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DOE OSTI2019

Installation and Commissioning of the Netzsch Skimmer

The DOE molten salt reactor (MSR) program has been focusing on the thermochemistry and transport properties of fuel, diluent/carrier, coolant, flush, and reprocessing salts proposed for various molten salt reactor technologies. Data from the chemistry project inform models of salt behavior used in dynamic models of nuclear salt reactors, for both high fidelity and systems simulations. Reactor and salt processing systems operate well above salt melting points (>600°C), so salt behavior can be well described by thermodynamics. However, much is not known about salt systems that contain two, three, or even more components. The Netzsch skimmer (commissioned this year) can be used in several configurations to collect data on molten systems. The skimmer uses mass spectrometry coupled with a thermogravimetric analyzer to accurately analyze systems up to 2000°C. Because the mass analyzer is coupled directly to the furnace through two nested orifices, there are no losses due to condensation. As a thermogravimetric analyzer, the skimmer measures weight loss and reaction energies simultaneously with evolved gas analysis. The mass spectrometry identifies gas phase species, and these volatility measurements can be used to derive vapor pressures. The skimmer was purchased in fiscal year 2018, delivered in April, installed in May 2019, and now is fully operational as verified by the analysis of known standards. The skimmer is unique because it will be available for study of salt systems that contain actinides or beryllium, although the additional work control to accommodate these hazards is still under review.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Physical Security Model Development of an Electrochemical Facility.

Nuclear facilities in the U.S. and around the world face increasing challenges in meeting evolving physical security requirements while keeping costs reasonable. The addition of security features after a facility has been designed and without attention to optimization (the approach of the past) can easily lead to cost overruns. Instead, security should be considered at the beginning of the design process in order to provide robust, yet efficient physical security designs. The purpose of this work is to demonstrate how modeling and simulation can be used to optimize the design of physical protection systems. A suite of tools, including Scribe3D and Blender, were used to model up a generic electrochemical reprocessing facility. Physical protection elements such as sensors, portal monitors, barriers, and guard forces were added to the model based on best practices for physical security. One outsider theft scenario was examined with 4-8 adversaries to determine security metrics. This work fits into a larger Virtual Test Bed 2020 Milestone in the Material Protection, Accounting, and Control Technologies (MPACT) program through the Department of Energy (DOE). The purpose of the milestone is to demonstrate how a series of experimental and modeling capabilities across the DOE complex provide the capabilities to demonstrate complete Safeguards and Security by Design (SSBD) for nuclear facilities. ACKNOWLEDGEMENTS This work was funded by the Materials Protection, Accounting, and Control Technologies (MPACT) working group as part of the Nuclear Technology Research and Development Program under the U.S. Department of Energy, Office of Nuclear Energy.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2019

Extraction of Mercury Utilizing 3D Printed Media-Mercury Extraction Coalescer (MEC)

This project was undertaken to address the concern of mercury species present in chemical reprocessing facilities and environmental remediation sites. Typical elemental mercury removal is performed by steam stripping. As efficient as this process is, the vaporized elemental mercury is often condensed along with acidic condensate. The acidic conditions re-dissolve the elemental mercury, transferring it back into the aqueous stream, and undoing the energy-intensive steam stripping process. This project looked to design and produce a 3D printed component to act as a mercury vapor/liquid coalescer utilizing structural and chemical features to segregate elemental mercury from an aqueous solution or process offgas and improving elemental mercury retention and removal from the recirculated condensate stream. The aim of this project is to take inspiration from nature, for example the lotus leaf, and the advantage of advanced manufacturing techniques, in this case 3D printing, to produce a functional solution to a serious environmental issue, the accumulation of elemental mercury in aqueous environments.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

