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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Integration of CO2 Electrolysis with Microbial Syngas Upgrading to Rewire the Carbon Economy

This project has worked towards demonstrating and understanding the conversion of waste CO2 into value added fuels and chemicals. If successful, our proposed technology will incentivize CO2 capture and can add carbon efficiency to biorefineries. We have integrated two conversion technologies to demonstrate a novel approach of combining electrolytic CO2 reduction with biocatalytic CO upgrading. Industrial partners 3M and Dioxide Materials have recently demonstrated electrolytic conversion of CO2 and water to syngas using inexpensive renewable energy. Concurrently, supporting partner Lanzatech has demonstrated industrial scale CO fermentation using steel mill waste CO as a substrate. This presentation will highlight the investment BETO has made to establish a CO2 electrolysis test station and CO fermentation at NREL. We will also discuss the process of increasing the size of the electrolyzers to meet fermentation needs, along with the investigation into the affect CO2 concentration and contaminates have on electrolyzer efficiency, lifetime, and specificity. Additionally, we have begun to examine the downstream impact of electrolyzer off-gas CO:H2 ratios have on the metabolism of C. autoethanologen. Experimental results along input from our industrial partners have led to a baseline techno-economic and life cycle analyses. These results enabled the identification of key cost drivers and carbon intensity of the process. We have used these metrics to identify technology gaps to iteratively inform the process optimization and potential deployment siting to achieve economically viable, sustainable conversion of biopower-derived flue gases to fuels and chemical intermediates.

bioenergy↗
DOE OSTI2014

Laboratory Evaporation Testing Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream, LAW Off-Gas Condensate, from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of canistered glass waste forms. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to be within acceptable concentration ranges in the LAW glass. Diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task examines the impact of potential future disposition of this stream in the Hanford tank farms, and investigates auxiliary evaporation to enable another disposition path. Unless an auxiliary evaporator is used, returning the stream to the tank farms would require evaporation in the 242-A evaporator. This stream is expected to be unusual because it will be very high in corrosive species that are volatile in the melter (chloride, fluoride, sulfur), will have high ammonia, and will contain carryover particulates of glass-former chemicals. These species have potential to cause corrosion of tanks and equipment, precipitation of solids, release of ammonia gas vapors, and scale in the tank farm evaporator. Routing this stream to the tank farms does not permanently divert it from recycling into the WTP, only temporarily stores it prior to reprocessing. Testing is normally performed to demonstrate acceptable conditions and limits for these compounds in wastes sent to the tank farms. The primary parameter of this phase of the test program was measuring the formation of solids during evaporation in order to assess the compatibility of the stream with the evaporator and transfer and storage equipment. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW facility melter offgas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet, and, thus, the composition will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. This report discusses results of evaporation testing of the simulant. Two conditions were tested, one with the simulant at near neutral pH, and a second at alkaline pH. The neutral pH test is comparable to the conditions in the Hanford Effluent Treatment Facility (ETF) evaporator, although that evaporator operates at near atmospheric pressure and tests were done under vacuum. For the alkaline test, the target pH was based on the tank farm corrosion control program requirements, and the test protocol and equipment was comparable to that used for routine evaluation of feed compatibility studies for the 242-A evaporator. One of the radionuclides that is volatile in the melter and expected to be in high concentration in this LAW Off-Gas Condensate stream is Technetium-99 (99Tc). Technetium will not be removed from the aqueous waste in the Hanford WTP, and will primarily end up immobilized in the LAW glass by repeated recycle of the off-gas condensate into the LAW melter. Other radionuclides that are also expected to be in appreciable concentrations in the LAW Off-Gas Condensate are 129I, 90Sr, 137Cs, and 241Am. The concentrations of these radionuclides in this stream will be much lower than in the LAW, but they will still be higher than limits for some of the other disposition pathways currently available. At this time, these scoping tests did not evaluate the partitioning of the radionuclides to the evaporator condensate, since ample data are available separately from other experience in the DOE complex. Results from the evaporation testing show that the neutral SBS simulant first forms turbidity at ~7.5X concentration, while the alkaline-adjusted simulant became turbid at ~3X concentration. The major solid in both cases was Kogarkoite, Na3FSO4. Sodium and lithium fluorides were also detected. Minimal solids were formed in the evaporator bottoms until a substantial fraction of liquid was removed, indicating that evaporation could minimize storage volume issues. Achievable concentration factors without significant insoluble solids were 17X at alkaline pH, and 23X at neutral pH. In both runs, significant ammonia carried over and was captured in the condenser with the water condensate. Results also indicate that with low insoluble solids formation in the initial testing at neutral pH, the use of Reverse Osmosis is a potential alternate method for concentrating the solution, although an evaluation is needed to identify equipment that can tolerate insoluble solids. Most of the ammonia remains in the evaporator bottoms during the neutral pH evaporation, but partitions to the condensate during alkaline evaporation. Disposition of both streams needs to consider the management of ammonia vapor and its release. Since this is an initial phase of testing, additional tasks related to evaporation methods are expected to be identified for development. These tasks likely include evaluation and testing of composition variability testing and evaluations, corrosion and erosion testing, slurry storage and immobilization investigations, and evaporator condensate disposition.

