintained at ≈2,250 kg/m 2 /day of glass, which corresponds to the “stretch-goal” rate of 22.5 MT/day specified by ORP for the WTP LAW melter, or ≈2,000 kg/m 2 /day of glass, which corresponds to the enhanced baseline specified by ORP for the WTP LAW melter (as compared to the nominal 1500 kg/m 2 /day). The bubbling rate was adjusted to achieve a near-complete cold-cap (90-100% of melt surface covered with feed). Quantitative measurements of glass production rates, melter operating conditions (temperatures, pressures, power, flows, etc.), gaseous emissions by Fourier Transform Infrared Spectroscopy (FTIR), and particulate emissions using Environmental Protection Agency (EPA) methods were made for each test. Discharged and melt pool glass samples were inspected for formation of any salt phases and analyzed for chemical composition, including technetium content. Samples of the process fluids from each of the off-gas treatment system components were collected and analyzed. Following each test, the off-gas system was partially disassembled and extensively sampled for any residual feed components residing within the off-gas system. Sampling and analysis during and subsequent to each test was intended to be sufficient for supporting a mass balance for all constituents of interest and particularly technetium, rhenium, iodine, sulfur, chlorine, and fluorine. Additional sampling and analysis of evaporator concentrates and condensates was conducted for various organic compounds.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

A system has been developed to perform measurements on irradiated, sodium bonded-metallic fuel elements to determine the amount of fission gas retained in the fuel material after release of the gas to the element plenum. During irradiation of metallic fuel elements, most of the fission gas developed is released from the fuel and captured in the gas plenums of the fuel elements. A significant amount of fission gas, however, remains captured in closed porosities which develop in the fuel during irradiation. Additionally, some gas is trapped in open porosity but sealed off from the plenum by frozen bond sodium after the element has cooled in the hot cell. The Retained fission Gas (RFG) system has been designed, tested and implemented to capture and measure the quantity of retained fission gas in characterized cut pieces of sodium bonded metallic fuel. Fuel pieces are loaded into the apparatus along with a prescribed amount of iron powder, which is used to create a relatively low melting, eutectic composition as the iron diffuses into the fuel. The apparatus is sealed, evacuated, and then heated to temperatures in excess of the eutectic melting point. Retained fission gas release is monitored by pressure transducers during the heating phase, thus monitoring for release of fission gas as first the bond sodium melts and then the fuel. A separate hot cell system is used to sample the gas in the apparatus and also characterize the volume of the apparatus thus permitting the calculation of the total fission gas release from the fuel element samples along with analysis of the gas composition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Test Results from Gamma Irradiation of Aluminum Oxyhydroxides

Hydrated metal oxides or oxyhydroxides boehmite and gibbsite that can form on spent aluminum-clad nuclear fuel assemblies during in-core and post-discharge wet storage were exposed as granular powders to gamma irradiation in a 60 Co irradiator in closed laboratory test vessels with air and with argon as separate cover gases. The results show that boehmite readily evolves hydrogen with exposure up to a dose of 1.8 x 10 8 rad, the maximum tested, in both a full-dried and moist condition of the powder, whereas only a very small measurable quantity of hydrogen was generated from the granular powder of gibbsite. Specific information on the test setup, sample characteristics, sample preparation, irradiation, and gas analysis are described.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Analysis of Defense Waste Processing Facility Sample: Off-Gas Condensate Tank Sample Batch 27862

The Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (now Savannah River Mission Completion (SRMC)), through a Technical Task Request (X-TTR-H-00107), to characterize the “as-received” Off-Gas Condensate Tank (OGCT) Sample identified as sample batch 27862, which was delivered to SRNL Shielded Cells on April 30, 2021. This OGCT report is the third of three sample characterization reports dealing with this Defense Waste Processing Facility (DWPF) Recycle Diversion Project. The other two recycling stream characterization reports are the Recycle Collection Tank (RCT) and the Slurry Mix Evaporator Condensate Tank (SMECT) sample report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Mercury speciation using microcolumns with a direct mercury analyzer (DMA). Development of analytical method for the Savannah River Site liquid system