First ARM Aerosol Chemical Speciation Monitor Users’ Meeting Report

The first U.S. Department of Energy (DOE) Atmospheric Radiation Measurement (ARM) user facility aerosol chemical speciation monitor (ACSM) users’ meeting was held on April 11-13, 2017, at Aerodyne Research, Inc. in Billerica, Massachusetts, to discuss the Southern Great Plains (SGP) atmospheric observatory ACSM data quality and establish best practices for data collection and processing. The participants examined six years of calibration and processed data. Specific issues raised by data users were addressed and case studies from two field experiments were examined. The most recent data from the ACSM installed in the newly commissioned SGP ARM Mobile Facility 7 (AMF7) were also evaluated. The participants recommended that the SGP ACSM data be reprocessed using calibration values averaged over the history of SGP ACSM calibrations. They also recommended that the data quality be evaluated by comparing (1) observed versus predicted particulate ammonium (NH 4 + ) mass loadings, and (2) ACSM mass loadings versus mass loadings calculated from particle size and light scattering data. The contents of the datastreams from the ACSM were defined based on ARM requirements and the necessary tasks to implement these recommendations were assigned to the mentor and the instrument manufacturer. Some of the recommendations were implemented for the SGP data and are presented in this report. Data from two ARM sites located in marine environments were also analyzed. These analyses as well as instrument intercomparisons have led to further questions about data treatment. These will be addresses in the second users’ meeting scheduled for the summer of 2018.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2019

Iodine Capture and Mechanical Stability of Modified Silver-Functionalized Silica Aerogel

Silver-functionalized silica aerogel (AgAero) is a material proposed for use in the removal of iodine from the off-gas streams arising from nuclear fuel reprocessing. This material was developed by Pacific Northwest National Laboratory and has been shown to adsorb up to 300 mg of iodine per gram of AgAero. One of the challenges noted for AgAero as an iodine sorbent is its mechanical stability. Its initial particle size is >0.85 mm, but previous testing has shown that extended exposure to heated gas streams results in the degradation of the material such that the sorbent bed is compacted with a substantial amount of fines produced. The observed mechanical degradation prompted an effort to reengineer the silica aerogel to improve the mechanical stability of the material without compromising its iodine capture characteristics. Pacific Northwest National Laboratory has modified the material with the intention of strengthening the aerogel backbone and improving mechanical durability. This reengineered material was transferred to Oak Ridge National Laboratory for extended testing to assess its mechanical durability. Modified AgAero was continuously exposed to a CH 3 I-bearing humid gas stream for 84 days. Testing was performed at 150°C with a superficial gas velocity of 10 m/min., a CH 3 I concentration of 554 ppb, and an inlet dew point of 10°C. The density of the AgAero increased by 16% during testing, and the height of the column was observed to decrease by 11%. The fines produced by the material were comparable to that of previously supplied material. Additionally, iodine loading of a thin bed of modified AgAero was performed at 150°C with a superficial gas velocity of 10 m/min. and a CH 3 I concentration of 50 ppm (balance dry air) with a total test duration of 1 week. The CH 3 I capture capacity of the reengineered material was 225 mg iodine per gram AgAero, as compared to a CH 3 I capacity of 250 mg iodine per gram AgAero for the original material. The primary data collected were visual observations, density measurements before and after testing, and the extent of column compaction during testing. Additionally, the material was tested to ensure that the iodine capacity of the material was not adversely impacted by the alternative methods used in the production of the reengineered material. Based on these criteria, no distinguishable improvements were observed between the material produced in fiscal years 2017 and 2019 and the mechanical stability of the aerogel in prototypic off-gas conditions remains a challenge to large-scale deployment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Extraction of Mercury Utilizing 3D Printed Media – Mercury Extraction Coalescer (MEC) (Revision 1)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Collaborative Research: Natural Organic Matter and Microbial Controls on Mobilization/Immobilization of I and Pu in Soils and Waters Affected by Radionuclide Releases in USA and Japan

