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Cooperative self-assembly for structure and morphology control of energetic materials

The performance of energetic materials (EM) varies significantly across production lots due to the inability of current production methods to yield consistent morphology and size. Lot-to-lot variations and the inability to remake the needed characteristics that meet specification is costly, increases uncertainty, and creates additional risk in programs using these materials. There is thus a pressing need to more reliably formulate EMs with greater control of morphology. The goal of this project is to use the surfactant-assisted self-assembly to generate EM particles with welldefined size and external morphologies using triaminotrinitrobenzene (TATB) and hexanitrohexaazaisowurtzitane (CL-20) as these EMs are both prevalent in the stockpile and present interesting/urgent reprocessing challenges. We intend to understand fundamental science on how molecular packing influences EM morphology. We develop scale up fabrication of EM particles with controlled morphology, promising to eliminate inconsistent performance by providing a trusted and reproducible method to improve EMs for NW applications.

36 MATERIALS SCIENCE↗

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↗

Electrolytic Oxide Reduction of Plutonium Oxide Surrogates

Our research team has investigated electrolytic oxide reduction (EOR) as an improved metal production process versus the currently employed multiple cycle direct oxide reduction (MCDOR) process for the conversion of plutonium oxide to plutonium metal. Inherent advantages in safety, waste generation, efficiency, and operational ease make EOR an attractive alternative to traditional metal production practices. Phase I of the project consisted of the construction of a prototype electrochemical cell, and proof of concept experiments towards the reduction of SnO 2 and CeO 2 compounds. Phase II of the project was focused on the optimization of the cerium system in preparation for future plutonium EOR experiments. We observed the electrolytic reduction of cerium oxide resulting in mixed degrees of success, and discovered this was a result of parasitic side reactions and partial reduction. We employed cyclic voltammetry (CV) to identify the products of these side reactions and characterize partially reduced species. We also observed that, when metal was successfully produced by EOR, a large amount of passed charge was required. We concluded that this observation was due to several inefficiencies of the process relating to the reduction mechanism (e.g. inefficient use of cathodically produced calcium). We used our CV results to guide our optimization of the EOR of CeO 2 by refining our electrochemical method, cell geometry, and materials of construction. Finally, we developed a next generation cell, worked to establish a new molten salt electrochemical capability in the Plutonium Facility, and completed a strategy for plutonium experiments. Our electrochemical method to produce plutonium metal clearly supports the core mission of the laboratory. Furthermore, this project has allowed for the development of capabilities that will provide valuable data to programs interested in characterization and detection, as well as science of signatures. Additionally, establishing this capability in the plutonium facility will facilitate future safeguard studies that can assist with safeguard development in molten salt reprocessing flowsheets, and molten salt reactor designs.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Dissolution of Used Nuclear Fuel using a TBP/n-paraffin Solvent

The complete recycle of used nuclear fuel (UNF) is desirable to recover the energy value from the remaining U and the higher actinides produced during fuel irradiation; however, the construction of a large-scale reprocessing facility is prohibitively expensive. A simplified flowsheet with a reduced equipment footprint is needed to improve the economic viability. To address this issue, we evaluated a hybrid process which combines the dissolution of UNF pretreated (oxidized) for tritium removal in a tributyl phosphate (TBP) solvent with the solvent extraction cycles required for recovery of the actinide and lanthanide elements. To demonstrate the feasibility of the hybrid process, we prepared actinide solid solutions containing selected fission product elements as surrogates for pretreated, irradiated light water reactor fuel. The distribution of the actinide and fission product elements was measured between the solvent and residual solids to determine dissolution efficiencies. Dissolution of the solid solutions in the TBP solvent was effective for the recovery of both the actinide and lanthanide elements. An increase in the TBP concentration and temperature improved the dissolution efficiencies for these elements. Two-stage dissolutions were successfully used to recover nearly all the actinides and most of the lanthanides from the solid solutions. Our data indicate that most of the fission products generated during fuel irradiation except for the lanthanides and Tc (based on data for Re) would remain in the undissolved solids. The Am/Cm, lanthanides, and other fission product are easily scrubbed from the solvent by contact with nitric acid generating a pure U/Np/Pu stream.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Steam Oxidation and Burst Testing in the Severe Accident Test Station