Mercury in the Savannah River Site (SRS) Liquid Waste System (LWS) exists in various forms, including: a) ionic inorganic mercury, organomercury (e.g., methyl Hg), and other less abundant species dissolved in LWS fluids, b) mercury solids such as oxides, hydroxides, amalgams, sulfides, and sorbed mercury, c) accumulations of dense liquid elemental mercury, and d) vapor phase elemental and organomercury mercury in tank headspace gas and in evaporators. Strategic and proactive management of the estimated 60,000 Kg mercury in the LWS requires deployment of efficient and effective paradigms for sampling and analysis of the total quantity of mercury in the various physical-chemical forms. Scoping tests of selective gas-phase sorbents for mercury speciation were performed to support LWS objectives. This research specifically focused on developing and testing streamlined methods to obtain high-quality analytical results for gaseous mercury species. The proposed methods are straightforward, using small columns (microcolumns) filled with materials that selectively sorb one or more target forms of mercury. Simple differentiation of mercury species and quantification is then achieved using an efficient thermal-desorption-based total mercury analyzer as the final step. Benefits associated with transitioning from multistep procedures such as EPA Methods 245.1, 1630 and 1631to streamlined microcolumn approaches include reduced analysis time, labor, and waste generation. Such transitioning would maximize the value of the current deployment of direct mercury analysis (EPA Method 7473) at SRS. This work has two target end uses: 1) simplifying the analysis steps and reducing costs for liquid samples (e.g., developing a total organic mercury (TOM) method), and 2) providing options for high-quality onsite analysis and rapid turnaround for gas samples in support of SRS LWS industrial hygiene (IH) objectives.

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

Trials and tribulations in the first three years in operation of the SSAMS for biomedical 14 C-AMS at LLNL

We report on the first several years of operation of our recently installed 250 kV SSAMS at LLNL, purchased to replace our 1-MV AMS system for the measurement of 14 C from labeled biochemical samples. We have modified the ion source region to improve ion output. Additionally, the SSAMS required significant software modifications to the data acquisition system in order to accurately measure 14 C at the high-count rates typically encountered with labeled biochemical samples. Here, we found that the data can be corrected assuming a nonparalyzable dead time response with a single event dead time of 6 µs. Since operation began, we have measured over 13,000 graphitic unknowns and over 1900 standards with an overall precision of 1.0%. We have optimized our system for the analysis of CO 2 gas samples. We compared aliquots of identical samples measured as solid graphite and as liquid drops. Excellent agreement was found between the two, although the average precision of the graphite targets was an order of magnitude better than the liquid drop analysis due to the much larger number of 14 C atoms available for measurement.