In this project, the relationship between natural organic matter (NOM) and two radioactive elements that are relevant to nuclear waste disposal were studied: Plutonium (Pu) and Iodine. The human and environmental risks associated with Pu stem mainly from the very long half-lives of several of its isotopes ( 238 Pu, 88 yr; 239 Pu, 24,100 yr; 240 Pu, 6560 yr) and its radiotoxicity. Understanding Pu biogeochemical behavior in both near-field (>10-11M) and far-field scales (<10-11M) is imperative to the development of approaches for reprocessing Pu, remediation of Pu contamination and accurate assessment of risks posed by disposal practices for Pu-bearing wastes. The environmental mobility of Pu can be affected by redox potential, pH, adsorption, precipitation, complexation, colloid formation, and microbial activity, of which the first characteristic has the most profound influence. Numerous studies have shown high affinity of Pu towards NOM, as well as to mineral phases. NOM is ubiquitous in the environment, e.g., both fulvic and humic acids are able to reduce Pu(V,VI) to Pu(IV) and the redox potential of NOM is positively related to the abundance of phenolic/acidic OH groups. NOM can either facilitate or limit actinide migration, depending on specific biogeochemical conditions including pH, mineral and organic matter characteristics, etc. The other radionuclide of interest is radioiodine ( 129 I). 129I is a major by-product of nuclear fission and of serious concern to the Department of Energy (DOE) as it is among the top risk drivers at existing and potential radiowaste-contaminated sites. The risk of 129 I stems largely from its high bioconcentration factor (90% of the body’s iodine is accumulated in the thyroid), a high inventory at source terms, a very long-half life (16M years), and rapid mobility in the subsurface environment. As a consequence, 129 I has the lowest drinking water standard (1 pCi/L) among all radionuclides in the Federal Register. With a novel and sensitive gas chromatography-mass spectrometry (GC-MS) method developed in our lab, it is possible to quickly and simultaneously determine the distribution of 129 I and stable 127 I forms in environments, as low as 2 pCi/L for 129 I. This method was subsequently validated using accelerator mass spectroscopy, AMS. IO 3 - and organo-I were determined as major species in the groundwater of SRS and the Hanford Site, contrary to thermodynamic predictions that I- should be the dominant species at these sites. Mobility of 129 I was also demonstrated to depend greatly on the I species and its concentration, sediment pH, and redox state, with times to achieve equilibrium taking up to 12 weeks. Along the groundwater pathway in the F-Area of SRS, 129 I- supplied from the seepage basins was transformed to 129 IO 3 - and organo- 129 I with increasing iodine sediment sorption, causing the lower total 127 I and 129 I concentrations along the gradient transect of the waste plume. By contrast, groundwater 129 I concentrations in the wetlands (as high as 1617.3 pCi/L) were greatly elevated with respect to the source term (159.3 pCi/L). While the NOM promoted the uptake of 129 I to the wetland sediment, it also promoted the formation of soluble organic fraction. A small fraction of NOM that is bound to iodine can behave as a mobile organo-I source. Iodide was enzymatically incorporated into NOM, whereas both iodide and iodate were abiotically bound to NOM, under certain conditions. Iodate removal from the mobile aqueous phase can also occur through incorporation into carbonate (e.g., at the Hanford Site, USA). Thus immobilization and re-mobilization of iodine species were influenced by pH, Eh and the presence of NOM and metal oxides, which adds to the complexity of site remediation action. A ground-breaking result was to elucidate the products (i.e. organo-iodine moieties formed via enzymatic and non-enzymatic processes) at the molecular level by nuclear magnetic resonance (NMR) and electrospray ionization Fourier transform ion cyclotron resonance mass spectrometry (ESI-FTICRMS). We found that iodine-NOM interactions may be influenced by NOM hydrophobic aliphatic moieties. From the perspective of ESI-FTICRMS, organo-iodine formulas were ascribed to the groups of unsaturated hydrocarbons, lignins and proteins. Iodate is likely abiotically reduced to reactive iodine species by lignin- and tannin-like compounds or carboxylic-rich alicyclic molecules (CRAM). We also investigated microbial mechanisms in iodine incorporation into NOM. We established that soil bacteria isolated from F-Area of SRS did not accumulate significant amounts of I- (0.2-2%). Intracellular uptake of I- decreases with increasing pH when pH ranged from 4 to 6. In contrast, 44 out of 84 strains isolated from the F-Area of SRS can transform I- to IO 3 - and organo-iodine. In some cases, oxidation was facilitated in the presence of H 2 O 2 . Microbes can also excrete organic acids that enhance I- oxidation by lowering the ambient pH and reacting with H 2 O 2 to form peroxy carboxylic acids. At lower pH values (≤5), H 2 O 2 hydrolysis was the driving force for iodide-oxidation; whereas, at pH ≥ 6, spontaneous decomposition of peroxy carboxylic acids, originating from H 2 O 2 and organic acids were the primary cause of iodide oxidation. Lastly, it was determined that microbial processes involved in Mn (II) are capable of directly oxidizing I- via enzymatic catalysis (i.e., multicopper oxidases), or indirectly through the formation of reactive oxygen species (ROS) and/or biogenic manganese oxides. ROS-mediated oxidation of I- was found to predominate at pH >5, whereas the enzymatic and Mn oxide pathways were more active at pH < 5. Together, this project has resulted in 9 publications in high-impact journals, and the training of 1 Ph.D and 4 undergraduate students.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2020