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

Xe Concentration Activity

INL conducted extensive studies to understand how the Xe capacity of AgZ-PAN changes under various conditions including flowrate, temperature, concentration, carrier gas, and irradiation of the sorbent. Therefore, it is well understood that AgZ-PAN can capture Xe under a wide range of conditions. While these studies were performed to prove that Xe can be completely desorbed from the column, the ability to concentrate Xe during desorption was not studied. Concentrating Xe during desorption would be ideal so as to produce Xe that can be used for commercial and research applications. This report investigates the Xe concentrating potential of AgZ-PAN by running multiple Xe adsorption-desorption cycles, collecting the desorbed effluent from a saturated AgZ-PAN column, and then loading it onto a fresh column.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Alternative Materials Literature Review: A Review of Alternative Materials for Xenon Capture in Comparison to AgZ-PAN

A collaborative literature review between Idaho National Laboratory (INL) and Pacific National Laboratory (PNNL) was conducted to research alternative metals/materials for xenon adsorption in gaseous effluents generated from used nuclear fuel (UNF) reprocessing. The review was conducted to research sorbents that could outperform silver mordenite (AgZ) at ambient temperature. The use of silver requires complex and expensive waste disposal processes due to it being an RCRA metal, therefore alternative materials need to be researched and evaluated. The report details two different types of sorbent material: alternative zeolite materials and non-zeolite-based materials (i.e., MOFs, COFs, etc.). This report documents a brief history of the materials, what testing has been conducted on them, and what results suggest the possibility of outperforming the silver-based materials. Sorbent performance characteristics such as durability, capacity, and operating temperatures were evaluated. The simplicity of the sorbents was also evaluated, meaning how they are acquired (synthesized vs. naturally abundant). Recommendations were made for the most promising sorbents based on available research in the subject area and these recommended sorbents will be tested in the upcoming fiscal year. The recommended sorbents that are zeolite based are CHA zeolites, ZIF-69, and Na-ETS-10. The recommended sorbents that are non-zeolite based are SBMOF-1 and CC3.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Review of molten salt reactor off-gas management considerations

Molten salt reactors offer a wide range of potential benefits but pose some unique challenges, particularly for designs that use an unclad liquid salt fuel. This type of fuel will result in the transport of fission gases into the headspace of the reactor where in some designs a cover gas can be circulated to remove certain fission products and maintain an inert atmosphere. The cover gas leaving the reactor core is expected to contain both noble and non-noble gases, aerosols, volatile species, tritium, radionuclides, and their daughters. To remove these radioactive gases, it is necessary to develop a robust off-gas system for molten salt–fueled reactors. Various treatment systems must be staged in series to remove the off-gas constituents from the stream before recirculating the gas back to the headspace of the reactor. Treatment options vary greatly depending on what they are designed to remove from the gas stream. This paper reviews the anticipated composition of a typical molten salt reactor off-gas stream and subsequently the available resources that could be employed to remove these species from the gas stream. An example off-gas system is then detailed, along with important design considerations, exemplifying the necessity for high-fidelity modeling. Lastly, the need for further thermophysical property research and the employment of advanced sensor technology for treatment component testing are discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Final Report - Advanced Laser-Based Sensors for Industrial Process Control

