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DOE OSTI2020

An Analysis of Abandoned Oil Well Characteristics Affecting Methane Emissions Estimates in the Cherokee Platform in Eastern Oklahoma

Anthropogenic activities increase methane emissions, contributing to greenhouse gas levels and adversely affecting the environment. Abandoned oil and gas wells potentially leak methane, but data are limited. We analyze methane emissions from abandoned wells (n = 179) in the Cherokee Platform in Oklahoma, a previously unaccounted basin, and compare emissions factors (EFs) to those in the Greenhouse Gas Inventory. We compare the contribution of various characteristics to the propensity for leakage. Higher emissions were observed with shallower wells and with unplugged wells. Plugged wells (n = 20) had an average EF of 96 ± 429 g/day and 65 ± 294 g/day for unplugged wells (n = 159). The majority of wells had no detectable leak. We calculated ethane EFs based on geochemical analysis of gas samples, finding higher EFs for unplugged (1.2 ± 5.5 g/day) versus plugged (0.9 ± 4.6 g/day) wells. The data indicate that in addition to the location of abandoned wells, physical characteristics are necessary to consider in estimating methane emissions.

58 GEOSCIENCES↗

DOE OSTI2018

Off-Gas Condensate and Cast Stone Analysis Results

A sample of off-gas condensate produced during a demonstration test of the direct-feed low-activity waste (DFLAW) Radioactive Waste Test Platform using waste retrieved from Hanford storage tank AP-105 waste was analyzed for major cations, anions, total cyanide, Resource Conservation and Recovery Act (RCRA) metals, selected radionuclides, and selected organic compounds. The remaining off-gas condensate was concentrated through evaporation and then used to produce Cast Stone waste forms. After the solid waste forms cured for 28 days, U.S. Environmental Protection Agency (EPA) Method 1311 (EPA 1992) Toxicity Characteristic Leach Procedure (TCLP) tests were conducted on two samples. The TCLP leachates were analyzed for RCRA metals, fluoride, and total cyanide. These chemical analyses were conducted to collect data on the waste form performance and to determine if the waste form will meet the disposal requirements of the Waste Control Specialists LLC (WCS) Federal Waste Disposal Facility (FWF) in Texas (WCS 2015) and the Integrated Disposal Facility at Hanford. The radionuclide results in the off-gas condensate were used to estimate concentrations in the Cast Stone waste forms; these estimates were then compared to Class C limits (30 TAC 336.362 Appendix E). With the possible exception of U-232, all measured radionuclides were determined to be below their respective Class C limits. Inductively coupled plasma-mass spectrometry (ICP-MS) was used to measure most of the radionuclides. It is not possible to differentiate the isotope measured at mass 232 between U 232 and Th-232 by mass spectroscopy. Class C limit comparisons were made by assuming two maximum possible scenarios; one in which all the isotope at mass 232 occurred as Th-232, and another assigning the entire mass contribution to U-232. If the isotope at mass 232 is all Th-232 then the waste form is well below its Class C limit; however, if the isotope at mass 232 is all U-232, then the Class C limit would be exceeded by more than a factor of 8. A chemical procedure to separate uranium from thorium followed by reanalysis by ICP-MS would be required to resolve this issue. A similar issue occurs at mass 238, which could be U-238 or Pu-238; however, in this case Pu-238 can be ruled out as a contributor to the isotope measured at mass 238 for the following reasons. Pu-238 has a half-life of 87.7 years and its daughter product (U-234) has a half-life of 2.5 × 105 years and a measured concentration of < 0.002 µg/L. If significant concentrations of Pu-238 were present in this waste, then the amount of measurable U 234 would have increased over the decades of storage in the Hanford tanks. Because this was not observed, it is concluded that the isotope measured at mass 238 is all U-238. A total of 68 organic compounds were targeted for analysis in the off-gas condensate. Of these compounds, only nine were measured above their respective minimum reportable concentration. These compounds were acenaphthene, acetone, n-butyl alcohol, p,p’-DDT, methyl ethyl ketone, nitrobenzene, o nitrophenol, pyridine, and toluene. Of the detectable compounds, only acenaphthene, acetone, o nitrophenol, and pyridine were above wastewater standards (40 CFR 268.48 - Universal Treatment Standards). None of the organic compounds exceeded the non-wastewater standards (40 CFR 268.48 - Universal Treatment Standards). Analysis of the TCLP leachates from the Cast Stone waste forms indicated that none of the analytes exceeded the Universal Treatment Standard and, in general, all were well below the standards.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Conversion kinetics of container glass batch melting