Demand Driven Cycamore Archetypes

Future nuclear fuel cycle options may present advantages over today’s once-through fuel cycle. Nuclear fuel cycle simulation tools assess the performance of those fuel cycles as well as the dynamics of long-term technology transitions. In many nuclear fuel cycle simulation tools, it has historically been the responsibility of the user to manually define facility deployment schemes and all facility parameters. While this is straightforward in simple fuel cycles, transitions from one fuel cycle to another can be more complex. In particular, deployment schemes for supportive fuel cycle facilities beyond the reactor become complex if the analyst desires to avoid gaps in the nuclear fuel and power supply chain during those transition scenarios. As nuclear fuel cycle analysis approaches questions regarding the feasibility and performance of the deployment schemes and technology choices during technology transition, automation of this historically manual process is necessary. The main objective of this work was to develop and demonstrate Cyclus automation capabilities toward key nuclear fuel cycle transition scenarios. While deploying reactors to meet power demand is trivial, and existed in the earliest versions of CYCLUS, automated, predictive deployment and decommissioning of other facilities is more complex. These include mining, milling, enrichment, fuel fabrication, reprocessing, and others. For example, a balanced closed fuel cycle may require ensuring that there is enough fast reactor fuel for their operation and may drive deployment of a fleet of light water reactors. This concern comprises the main challenge that drove the project effort. The Demand-Driven Cycamore Archetype project (NEUP-FY16-10512) aimed to develop CYCAMORE demand-driven deployment capabilities and thereby automate transition scenario definition. The developed software package, d3ploy, in the form of a CYCLUS Institution agent, deploys Facilities to meet the front-end and back-end demands of the fuel cycle. The University of South Carolina and the University of Illinois applied multiple algorithmic approaches to this challenge. This project developed an in situ demand-driven development schedule calculation through non-optimizing, deterministic-optimizing, and stochastic-optimizing algorithms as CYCLUS archetypes and demonstrated these new archetypes in program-supporting fuel cycle transition scenarios. Both objectives were achieved. This report documents the results and deliverables obtained toward these achievements in detail.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Testing of an Iodine and Tritium Capture System for an NO₂-Based Tritium Pretreatment Process