Within six months of the March 2011 nuclear accident at Fukushima Daiichi, research began at Oak Ridge National Laboratory (ORNL) to simulate the high temperature steam oxidation accident conditions and identify potential accident tolerant fuel (ATF) cladding [1-14]. The ORNL laboratory equipment for steam oxidation equipment was expanded and made available to the entire ATF research community as the Severe Accident Test Station (SATS) in 2012 [5,1 5], including the capability to conduct the standard integral loss of coolant accident (LOCA) experiment [9]. A second SATS has now been deployed in an ORNL hot cell for handling irradiated specimens including reprocessed UO 2 /Zr fuel [16].

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Global Nuclear Markets – Market Arrangements and Service Agreements

The U.S. Department of Energy’s Office of Energy Policy and Systems Analysis (EPSA) requested an assessment of global nuclear markets, including the structure of nuclear companies in different countries and the partnerships between reactor vendors and buyers. EPSA was interested in obtaining information on the competitive context of international sales of reactors and fuel services. The Idaho National Laboratory responded to this request with a plan for information gathering and assessment of global markets in several phases. The first phase researched global sources and developed a collection of information to assist in analyses of global market status and trends in services provided in conversion, enrichment, reactor design, construction and operation, and discharged fuel storage and reprocessing. This report summarized this first phase, including analysis conclusions about current global markets. Additional phases will address specific topics that are of particular interest to EPSA.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

Performance Evaluation of the CHKED Multi-Elemental Analysis and Software

The Hybrid K-edge Densitometer (HKED) is a high precision safeguards assay system developed by Ottmar and Eberle [1] for quantification of the U and Pu concentrations of dissolver solutions in spent nuclear fuel reprocessing. The HKED measurement supplements traditional destructive assay methods such as Isotope Dilution Mass Spectrometry (IDMS) and is an essential component of the international safeguards program implementation at these facilities. The Canberra Hybrid K-Edge Densitometry (CHKED) software was recently developed by Canberra Industries to better address the changing composition of the nuclear fuel cycle by addition of a multi-elemental analysis capability to the K-Edge transmission measurement. Our evaluation of the CHKED software was performed at the request of the International Atomic Energy Agency (IAEA) to examine the functionality and performance of the CHKED software and algorithms and identify potential areas for improvement of the software. The reader is assumed to be familiar with the HKED technique and its current implementation in safeguards assay of U and Pu in the nuclear fuel cycle. The Canberra Hybrid K-Edge Densitometry CHKED software replicates functionality of the older VMSbased Hybrid K-Edge Densitometer (HKED) software and incorporates additional analysis algorithms and tools. When the traditional HKED analyses are selected, the CHKED software appears to faithfully reproduce the measurement results as reported by the older software. The CHKED software includes a Multi-elemental K-Edge (MEKED) analysis based in part on algorithms devised at Los Alamos National Laboratory. In principle, the MEKED analysis represents a significant advancement over the traditional HKED analysis, providing a capability to analyze the actinide concentrations from complex solutions, such as low U:Pu ratios (e.g., U:Pu = 1:1) including additional minor actinide components such as Np and Am. However, this MEKED analysis is somewhat limited by reliance on a region of interest (ROI) analysis of the spectrum, the need for extensive calibration measurements and a simplistic interpretation of the KED transmission spectrum. Finally, the MEKED algorithm set associated with the CHKED software (V1.0) acquired by Oak Ridge National Laboratory (ORNL) from Canberra Industries does not seem to be “finished” (e.g., the MEKED analysis seems to only determine measurement results for two actinides with while results for the minor actinides appear to be highly dependent on declaration values).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Design of, and Test Plan for, an Iodine and Tritium Capture System for a NO 2 -Based Tritium Pretreatment Process