42 ENGINEERING↗
DOE OSTI2023

Specifications of FIPD Fission Gas Release Data

All fission gas release data stored in the Fuels Irradiation & Physics Database (FIPD) was originally measured using the Gas Assay, Sample and Recharge (GASR) System in the Hot Fuel Examination Facility (HFEF). It is therefore called GASR data in FIPD. During the measurement of a sample, such as an irradiated EBR-II fuel element/capsule, a pinhole-sized region near the top of the element plenum was melted by a laser. Plenum gas then expanded into a calibrated volume (note: in this document, “sample” and “capsule/element” are used interchangeably consistent with GASR documents in FIPD). The pressure rise in the volume was recorded. Helium backfilling and expansion was then performed to determine the sample (e.g., fuel element plenum) volume using Boyle’s Law and assuming ideal gas behavior at constant temperature. With the plenum volume and the recorded pressure rise, the sample (e.g., fuel element plenum) pressure was derived with assumption of ideal gas law behavior. The plenum volume and pressure as well as the cladding temperature during the measurement were collected (GASR data in FIPD). Other records associated with the fission gas release data include: raw GASR data records including volumes and post-puncture pressures of seal head/sealing head and manifold, calibration data, backfilling gas pressure data, and the data analysis records. A sample(s) of the fission gas released from the plenum was collected by the GASR system into sample bottles. The chemical and isotopic composition of the gas sample was analyzed separately from GASR data, and will be discussed in a separate specification. The plenum volume, pressure, and cladding temperature during the measurement are typically utilized to determine the number of moles of gas in the plenum. This quantity is often compared to the number of moles of gas generated by fission events. However, calculating these values and their associated uncertainty is beyond the scope of this document, as it necessitates additional assumptions. The most important document to understand the FIPD fission gas data is the GASR operational manual (title: Gas Assay, Sample and Recharge System (GASR) operation and maintenance manual, HFEF/N OMM 4381, DOC. NO. W0018-0032-ES-00). This manual provides: (1) description of the GASR and the functions of each component (laser drilling, welding, seal head/sealing head, manifold, vacuum system, sample system, purge and gas tag system, etc.); (2) step-by-step guidance on calibrations, operations, and measurements; and (3) maintenance procedures and other details relating to the structure and operation of the GASR. Note that the original GASR operated until 2020. A new GASR with the same design and measurement methodology was installed in 2021. The specifications of the GASR presented on the HFEF website at this time are consistent with the ones given in the operational manual. The methods to calculate the plenum volume and pressures were not included in the operational manual, but were recorded in the legacy data analysis files. Details of the methods are given in Chapter 3.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

The Direct Path To WIPP - 12471

Sandia National Laboratories/New Mexico (SNL/NM), designated as a small quantity site (SQS) by the National TRU Program (NTP), generated contact-handled (CH) and remote-handled (RH) transuranic (TRU) waste primarily from the decontamination and clean-out of glove boxes at the Hot Cell Facility (HCF) at Technical Area (TA) V. All of the waste required repackaging, with the CH TRU waste being repackaged from late 2007 through 2011. Three shipments of CH were completed in October 2011, which de-inventoried SNL/NM's legacy TRU waste. In FY11, RH TRU waste was repackaged at the Auxiliary Hot Cell Facility (AHCF) located in TAV with the support of the Central Characterization Project (CCP). The waste was originally packaged in SNL/NM fabricated casks, cement or lead-lined 55-gallon drums, or 30-gallon drums. The AHCF is a small hot cell, with access only through a roof port which presented challenges for inserting and removing waste from the hot cell. The CCP provided visual examination operators (VEOs) to observe and document each waste item repackaged, removal of prohibited items, and radiological sampling. Dose-to-Curie measurements were calculated by CCP after a radiological report was prepared using scaling factors determined by the analysis of swipe samples. Finally, headspace gas samples were taken and sent to the Advanced Mixed Waste Treatment Project (AMWTP) for analysis. Despite the challenges, the RH waste is on track to be shipped to WIPP in early FY12. The processes used and procedures developed to conduct the repackaging operations, the issues identified and mitigated were challenging but the cooperation between SNL/NM and the Central Characterization Program (CCP) enabled SNL/NM to complete the repackaging and support the characterization and shipment. An inventory list, identification of the campaigns, discussion of the challenges and mitigations, and the final loading of the RH 72-B casks at TA-V for direct shipment to the Waste Isolation Pilot Plant (WIPP) will be discussed. Lessons learned from the RH campaigns are: - Some containers that were originally identified as HC-3 have been re-evaluated and became < HC-3 due to the conservative estimates made by the original generators - Operators at the AHCF were not accustomed to the detail required by the VE operators. However, they worked well together and the repackaging was completed ahead of schedule. - The AK was not always accurate as was demonstrated by the solid waste found in the drum during the first visit by EPA. That waste has since been determined to be low-level. - Two drums originally thought to be RH turned out to be CH and arrangement for RTR had to be made quickly. - Six of the original RH repacked drums became low level. - Lessons learned from the CH campaigns were helpful in avoiding many issues. The RH repackaging effort has been a success due to the expertise of the AHCF operators, supervisor, and manager, the conscientious attention to detail of the CCP VE operators, the experience of the CCP DTC and headspace gas sampling staff, and the guidance and support from CCP and CBFO. Sometimes schedules had to be adjusted, processes updated, and issues discussed, but the communication between CCP and SNL/NM was good. SNL/NM hopes to have the legacy RH TRU waste shipped off-site by early 2012. (authors)