The reprocessing of used nuclear fuel would release volatile radionuclides into the off-gas streams of a processing plant, including ³H, ¹⁴C, ⁸⁵Kr, and ¹²⁹I. One potential simplification to the management of the off-gas streams could be achieved through an efficient tritium pretreatment (TPT) step in which the UO₂ fuel is oxidized by either air or NO₂ before dissolution. The oxidation of the UO₂ fuel matrix will result in the release of tritium from the fuel. Upfront release of tritium from the fuel in a pretreatment step with subsequent tritium abatement can minimize or eliminate the distribution of tritium throughout a plant. This can decrease or eliminate the need for tritium capture on multiple off-gas streams and prevent distribution to the aqueous inventory. The use of NO₂ as the oxidant in an advanced tritium pretreatment (ATPT) allows for the oxidation to be performed at lower temperatures and may also result in the quantitative release of iodine from the fuel. The ATPT off-gas will contain up to 75 vol% NO₂ gas. Spencer et al. identified the most promising iodine sorbent as silver nitrate-impregnated alumina (AgA) and the most promising tritium sorbents as 3Å molecular sieve (3Å MS) and silica gel for radionuclide capture from ATPT off-gas streams. An initial series of tests were conducted in fiscal years 2017–2018 with the intent of demonstrating the recovery of iodine and tritium using these sorbents from a recirculating gas stream that simulated the NO₂ recycle loop envisioned for the ATPT system. These initial tests showed low tritium recoveries and significant variations in the iodine recovery on the AgA. Ultimately it was concluded, following disassembly of the system, that corrosion within the system was partially the cause of the poor recovery of both species. All components of that system were disposed of at the end of the analysis. One of the conclusions drawn from the initial tests pointed to the reconsideration of the materials of construction. Based on that recommendation a new test bed for the advanced TPT off-gas system, that addresses the issues identified in the fiscal years 2017 and 2018 scoping tests, was designed and constructed. The test bed used high nickel alloys wherever possible to address the previously observed corrosion issues, a redesigned surge tank to permit ease of purging, an improved feed injection system and a more modular system design that allowed testing of incremental increases in system complexity to allow the isolation of any element that results in changes to recovery of target species. Accompanying that design, a multifaceted test plan was completed. The overarching goals of this test series were two-fold. The first goal was to demonstrate the ability to recover both the iodine and tritium that would be released during TPT in a manner that the releases from the fuel can be quantified. The second goal was to demonstrate that the combined recovery system could operate in any of the three proposed TPT modes (once-through air or O₂, recirculating air or O₂, or recirculating NO x ). A total of 22 tests were completed. Ten tests examined tritium behavior under varying operating parameters, 8 tests examined iodine behavior under varying operating parameters, and 4 tests examined both tritium and iodine behavior. Silica gel was used for the tritium sorbent, and AgA or Ag-mordenite (AgZ) was used as the iodine sorbent. The following conclusions were drawn: Full recovery is achieved for both sorbates when no sorbent is in place indicating no accumulation within the test system. In even the simplest of the configurations (i.e., no sorbents in the columns), iodine is more difficult to transport and recover from the system than tritium. Analysis of the AgA iodine sorbent beds from the integrated tritium and iodine tests indicate quantitative iodine recovery can be achieved (within anticipated experimental errors). This system may provide a means for the quantification of the iodine released by the ATPT process within the experimental uncertainties. This would support 0.33 to 1.00 kg demonstrations of the ATPT process with actual used nuclear fuel. Iodine penetrated further into AgZ than into AgA when operated in gas recirculation mode. Tritium adsorbs to and cannot be fully recovered from the 10-cm column of silica gel at regeneration temperatures up to 140 °C.; Tritium adsorbs to and cannot be recovered from the 10-cm column of AgA at desorption temperatures up to 200 °C. Co-adsorbed tritium was recovered to a greater extent from AgZ than from AgA, but recovery was not complete. No visible corrosion was observed in the test system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

The Fission-Fusion Hybrid Molten Salt Reactor: A Clean Energy Game Changer

This White Paper proposes the Fission-Fusion Hybrid Molten Salt Reactor (FFHMSR) as the solution to the world’s growing need for abundant, carbon-free, safe electricity. The FFHMSR combines a critical molten salt fission reactor with a fusion source of energetic neutrons in a single connected system, offering advantages that neither has independently. The molten salt reactor provides improved safety and greater efficiency, while the neutron source’s high energy neutrons eliminate the need for fuel enrichment and reprocessing and can remove all actinides from the waste stream. The dynamic balance between the dual processes capitalizes on their strengths and synergies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Critical Materials Capabilities at LANL [Slides]

Critical materials are a recognized problem. In addition to being it’s own cross-cutting topic area, it is called out in Advanced Energy Storage Initiative, Transportation Sector Priorities, and Energy Efficiency Sector Priorities. The need for domestic battery technology is a priority. Domestic supply, separations and processing technologies are required to reduce dependence on foreign capabilities. LANL maintains many capabilities that are applicable to REEs and critical materials: Actinide processing capability for defense programs and extensive separation capabilities – trace analysis up to pilot scale. Development of new approaches for reprocessing technologies are often tested first on lanthanides.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Current Needs in Radiochemistry