The objective of this work is to capture the potential of real-time monitoring and overcome the challenges of harsh industrial environments, Los Gatos Research (LGR) is fabricating, deploying, and commercializing advanced laser-based gas sensors for process control monitoring in industrial furnaces (e.g. electric arc furnaces). These sensors can achieve improvements in process control, leading to enhanced productivity, improved product quality, and reduced energy consumption and emissions. The first sensor will utilize both mid-infrared and near-infrared lasers to make rapid in-situ measurements of industrial gases and associated temperatures in the furnace off-gas. The second sensor will make extractive measurements of process gases. During the course of this DOE project, Los Gatos Research (LGR) fabricated, tested, and deployed both in-situ tunable diode laser absorption spectrometry (TDLAS) analyzers and extractive Off-Axis Integrated Cavity Output Spectroscopy (Off-Axis ICOS) analyzers.

47 OTHER INSTRUMENTATION↗
DOE OSTI2014

Iodine Loading of NO Aged Silver Exchanged Mordenite

In an off-gas treatment system for used nuclear fuel processing, a solid sorbent will typically be exposed to a gas stream for months at a time. This gas stream may be at elevated temperature and could contain water vapor, gaseous nitrogen oxides (NO{sub x}), nitric acid vapors, and a variety of other constituents. For this reason, it is important to evaluate the effects of long-term exposure, or aging, on proposed sorbents. Silver exchanged mordenite (AgZ) is being studied at Oak Ridge National Laboratory (ORNL) to determine its iodine sorption capacity after long term exposure to increasingly more complex chemical environments. Studies previously conducted at ORNL investigated the effects of aging reduced silver exchanged mordenite (Ag{sup 0}Z) in dry air, moist air, and NO2. This study investigated the effects of extended exposure to nitric oxide (NO) gas on the iodine capture performance of Ag{sup 0}Z. A deep bed of Ag{sup 0}Z was aged in a 1% nitric oxide (NO) air stream, and portions of the bed were removed at pre-determined intervals. After being removed from the NO stream, each sample was loaded with iodine in a thin bed configuration. These samples were analyzed by neutron activation analysis (NAA) to quantify the iodine content in the sample. Samples were removed at one week and one month. A 78% decrease in sample capacity was seen after one week of exposure, with no further decrease observed after 1 month of aging. The observed loss in capacity is larger in magnitude than previous studies exposing Ag{sup 0}Z to dry air, moist air, or NO2 gas. The aging study was terminated after one month and repeated; this successfully demonstrated the reproducibility of the results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Adsorption Isotherms for Xenon and Krypton using INL HZ-PAN and AgZ-PAN Sorbents

The generation of adsorption isotherms compliments the scale-up of off-gas processes used to control the emission of encapsulated radioactive volatile fission and activation products released during Used Nuclear Fuel (UNF) reprocessing activities. A series of experiments were conducted to obtain capacity results for varying Kr and Xe gas concentrations using HZ-PAN and AgZ-PAN engineered form sorbents. Gas compositions for Kr ranged from 150-40,000 ppmv and 250-5020 ppmv for Xe in a helium balance. The experiments were all performed at 220 K at a flowrate of 50 sccm. Acquired capacities were then respectively fit to the Langmuir equation using the Langmuir linear regression method to obtain the equilibrium parameters Qmax and Keq. Generated experimental adsorption isotherms were then plotted with the Langmuir predicted isotherms to illustrate agreement between the two. The Langmuir parameters were provided for input into the OSPREY model to predict breakthrough of single component adsorption of Kr and Xe on HZ-PAN and AgZ-PAN sorbents at the experimental conditions tested. Kr and Xe capacities resulting from model breakthrough predictions were then compared to experimental capacities for model validation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Xe/Kr Selectivity Measurements using AgZ-PAN at Various Temperatures

In preparation for planned FY-15 Xe/Kr multi-column testing, a series of experiments were performed to determine the selectivity of Xe over Kr using the silver converted mordenite-polyacrylonitrile (AgZ-PAN) sorbent. Results from these experiments will be used for parameter selection guidelines to define test conditions for Kr gas capture purity evaluations later this year. The currently configured experimental test bed was modified by installing a new cooling apparatus to permit future multi-column testing with independent column temperature control. The modified test bed will allow for multi-column testing to facilitate a Xe separation followed by a Kr separation using engineered form sorbents. Selectivity experiments were run at temperatures of 295, 250 and 220 K. Two feed gas compositions of 1000 ppmv Xe, 150 ppmv Kr in either a He or an air balance were used. AgZ-PAN sorbent selectivity was calculated using Xe and Kr capacity determinations. AgZ-PAN sorbent selectivities for Xe over Kr of 72 were calculated at room temperature (295 K) using the feed gas with a He balance and 34 using the feed gas with an air balance. As the test temperatures were decreased the selectivity of Xe over Kr also decreased due to an increase in both Xe and Kr capacities. At 220 K, the sorbent selectivities for Xe over Kr were 22 using the feed gas with a He balance and 28 using the feed gas with an air balance. The selectivity results indicate that AgZ-PAN used in the first column of a multi-column configuration will provide adequate partitioning of Xe from Kr in the tested temperature range to produce a more pure Kr end product for collection.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using nonradioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