The reprocessing of used nuclear fuel would release volatile radionuclides into the off-gas of a processing plant, including 3 H, 14 C, 85 Kr, and 129 I. One potential simplification to the management of the off-gas streams could be achieved through an efficient tritium pretreatment (TPT) step in which the UO 2 fuel is oxidized either by air or by NO2 before dissolution. Oxidation of the UO2 fuel matrix results in the release of tritium contained within. Upfront removal of tritium from the fuel in a pretreatment step, followed by the abatement of tritium, can minimize or eliminate the distribution of tritium throughout the plant, also decreasing or eliminating the need for tritium capture on multiple off-gas streams. The use of NO 2 as the oxidant in an advanced tritium pretreatment (ATPT) allow the oxidation to be performed at lower temperatures and may result in the quantitative release of iodine from the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Phase separation and crystallization of complex borosilicate melts for glass-ceramic waste forms

Waste loadings of reprocessed spent nuclear fuel vitrified into borosilicate glass can be increased by precipitating environmentally stable phases concentrated with waste components in a chemically stable glass matrix. The principal objective of this study was to characterize the development of crystalline powellite (CaMoO 4 and related phases) and oxyapatite (Ca 2 LN 8 Si 6 O 26 ) in borosilicate glass-ceramics and to determine how the formation of those phases affected its chemical durability. Borosilicate glasses provided by PNNL were re-melted and quenched at rates from over 300°C/s to ~0.05°C/s. Isothermal heat treatment experiments were conducted by quenching melts in a molten tin bath at various temperatures, holding for periods of time, and then quenching in a water bath. Analytical electron microscopy and x-ray diffraction provided information about the kinetics of the phase separation and crystallization processes responsible for microstructural development. Powellite and oxyapatite crystals formed during slower quench rates and longer isothermal times, and time-temperature-transformation (TTT) diagrams were developed from the latter experiments. A hot thermocouple test (HTT) system was built to rapidly quench melts and hold them at set times to provide supplementary TTT information, but that system was not as sensitive to the phase transformations as other tests. Corrosion tests were performed to understand how the individual phases in the glass-ceramic affect its overall chemical durability. Product consistency tests, performed at Missouri S&T and at Savannah River National Lab, provided release rates of major elements from samples as a function of cooling rate, and atomic force microscopy and profilometry measurements of surface topology determined the relative corrosion rates of the residual glass and oxyapatite phases. Faster dissolution rates were measured from samples cooled more slowly and these were explained by the greater fractions of B 2 O 3 in the residual glass phase after the formation of oxyapatite and powellite. Oxyapatite was found to be more durable than the residual glass.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Solid-liquid Separation Testing for the Remediation of Hanford Waste Treatment Plant Low Activity Waste Melter Off-Gas Condensate

The Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate stream (LAW Off-Gas Condensate) from the off-gas system. The plan for disposition of this stream during Direct Feed LAW operations is to transfer it to the Effluent Management Facility (EMF), where it will be concentrated by evaporation and recycled to the LAW vitrification facility again. The primary reason to recycle this stream is so that the semi-volatile 99Tc isotope eventually becomes incorporated into the glass. However, this stream also contains non-radioactive salt components that are problematic in the melter, so diversion of this stream to another process would eliminate recycling of these salts and would enable simplified operation of the LAW melter and the Pretreatment Facilities. This diversion from recycling this stream within WTP would have the effect of decreasing the quantity of glass waste produced by allowing further increases in waste loading, resulting in a shortened LAW vitrification mission duration. The concept being tested here involves selectively removing the 99 Tc so that the decontaminated aqueous stream, with the problematic salts, can be disposed as a different waste form and avoid reprocessing through vitrification.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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↗

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↗

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↗

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↗

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↗

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

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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↗

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↗