There are a number of current needs in the development of the Pu and Am separations that underpin larger interest areas including nuclear waste issues, nuclear weapons, nuclear forensics, nuclear fuels and similarly related fields. In the large areas of interest, there are a number of sub-fields that include corrosion chemistry, pyrochemistry, solution-phase chemistry, rapid separations, environmental speciation, and other similar efforts. However, in all of the these fields it is important to realize that the underlying chemistry that underpins these efforts is similar in that Pu and Am are bound to ligands that can be in the solid, aqueous, or gas phase and separations occurs due to the preferential binding of the ligand to Pu or Am based on a number of factors that include pH, concentration, oxidation state, etc. In the past, most of the Pu and Am chemistry has occurred in limited scope within universities where micro-scale and bench-top scale chemistry are the dominate investigative theme. At US national laboratory and their international counterparts, larger scale reactions have been attempted such that industrial scale chemistry has been completed. Some of this technology has been harnessed in radiopharmaceuticals and nuclear fuel reprocessing. However, such bench-top research is expensive to due to the safety and security that is concerned when dealing with Pu, and scale-up of reactions is even more resource intensive. In an effort to offset this expensive, it would be of interest to explore the ability to predict selectivity based on the computational modeling of actinide chemistry. This has historically been limited by the abilities of large supercomputers, but with the emphasis on developing exascale computers within the US and similar efforts by other countries this predictive modeling has potential to produce significant results that would yield improved understanding of actinide chemistry.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2020

Effect of NO x and Water Variations on Iodine Loading of AgZ

Reprocessing of used nuclear fuel will result in the release of several volatile radionuclides that need to be removed from facility off-gas streams before their release to the environment. Iodine, one of these radionuclides, is expected to be primarily released in the dissolver off-gas stream as iodine (I 2 ) or methyl iodide (CH 3 I). Sorbents that target I 2 and CH 3 I removal from the DOG stream will need to adsorb iodine under elevated temperatures and in the presence of water vapor and nitrogen oxides (NO x ). Previous studies examining the adsorption of CH 3 I and I 2 in the presence of NO x gases by silver-exchanged mordenite (AgZ), a zeolite mineral considered for use in this application, have resulted in slightly inconsistent conclusions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

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 OSTI2020

Removal of High Specific Activity Fission Products from Uranyl Sulfate Waste Solutions

The Savannah River National Laboratory (SRNL) is currently providing support to SHINE Medical Technologies (SHINE) which plans to deploy a low energy, accelerator-based neutron source to fission low enriched U in a uranyl sulfate target solution for 99 Mo production. The 99 Mo is initially separated from the fission products and target solution by an extraction column. Subsequent washing of the column will generate waste solutions containing residual U and fission product activity. A small number of high specific activity fission products (e.g., 90 Sr, 137 Cs, and 144 Ce) in these streams will likely control the classification of the low level waste (LLW). If a sufficient amount of the high specific activity isotopes are separated from the SHINE waste streams and concentrated into a waste form, it would be possible to treat a majority of the wash solutions from the column operations as a lower class of LLW (Class A versus Class B or C or Class B versus Class C). The high specific activity fission product elements could then be dispositioned as a much smaller volume of waste rather than requiring the disposal of the entire waste stream at the higher waste classification. The Savannah River Site (SRS) has experience with using monosodium titanate (MST) and crystalline silicotitanate (CST) to remove Cs and Sr from high salt content waste solutions generated during the reprocessing of nuclear fuels and targets. Both of these materials have worked very well for their intended purposes at the SRS where the fission product elements are removed from highly alkaline waste. On the other hand, SHINE waste streams from the extraction column contain H 2 SO 4 which makes the solution acidic. Additionally, the SRS waste streams do not contain other fission product elements such as transition metals or lanthanides because they precipitate upon neutralization of the SRS waste and are not present in the supernate which is dispositioned as LLW following treatment. As such, there are inherent differences between SHINE and SRS waste treatment strategies. Savannah River National Laboratory was tasked with performing scoping studies to see if MST and CST would remove Sr, Cs, and Ce from an acidic mixed metal simulant solution. Batch contact experiments were performed using MST and two CST type materials. The MST material is a 15 wt % powder in 0.15 M NaOH slurry. The MST showed low adsorption for elements of interest from acidic solution. Furthermore, the powder size makes MST non-ideal for column operations. A CST IE-911 ion exchange material had high Cs adsorption, moderate Sr, and marginal Ce adsorption. Based on adsorption of all species, the ion exchange capacity was found to be 0.032 meq/mL. A bench-top column experiment to measure elemental breakthrough curves was performed using CST IE-911 where chromatographic separations of the mixed simulant were expected to occur. While most elements behaved as expected, the lanthanide series, containing Ce, broke though the column earlier than expected. The second CST material, CST R9120, displayed high adsorption for all elements in the acidic mixed simulant solutions in a batch contact study, and had a calculated loading capacity of 0.091 meq/mL. Future studies to develop a waste treatment flowsheet should focus on CST R9120 to treat SHINE waste solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Actinide N-Donor Thermodynamics: Expanding the f-element Covalency Dialogue. Final report