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

Application of Indirect Quantification of 133mXe to Calibration of HPGe Detector

The INL Noble Gas Laboratory provides intercomparison samples for the noble gas analysis laboratories as part of the CTBTO PrepCom IMS. Xe-133m is one of the four relevant radionuclides in nuclear explosion monitoring. Without commercially available Xe-133m calibration standards laboratories must create and improve calibration methods. Improvements in calibration methods at the INL NGL benefit the CTBTO PrepCom through better certified values for Xe-133m intercomparison samples. Calibration of High Purity Germanium detectors for Xe-133m quantification is complicated by the coexistence of Xe-133 in samples under analysis. Xe-133 is typically produced in larger quantities, has higher gamma emission probabilities, and its gammas are detected more efficiently than Xe-133m. Xe-133m activity of samples can be indirectly inferred through the 133:133m activity ratio of a batch of material, and the Xe-133 counts in the assay of a small aliquot of the same material. This indirect quantification method can be leveraged to perform detector calibrations for quantification of Xe-133m. Calibrations can be performed by inferring the Xe-133m to certify the sample, and direct counting to determine detector efficiency. A comparison of method results will be shown.

133mXe↗
DOE OSTI2024

Pre-Transient Characterization of Historic EBR-II Pins for Transient Testing

Current interest in sodium-cooled fast reactor (SFR) designs, such as TerraPower’s Natrium Reactor, has highlighted the need for advanced reactor fuel technology development. Modern U-Zr and U- Pu-Zr pin designs are primary candidates to fuel SFRs and boast high fuel utilization capacity, increased fuel-cladding compatibility, and improved safety through inherent feedback mechanisms. Despite over 60 years of metallic fuel irradiation, uncertainties exist in the performance of the fuel system, particularly under transient overpower (TOP) and loss of flow (LOF) scenarios. Throughout historical testing within the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF), fuel behavior has demonstrated benign response to transient reactor conditions; however, accurate predictions of failure thresholds to inform operational limitations rely heavily on fuel composition, burnup, and irradiation history. In expanding TOP and LOF testing, the Transient Heat sink Overpower Response (THOR) Capsule will be used to test modern fuel technologies in a static sodium environment in the Transient Reactor Test (TREAT) Facility. The THOR capsule is highly instrumented and will provide time-dependent thermal behavior of SFR fuel pins subjected to accident conditions within TREAT. The THOR-Metallic (THOR- M) campaign aims to validate and expand historical TOP and LOF testing on high burnup U-Zr and U-Pu- Zr fuel alloys previously irradiated in EBR-II by running the rods to failure. This contribution focuses primarily on the pre-transient engineering-scale destructive and non- destructive characterization that has been conducted on both the test and sibling pins used for the TOP and LOF tests. All pins underwent visual examination, neutron radiography, element contact profilometry, and precise gamma scan. The sibling pins used for each test were further analyzed using gas assay, sampling, and recharge analysis (GASR), and optical microscopy. The results from each technique confirmed that the fuel pins were intact and devoid of any atypical developments when compared to historical data. Additionally, the analyzed measurements establish a baseline for comparison to post-transient analysis. Key fuel behaviors quanitifed include axial elongation of the fuel column, diametral strain of the pin, patterns in fluff structure geometry, changes in axial isotope distribution, evolution of constituent redistribution, porosity, and fission gas release. The pre-transient measurements and changes attributed to transient behavior from post-transient measurement will be compared to historical data to capture the behavioral dependence on composition, burnup, and irradiation history. Results from this work advance the initiatives of the THOR-M campaign, which aid in informing fuel performance models and establishing safety criteria for SFR operational limits. The novel combination of test environment, in-situ instrumentation, and comprehensive suite of characterization methods provides greater understanding of transient fuel behavior. Overall, information on the time and condition of pin failure for high burnup U-Pu-Zr will greatly expand the limited existing TOP and LOF test data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Pre-Transient Characterization of Historic EBR-II Pins for Transient Testing