Abstract Understanding the batch‐to‐glass conversion process is fundamental to optimizing the performance of glass‐melting furnaces and ensuring that furnace modeling can correctly predict the observed outcome when batch materials or furnace conditions change. To investigate the kinetics of silica dissolution, gas evolution, and primary foam formation and collapse, we performed X‐ray diffraction, thermal gravimetry, feed expansion tests, and evolved gas analysis of batch samples heated at several constant heating rates. We found that gas evolving reactions, foaming, and silica dissolution depend on the thermal history of the batch in a similar manner: the kinetic parameters of each process were linear functions of the square root of the heating rate. This kinetic similarity reflects the stronger‐than‐expected interdependence of these processes. On the basis of our results, we suggest that changes in furnace operating conditions, such as firing or boosting, influence the melting rate less than what one would expect without consideration of batch conversion kinetics.

Ueda, Nanako↗

DOE OSTI2021

G-LiHT Campaign Leaf Sample details & photos, March 2017: Puerto Rico

This data package includes details of leaves sampled for leaf spectra and chemistry from 5 sites in Puerto Rico, in March of 2017. Sunlit canopy and shaded leaves of 66 species were collected. Data for each sample includes species, leaf age, type of analysis (spectroscopy, gas exchange, chemistry), sample number and sample photographs. The data package includes a spreadsheet with sample information and a zip file of photographs (1.6 GB). This data was collected as part of the 2017 BNL–G-LiHT leaf spectra campaign. See related datasets for leaf spectral reflectance and transmittance, leaf mass area (LMA), and leaf chemistry. Note that leaf sample details are also included in related datasets.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2024

Creating Accurate Methane Emission Inventories through Data-Driven Airborne Survey Strategies: Methods and Results from the Haynesville, Anadarko, and Permian Basins

Significantly reducing methane emissions from the oil and gas sector can decrease the rate of climate change over the next two decades, buying critical time for a global energy transition. However, emissions inventories that can be used by oil and gas operators and environmental regulators to identify optimal methane emission mitigation strategies are either based on conservative emission factor methods, or are inconsistent between studies due to differences in sampling strategies or survey technologies. We developed a new approach for methane emissions survey design that yields representative basinwide methane emissions inventories by surveying a subset of total assets in a given oil and gas basin. 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 that can be reconciled with future studies. We further present results from deploying this strategy in two 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. Aerial surveys were performed in 2023 using the Insight M LeakSurveyor™ technology. Preliminary results from methane emissions detected by Insight M indicate that aerially detected emissions above roughly 30 kg(CH4)/hr by themselves contribute a fractional loss rate of 1.13% of gross gas production across oil and gas operations in the Haynesville Basin, with aerially detected emissions equivalent to 2.67% of gross gas production in the Woodford Shale. We supplement these aerial estimates with modeled emissions that are below the LeakSurveyor’s survey sensitivity using a recently published inventory-based model of methane emissions, which we update for our survey areas. We then combine our aerial detections with modeled emissions to yield methane emission distributions and inventories that incorporate the full range of potential methane emissions from the smallest to the largest. These results can be used to identify the most effective methane mitigation strategies for our study areas, and can be reconciled with future methane emissions surveys that use different technologies.

Sherwin, Evan (ORCID:0000000321804297)↗

DOE OSTI2016

Technical Proposal for Loading 3000 Gallon Crude Oil Samples from Field Terminal to Sandia Pressurized Tanker to Support US DOE/DOT Crude Oil Characterization Research Study

Sandia National Laboratories is seeking access to crude oil samples for a research project evaluating crude oil combustion properties in large-scale tests at Sandia National Laboratories in Albuquerque, NM. Samples must be collected from a source location and transported to Albuquerque in a tanker that complies with all applicable regulations for transportation of crude oil over public roadways. Moreover, the samples must not gain or lose any components, to include dissolved gases, from the point of loading through the time of combustion at the Sandia testing facility. In order to achieve this, Sandia designed and is currently procuring a custom tanker that utilizes water displacement in order to achieve these performance requirements. The water displacement procedure is modeled after the GPA 2174 standard “Obtaining Liquid Hydrocarbons Samples for Analysis by Gas Chromatography” (GPA 2014) that is used routinely by crude oil analytical laboratories for capturing and testing condensates and “live” crude oils, though it is practiced at the liter scale in most applications. The Sandia testing requires 3,000 gallons of crude. As such, the water displacement method will be upscaled and implemented in a custom tanker. This report describes the loading process for acquiring a ~3,000 gallon crude oil sample from commercial process piping containing single phase liquid crude oil at nominally 50-100 psig. This document contains a general description of the process (Section 2), detailed loading procedure (Section 3) and associated oil testing protocols (Section 4).

02 PETROLEUM↗

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 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 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

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↗