Resolving the chemistry and physics of f-electrons is one of the grand challenges of science for energy technology. A central component of this challenge is describing covalency in actinide-ligand interactions. Crystallography, x-ray spectroscopy and computational techniques have provided reasonable, and in some cases remarkable, support for covalency in actinide-ligand interactions. Courtesy the quality of the results obtained, a significant portion of the covalency dialogue has been forwarded through solid-state and computational studies of the more stable actinides (Th, U, Pu, Np). However; recent reports have questioned whether covalency in actinide interactions increases or decreases across the series. Examination of less-stable trans-actinides (Am, Cm, Bk, Cf and Es) by the aforementioned approaches can be rapidly limited due to material availability, radiological hazards and experimental data to validate computational models. Furthermore, limited thermodynamic data exists to confirm covalency in actinide solution phase interactions that are highly relevant to the remediation and reprocessing of used nuclear fuel. The best-defined interactions of actinides with soft donors in aqueous solutions involve (poly)aminopolycarboxylate (APC) ligands. The chelate effect and binding affinity with the acetic acid APCs subgroups encourages amine interactions with the actinide metal center. In the absence of these factors, amine interactions with actinides are too weak to overcome the protective hydration shell of the dissolved ion. The ability for APCs to force actinide interactions with soft nitrogen donors (as defined by Pearson’s Hard Soft Acid Base theory) encourages the application of these ligands in a variety of processes for actinide recovery from nearly chemically identical lanthanides. The ability to functionalize the amine center of the ligand in a variety of capacities (adding additional amine groups, exchanging the conventional acetate group for an acetate group, etcetera), allows for the APC ligand to serve as a thermodynamic probe for actinide-nitrogen interactions. Perhaps the single most significant breakthrough during the previous funding cycle was the observation that covalency for the transplutonium part of the actinide series can be increasingly influenced with energy degeneracy driven covalency as the actinides become heavier. This was observed most predominantly with dipicolinic acid, but extensions of f-orbital degeneracy were found to affect aliphatic aminopolycarboxylate-actinide complexes through einsteinium.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2020

Physical Security Model Development of an Electrochemical Facility

Nuclear facilities in the U.S. and around the world face increasing challenges in meeting evolving physical security requirements while keeping costs reasonable. The addition of security features after a facility has been designed and without attention to optimization (the approach of the past) can easily lead to cost overruns. Instead, security should be considered at the beginning of the design process in order to provide robust, yet efficient physical security designs. The purpose of this work is to demonstrate how modeling and simulation can be used to optimize the design of physical protection systems. A suite of tools, including Scribe3D and Blender, were used to model a generic electrochemical reprocessing facility. Physical protection elements such as sensors, portal monitors, barriers, and guard forces were added to the model based on best practices for physical security. Two theft scenarios (an outsider attack and insider diversion) as well as a sabotage scenario were examined in order to optimize the security design. Security metrics are presented. This work fits into a larger Virtual Facility Distributed Test Bed 2020 Milestone in the Material Protection, Accounting, and Control Technologies (MPACT) program through the Department of Energy (DOE). The purpose of the milestone is to demonstrate how a series of experimental and modeling capabilities across the DOE complex provide the capabilities to demonstrate complete Safeguards and Security by Design (SSBD) for nuclear facilities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Fission Gas Measurements in Spent Fuel: Literature Review

The fission-derived noble gases, xenon(Xe) and krypton(Kr), have long been proposed as a tool for supporting nuclear safeguards and declarations verification. A vast body of literature outlines how these non-reactive fission gases released during reactor operation and shutdowns correlate to burnup level, power history, other reactor specifications and reprocessing activities. However, irradiation conditions can vary appreciably both radially and axially within a fuel pin.

07 ISOTOPE AND RADIATION SOURCES↗