Current interest in sodium-cooled fast reactor (SFR) designs, such as TerraPower’s Natrium Reactor, has highlighted the need for advanced reactor fuel technology development. Modern U-Zr and U- Pu-Zr pin designs are primary candidates to fuel SFRs and boast high fuel utilization capacity, increased fuel-cladding compatibility, and improved safety through inherent feedback mechanisms. Despite over 60 years of metallic fuel irradiation, uncertainties exist in the performance of the fuel system, particularly under transient overpower (TOP) and loss of flow (LOF) scenarios. Throughout historical testing within the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF), fuel behavior has demonstrated benign response to transient reactor conditions; however, accurate predictions of failure thresholds to inform operational limitations rely heavily on fuel composition, burnup, and irradiation history. In expanding TOP and LOF testing, the Transient Heat sink Overpower Response (THOR) Capsule will be used to test modern fuel technologies in a static sodium environment in the Transient Reactor Test (TREAT) Facility. The THOR capsule is highly instrumented and will provide time-dependent thermal behavior of SFR fuel pins subjected to accident conditions within TREAT. The THOR-Metallic (THOR- M) campaign aims to validate and expand historical TOP and LOF testing on high burnup U-Zr and U-Pu- Zr fuel alloys previously irradiated in EBR-II by running the rods to failure. This contribution focuses primarily on the pre-transient engineering-scale destructive and non- destructive characterization that has been conducted on both the test and sibling pins used for the TOP and LOF tests. All pins underwent visual examination, neutron radiography, element contact profilometry, and precise gamma scan. The sibling pins used for each test were further analyzed using gas assay, sampling, and recharge analysis (GASR), and optical microscopy. The results from each technique confirmed that the fuel pins were intact and devoid of any atypical developments when compared to historical data. Additionally, the analyzed measurements establish a baseline for comparison to post-transient analysis. Key fuel behaviors quanitifed include axial elongation of the fuel column, diametral strain of the pin, patterns in fluff structure geometry, changes in axial isotope distribution, evolution of constituent redistribution, porosity, and fission gas release. The pre-transient measurements and changes attributed to transient behavior from post-transient measurement will be compared to historical data to capture the behavioral dependence on composition, burnup, and irradiation history. Results from this work advance the initiatives of the THOR-M campaign, which aid in informing fuel performance models and establishing safety criteria for SFR operational limits. The novel combination of test environment, in-situ instrumentation, and comprehensive suite of characterization methods provides greater understanding of transient fuel behavior. Overall, information on the time and condition of pin failure for high burnup U-Pu-Zr will greatly expand the limited existing TOP and LOF test data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Specification of FIPD Fission Gas Chemistry Data