This report is issued as the first revision to FCRD-MRWFD-2016-000297. Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using non-radioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Stability of Tritium and Iodine Sorbents in TPOG Conditions

US regulations could require the removal of both iodine and tritium from the off-gas stream of a used nuclear fuel (UNF) reprocessing facility. Advanced tritium pretreatment is a pretreatment step that uses high concentrations of NOR2R in a gas stream to volatilize tritium and iodine from UNF prior to traditional dissolution. The gaseous effluent from this process would then require abatement to remove tritium and iodine, but high levels of NOR2R could have a detrimental effect on the ability of various solid sorbents to remove the volatile radionuclides. For tritium and iodine, the sorbents of interest are 3Å molecular sieve (3AMS) for tritium and reduced silver mordenite (AgP 0 PZ), silver-functionalized silica-aerogel (AgAerogel), and silver-nitrate-impregnated alumina (AgA) for iodine. Prior research has demonstrated that exposure to high concentrations of NOR2R can reduce the iodine loading capacity of AgP 0 PZ by > 90% when exposed for 1 week. Research in Japan has demonstrated that AgA is more robust to NOR2R exposure than AgZ. The testing described here was intended to assess the effects of high concentrations of NOR2R on the iodine capture capacity of AgA and the water adsorption capacity of 3AMS. To determine the effect of extended exposure of the sorbents to NOR2R, both 3AMS and AgA were aged in a 75% NOR2R environment prior to loading. The 3AMS samples were aged for 1, 4, and 5.5 weeks at 40°C. They were then loaded with water in a 10°C dew point stream (corresponding to a water concentration of ~12,000 ppmv) at 40°C. There was no significant change in the water adsorption capacity of the 3AMS upon exposure to 75% NOR2R. The AgA samples were aged for 1, 2, and 4 weeks at 150°C and were loaded with 50 ppmv IR2R at 150°C. The results show that the iodine capture capacity of AgA is reduced by exposure to high concentrations of NOR2R. The iodine capacity reductions were 16%, 36%, and 76% for 1, 2, and 4 week exposures, respectively. This is less of a capacity loss than that seen in similar testing with the AgP 0 PZ sorbent.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Comparison of Extended and Accelerated VOG Tests

The reprocessing of used nuclear fuel would release volatile radionuclides into the off-gas streams of a processing plant, including ¹²⁹I. The dissolver off-gas and more dilute vessel off-gas streams (VOG) must both be targeted when mitigating ¹²⁹I environmental release because each contains some amount of iodine. Iodine-129 will likely be found as elemental iodine (I₂) and methyl iodide (CH₃I) in these off-gas streams. Reduced silver-exchanged mordenite (AgZ or Ag⁰Z) has been investigated as a potential sorbent for iodine abatement and is studied here under prototypic VOG conditions. Because of the relatively low iodine abundance and high flow rates of the VOG, total iodine concentrations are expected to be in the parts per billion range. Thus, VOG experiments need to run for extended durations at low concentrations to reach sorbent saturation. Because the sorbent will be exposed to oxidizing gas streams for extended periods of time, aging effects and sorbent degradation need to be considered when designing a sorbent-based abatement system. Three sets of experiments were performed with the aim of determining how the adsorption of iodine by AgZ is affected by differing test durations and gas compositions. The first tested the capacity of sorbent aged for 8 months under a humid air stream. The second tested differences in sorbent capacity and mass transfer zone (MTZ) length during high-concentration (1200 ppb v ) CH₃I loading in a humid nitrogen gas (N₂) stream and a humid air gas stream over 28 days. The third tested sorbent capacity and MTZ length during low concentration (< 200 ppb v ) CH₃I and I₂ loading in a humid air stream over 9 months. The results of these tests suggest that AgZ capacity drops by ~50% after 1 month of aging, by ~60% after 2 months of aging, and then does not continue to decrease significantly at up to 8 months of aging. One-month aging tests that compared N₂ and air as the gas stream diluent resulted in similar maximum loading capacities and overall loading curves, indicating that the effects of aging cannot be mitigated by avoiding air as the balance gas. The 9-month extended VOG tests did not result in clear sorbent saturation at the inlet, but it can be inferred using an assumption of a maximum capacity of 45 mg I/g AgZ. Applying this assumption, then the MTZ of CH₃I is 12.8 cm, and the MTZ of I₂ is 12.4 cm. These results are similar to previous tests. Comparisons to tests completed at Idaho National Laboratory suggest that the CH₃I MTZ may be dependent on concentration. Future work should re-evaluate the design of the VOG iodine capture system with this updated data and should seek to understand fundamental characteristics of CH₃I adsorption by AgZ, such as the effect of concentration, gas velocity, and gas composition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Development of Integrated Thermal Fluids Modeling Capability for MSRs

The DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program supports a full range of computational thermal fluids analysis capabilities and code developments for a broad class of light-water and advanced reactor concepts. The research and development approach under thermal fluids technical area synergistically combines three length and time scales in a hierarchal multi-scale approach. To demonstrate the feasibility and capabilities of a multi-scale thermal fluids capability using these codes, a key joint effort has been pursued to develop an integrated system-and engineering-scale thermal fluids analysis capability with the MOOSE-based codes, through integration of SAM and Pronghorn, both based on the MOOSE framework. This report summarizes the progress in developing an integrated system- and engineering-scale thermal fluids analysis capability based on SAM and Pronghorn for molten salt reactors (MSRs), which gained significant interest in recently years. Two coupling approaches were studied, i.e. separate domain or domain-segregated coupling approach and the domain-overlapping approach. A series of coupled multi-physics models have been developed for a common reference molten salt fast reactor (MSFR) concept, ranging from standalone SAM system model to integrated SAM-Pronghorn-Griffin models. Both the steady state and the transient simulations are performed to the state-of-the-art simulation capabilities of NEAMS software suite in MSFR system applications. This report also covers further development and testing of the gas transport model in SAM for MSR modeling support. The presence of noncondensable gas in MSR systems would have strong impacts on fission gas removal and transport of noble metals throughout the system. Fission products removed through the off-gas system can also impact reactivity and can act as an additional point of heat removal. To ensure that the gas transport model is adequately tested, the supported modeling capabilities and features of SAM were identified, and a suite of tests were developed to test the model for each identified feature. Validation and UQ testing were also performed for the model and demonstrated that the buoyancy term, which was originally developed using non-salt/helium gas experimental data, can capture the experimental gas bubble velocity and diameter to within experimental and code uncertainty. An existing MSRE model was also modified in this work to include the gas transport model and to demonstrate the gas model behavior for realistic conditions of interest.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Partitioning of Radionuclides in Various Streams of the Mo-99 Purification Process

The Universal Neutron Irradiator (UNI) was designed to irradiate a small volume of uranium solution and to achieve high fission power without creating a large radionuclide (RN) inventory. The UNI system is equipped with a fully enclosed off-gas system with online hydrogen gas monitoring to prevent the production of an explosive atmosphere, as well as storage cylinders for capturing fission gases. It also includes a catalytic recombiner in the target capsule (to immediately recombine radiolytic hydrogen and oxygen gas in the system) and access ports for sampling the target solution and capsule headspace. In UNI Phase I experiments, 50 mL of uranyl sulfate solution was irradiated at 0.3–0.5 W/mL fission power using a tantalum target assembly to produce photoneutrons via irradiation with an electron beam. The purposes of the Phase I experiments were to test the efficacy of adding 200 ppm Fe 2+ to the target solution to prevent uranyl peroxide precipitation under representative irradiation conditions, and to monitor iodine speciation in the irradiated uranyl sulfate solution. The focus of the UNI Phase II experiments was to use an irradiated uranyl sulfate solution containing a mixture of fission products and monitor their distribution in various streams during the primary recovery column (titania-based column) that separates Mo-99 from uranium. Irradiations were designed to create sufficient activity for tracking their distribution in various processing streams using a combination of gamma counting and inductively coupled plasma mass spectrometry (ICP-MS) analyses. Obtained data will be used to better understand the composition of waste streams, accumulation of RN on the titania column, and identify what RNs accumulate in the uranyl sulfate solution. Furthermore, this knowledge will allow us to determine waste classifications for various streams produced during the Mo-99 purification process and develop an appropriate waste clean-up strategy.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Bubble Mass Transport Measurement in the Large-Scale Test Loops at Oak Ridge National Laboratory