The current FIPD library contains two main sets of fission gas chemistry data. The first set is data collected during the Integral Fast Reactor (IFR) program from 1984 to 1994, using a gas mass spectrometry system located in the Analytical Laboratory (AL) at Argonne National Laboratory-West (Argonne-West). Throughout this period, numerous fission gas release and chemistry datasets were gathered from a variety of metallic fuel pins. The fission gas was sampled by the Gas Assay, Sample and Recharge (GASR) System in the Hot Fuel Examination Facility (HFEF) and transferred to the AL to perform gas composition and isotopic abundance analysis. The second set is data collected after the IFR program. The fission gas samples were also collected by the GASR system at the HFEF, but analyzed using a similar gas mass spectrometer located in Pacific Northwest National Laboratory (PNNL). Many fuel pins irradiated in Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF) were measured, including the fuel pins for the MFF series of experiments, designed to qualify metal fuel for use as driver fuel in the FFTF and X496 experiment. For either set of data, fission gas was sampled with the gas sampling line in GASR using sample bottles after the capsule/element volume has been determined and the system is still full of radioactive gas. The sample bottles were then transferred to the sample packaging cylinder or an approved storage location pending transfer to the AL or prepared for shipment to another laboratory (such as PNNL) for analysis of the collected gas as directed on the GASR data form, other approved form. The receiving laboratories (AL or PNNL) required their Analytical Service Request form to be completed prior to sample transfer. Typical sample transfer processes were initiated at HFEF by the principal or process engineer. The laboratories performing the analyses (AL or PNNL) use the sample bottle numbers as well as a sample number produced by the respective laboratory. HFEF and the responsible experimenter tracked the sample using the analysis number, the gas bottle number, and the fuel pin number.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Gamma Irradiation of NaCl-UCl 3 Salt for the Molten Chloride Fast Reactor

This report documents the results of the gamma ray irradiation of NaCl-UCl 3 eutectic salt in the Advanced Test Reactor (ATR) spent fuel pool gamma tube and subsequent analysis of capsules utilizing the Gas Assay Sample and Recharge (GASR) system. NaCl-UCl 3 salt capsules at four different temperatures, 75C, 150C, 300C and 600C, were irradiated shortly after the core internal changeout (CIC) of the ATR, beginning April 28 th of 2021, and ending on August 18 th of 2021. Re-positioning of the fuel elements in the spent fuel pool to move more fresh fuel elements around the gamma tube occurred on May 11 th of 2021, resulting in an approximate 20% increase in the dose rate and consequently the total absorbed dose. In total, the salt-filled capsules underwent 2638 hours of gamma irradiation from the adjacent freshly discharged ATR spent fuel. Capsule internal pressure measurements were taken via the Gas Assay Sample and Recharge (GASR) system in the INL Hot Fuels Examination Facility (HFEF) following completion of the irradiation. Based on the GASR analysis results, the primary conclusion from this experiment to-date is that radiolytic generation of chlorine gas is insignificant for the dose rates and total absorbed dose that these samples experienced. To elucidate more detailed effects of radiolysis on NaCl-UCl 3 eutectic salts and reach a more definitive conclusion, further advanced analyses such as electron microscopy, simultaneous thermal analysis, and electron paramagnetic resonance are recommended to analyze the capsule wall and salt samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Enabling liquid vapor analysis using synchrotron VUV single photon ionization mass spectrometry with a microfluidic interface

Vacuum ultraviolet (VUV) single photon ionization mass spectrometry (SPI-MS) is a vacuum-based technique typically used for analysis of gas phase and solid samples, but not for liquids, due to the challenge in proposing volatile liquids in vacuum. In this work we introduce the first demonstration of in situ liquid analysis by integrating the System for Analysis at the Liquid Vacuum Interface (SALVI) microfluidic reactor into VUV SPI-MS. Four representative volatile organic compound (VOC) solutions were used to illustrate feasibility of liquid analysis. Our results reflect accurate mass identification of the VOC molecules and reliable determination of appearance energy (AE) that is consistent with ionization energy (IE) for gaseous species in literature as reported. This work validates that the vacuum-compatible SALVI microfluidic interface can be utilized at the synchrotron beamline and enable in situ research of gas-phase molecules evaporating off the surface of a liquid, which holds importance in the study of condensed matter chemistry.

47 OTHER INSTRUMENTATION↗
DOE OSTI2025

Creating Accurate Methane Emission Inventories through Data-Driven Airborne Survey Strategies

Because natural gas emits less carbon than other fossil fuels, it holds promise as a green energy transition fuel. However, the overall carbon footprint of natural gas is significantly elevated by methane emissions that occur during its production and transmission (Cusworth et al. 2022). Methane “super-emitters,” while comprising only about 1% of sites, are responsible for the majority of oil- and gas-sourced methane emissions, making their detection and mitigation critical in reducing the climate impact of natural gas and in meeting national and global sustainability goals (Sherwin et al. 2024). Yet, despite advancements in detection, significant uncertainties remain regarding the size, frequency, and duration distributions of methane emissions (e.g., Frankenberg et al. 2016, Cusworth et al. 2022, Chen, Sherwin et al. 2022, Conrad et al. 2023, Johnson et al. 2023, Sherwin et al. 2024) underscoring the need for comprehensive emissions inventories segmented by basin across the US. Airborne surveys are well-suited for collecting data to build these comprehensive, basin-level inventories because they allow for extensive spatial coverage, and have the spatial resolution, and the sensitivity to pinpoint individual methane sources. As remote sensing technologies enable rapid basin-scale surveys, it is imperative to establish scientifically and statistically robust standards to generate reliable and actionable emissions inventories. Recent work has shown that differences in airborne sampling strategies, detection technologies, and analysis can lead to large differences between survey conclusions if not correctly accounted for (Chen et al. 2024). This elevates the importance of incorporating proper sampling and analysis techniques when designing a methane emissions monitoring campaign to produce accurate results and facilitate cross-study comparisons. In this paper, we describe a survey strategy designed using the latest conclusions from the literature to align results from different aerial surveys. We identify several sampling and analysis principles, including large sample sizes, balanced sampling across oil and gas production, careful survey area definition, and a unified protocol for analysis, to be vital to producing an unbiased estimate of basin-scale emissions. We present results from a Department of Energy-funded project that deployed this survey strategy in two understudied oil and gas- producing regions in the United States: the Haynesville Basin in Texas and Louisiana, and the Woodford Shale in the Anadarko Basin in Oklahoma.

03 NATURAL GAS↗
DOE OSTI2017

Direct Analysis of Organic Compounds in Liquid Using a Miniature Photoionization Ion Trap Mass Spectrometer with Pulsed Carrier-Gas Capillary Inlet

A miniature ion trap mass spectrometer with capillary direct sampling and vacuum ultraviolet photoionization source was developed to conduct trace analysis of organic compounds in liquids. Self-aspiration sampling is available where the samples are drawn into the vacuum chamber through a capillary with an extremely low flow rate (less than 1 μL/min), which minimizes sample consumption in each analysis to tens of micrograms. A pulsed gas-assisted inlet was designed and optimized to promote sample transmission in the tube and facilitate the cooling of ions, thereby improving instrument sensitivity. A limit of detection of 2 ppb could be achieved for 2,4-dimethylaniline in a methanol solution. The sampling system described in the present study is specifically suitable for a miniature photoionization ion trap mass spectrometer that can perform rapid and online analysis for liquid samples. .

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Investigation of the hydrolytic and radiolytic degradation of HEH[EHP]

The extractant 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) is a component used in both the Advanced TALSPEAK and ALSEP solvent extraction processes. The most likely compound formed via hydrolytic or radiolytic degradation of HEH[EHP] would be the phosphonic acid 2-ethylhexylphosphonic acid (H2EHP) that is formed by cleavage of the P-O-R bond. Thus far, attempts to detect H2EHP by gas chromatography or mass spectrometry have not been successful. The inability to detect this proposed degradation product in analytical samples is likely due to inadequate analysis techniques, lack of H2EHP production, further decomposition of H2EHP forming products not detectable by the employed analytical techniques, or a combination of all of the above scenarios. In order to address this problem, commercially available alkylphosphonic acids were acquired and used as surrogates for H2EHP in the gas chromatography and mass spectrometry analysis of samples. Once the ability to detect alkylphosphonic acid compounds was confirmed, these analytical techniques were used to confirm the production of H2EHP in samples of HEH[EHP] exposed to nitric acid and nitric acid plus gamma radiation. This report provides a brief summary of results and serves as documentation of the completion the level four milestone M4FT-16IN030102025 “Investigate the hydrolytic and radiolytic degradation of HEH[EHP]”.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