Molten salts are complex fluids that incorporate multi-phase behavior depending on the chemistry and the physical properties of the entrained components. These components include the carrier salt, the actinide fuel, and fission, activation, and corrosion products, the concentrations of which depends on the burnup history of the salt. Radionuclide transport from the salt into the cover gas / off-gas system depends on volatility as predicted by thermochemistry, but data from the Molten Salt Reactor Experiment (MSRE) conducted in the late 1960s suggest that bubble formation and transport are also important. Anomalously high fractions of noble metals were found in the off-gas system and were attributed to transportation with parent salt aerosols and their association with rising noble gas bubbles. The prediction of such phenomena requires coupled neutronic, thermal hydraulic, and chemical equilibrium calculations, the framework of which is being developed within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program supported by the US Department of Energy (DOE) Office of Nuclear Energy (NE). However, separate effects tests and models of experiments using tools such as SAM and Thermochimica can guide model developers through important processes. Conversely, model development guides the choice of experiments and systems to provide data that are relevant for validation. This report describes several experiments that tracked gas transport in molten salts, ranging from small-scale systems to large-scale loops. Gas transportation in a molten salt, LiCl-KCl, has been studied using the shadowgraph technique. Sensors such as residual gas analysis, Raman spectroscopy, and laser-induced breakdown spectroscopy have been tested for off-gas measurements. These data were used to interpret how the gases move through the upstream salt / cover gas and the interface between them. Differential pressure measurements were able to detect individual gas bubbles as they popped at the liquid–gas interface. Salt aerosols were collected on a cascade impactor. Their formation was also observed directly via the shadowgraph method, and most of these aerosols were launched ballistically into the cover gas. Fine mists could also be observed. Convection currents through the salt were visualized and can be used to calculate the thermophysical properties of the salt itself. The apparatuses described in this report and in a previous work (McFarlane et al. 2023) have been commissioned and are available for use in making further measurements of salt/surrogate fission product behavior. Novel approaches using neutron imaging are planned for the study of fluoride salts, which cannot be contained in quartz, so shadowgraph visualization is not available. Bubble transport in convective flow and in a slow-moving salt column are planned. The mobile laser-induced breakdown spectroscopy (LIBS) system is available for several applications, including iodine capture in a molten hydroxide scrubber, H2 transport though molten salts to complement Raman analysis, and online tracking of salt aerosol generation, transport, and deposition. This report summarizes the findings from FY24 and the plans for FY25. The work completes milestone M2AT-24OR0702013 of the DOE-NE Advanced Reactor Technology, Molten Salt Reactor Campaign, DOE-NE-5.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Separation of Xenon from Krypton in UNF Reprocessing Off-Gas

Reprocessing of used nuclear fuel (UNF) results emits several volatile radionuclides that must be removed from the bulk off gas prior to release in order to meet U.S. regulatory emission requirements. After other volatile radionuclides have been removed, krypton (Kr) remains, along with associated with xenon (Xe) in the gas stream. The half-life of Xe is short enough that by the time UNF is sufficiently cooled to be reprocessed Xe is no longer a radiological concern and it may be release to atmosphere or collected for other purposes. Kr, however, must be placed in long term storage. In order to ensure the smallest practical volume for long term storage, the Xe, which is nearly ten times the volume of Kr, must be separated from the Kr. Although cryogenic distillation has been used successfully for this purpose, high capital and operational cost, along with safety considerations, made development of gas adsorption processes for separation of Xe and Kr attractive. INL developed an engineered form sorbent consisting of silver mordenite in a polyacrylonitrile matrix (AgZ-PAN) capable of complete separation of Xe from Kr in a single pass, allowing Kr to be captured downstream with no detectable Xe present.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

MRWFD Kr Capture Project

Progress report presentation to DOE program representatives.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