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At least 1,207 records · Page 67

Chemical durability of multicomponent lanthanide zirconate solid solutions

Multicomponent zirconate solid solutions A 2 Zr 2 O 7 (A= (Sm,Yb), (Sm,Yb,Gd), (Sm,Yb,Gd,Er), and (Sm,Yb,Gd,Er,Dy)) were synthesized through spark plasma sintering and their chemical durability was investigated via semi-dynamic leaching tests in pH = 1 nitric acid at 90°C. Incongruent dissolution occurred for all samples with the preferential elemental release of zirconium in all A 2 Zr 2 O 7 except for high entropy (Sm,Yb,Gd,Er,Dy)Zr 2 O 7 which displays a congruent release of the A-site elements and Zr. A strong positive correlation was found between the lanthanide elemental release rates and the A-site cation size disorder, whereas a negative correlation was observed between the elemental release rates and mixing entropy.

36 MATERIALS SCIENCE↗

The function of Sn(II)-apatite as a Tc immobilizing agent

Technetium-99 is a radioactive contaminant of high concern at many nuclear waste storage sites. At the U.S. Department of Energy Hanford Site, 99Tc is a component of low-activity waste (LAW) fractions of the nuclear tank waste, which are highly caustic, high ionic strength and have high concentrations of chromate. Removal of 99Tc from LAW streams would greatly benefit the site remediation process. In this study, we investigated the removal of 99Tc(VII), as pertechnetate, from deionized water (DIW) and a LAW simulant using two solid sorbents, tin (II) apatite (Sn-A) and SnCl2 through batch sorption testing and solid phase characterization. Sn-A showed higher levels of removal of Tc from both DIW and LAW simulant compared with the SnCl2. Scanning electron microscopy/energy dispersive X-ray spectroscopy (SEM/XEDS) and X-ray adsorption spectroscopy (XAS) of Sn-A following batch experiments in DIW showed that TcO4- is reduced to Tc(IV) on the Sn-A surface with no incorporation into the lattice structure of Sn-A. The performance of Sn-A in the LAW simulant was lowered due to a combined effect of the high alkalinity, which lead to an increased dissolution of Sn from the Sn-A, and a preference for the reduction of Cr(VI) over Tc(VII).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Sustainable sanitary landfills for neglected small cities in developing countries: The semi-mechanized trench method from Villanueva, Honduras

Highlights: Black-Right-Pointing-Pointer Open dumping is the most common form of waste disposal in neglected small cities. Black-Right-Pointing-Pointer Semi-mechanized landfills can be a sustainable option for small cities. Black-Right-Pointing-Pointer We present the theory of design and operation of semi-mechanized landfills. Black-Right-Pointing-Pointer Villanueva, Honduras has operated its semi-mechanized landfill for 15 years. Black-Right-Pointing-Pointer The cost of operation is US$4.60/ton with a land requirement of 0.2m{sup 2}/person-year. - Abstract: Open dumping is the most common practice for the disposal of urban solid wastes in the least developed regions of Africa, Asia and Latin America. Sanitary landfill design and operation has traditionally focused on large cities, but cities with fewer than 50,000 in population can comprise from 6% to 45% of a given country's total population. These thousands of small cities cannot afford to operate a sanitary landfill in the way it is proposed for large cities, where heavy equipment is used to spread and compact the waste in daily cells, and then to excavate, transport and apply daily cover, and leachate is managed with collection and treatment systems. This paper presents an alternative approach for small cities, known as the semi-mechanized trench method, which was developed in Villanueva, Honduras. In the semi-mechanized trench method a hydraulic excavator is used for 1-3 days to dig a trench that will last at least a month before it is filled with waste. Trucks can easily unload their wastes into the trench, and the wastes compact naturally due to semi-aerobic biodegradation, after which the trenches are refilled and covered. The exposed surface area is minimal since only the top surface of the wastes is exposed, the remainder being covered by the sides and bottom of the trench. The surplus material from trench excavation can be valorized for use as engineering fill onsite or off. The landfill in Villanueva has operated for 15 years, using a total land area of approximately 11 ha for a population that grew from 23,000 to 48,000, with a land requirement of 0.2 m{sup 2}/person year, a cover to waste ratio of 0.2, and an estimated soil surplus of 298,000 m{sup 3} that is valorized and used onsite. The landfill has been operated solely by the municipality with an operational cost in 2010 estimated at US$4.60 per ton. A modified water balance analysis at Villanueva shows negligible leachate generation from covered trenches and 700 m{sup 3}/yr (60 m{sup 3}/ha yr) from the two open trenches required for daily operation. If the site were an open dump, however, leachate generation is estimated to be 3900 m{sup 3}/ha yr and contaminated runoff 5000 m{sup 3}/ha yr. A simple model used to estimate dilution of generated leachate based on groundwater flow data and aquifer stratigraphy suggests that the leachate will be diluted by a factor of 0.01 in the aquifer. Leachate contaminants will not accumulate because the aquifer discharges to the Ulua River 2 km south of the landfill. While not suitable for all sites, the Villanueva method nevertheless serves as an excellent example of how a small city landfill with natural compaction of waste and attenuation of leachate can be sustainably operated.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Thermodynamics of CeSiO 4 : Implications for Actinide Orthosilicates

Zircon (ZrSiO 4 : I4 1 /amd) can accommodate actinides, such as thorium, uranium, and plutonium. The zircon structure has been determined for several of the end member compositions of other actinides, such as plutonium and neptunium. However, the thermodynamic properties of these actinide zircon structure-types are largely unknown due to the difficulties in synthesizing these materials and handling transuranium actinides. Thus, we have completed a thermodynamic study of cerium orthosilicate, stetindite (CeSiO 4 ), a surrogate of PuSiO 4 . For the first time, the standard enthalpy of formation of CeSiO 4 was obtained by high temperature oxide melt solution calorimetry to be -1971.9 ± 3.6 kJ/mol. Stetindite is energetically metastable with respect to CeO 2 and SiO 2 by 27.5 ± 3.1 kJ/mol. The metastability explains the rarity of the natural occurrence of stetindite and the difficulty of its synthesis. Applying the obtained enthalpy of formation of CeSiO 4 from this work, along with those previously reported for USiO 4 and ThSiO 4 , we developed an empirical energetic relation for actinide orthosilicates. Here, the predicted enthalpies of formation of AnSiO 4 are then made with a discussion of future strategies to efficiently immobilize Pu or minor actinides in the zircon structure.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

High-Temperature Thermodynamics of Cerium Silicates, A-Ce 2 Si 2 O 7 , and Ce 4.67 (SiO 4 ) 3 O

Lanthanide disilicates and oxyapatites have potential roles in high temperature applications as thermal (TBC) and environmental barrier coatings (EBC), or possible alteration phases in geological nuclear waste repositories. However, those Ce 3+ -bearing silicates have only been limitedly studied. In this work, we performed detailed structural and thermodynamic investigations on A-Ce 2 Si 2 O 7 (tetragonal, P4 1 ) and Ce 4.67 (SiO 4 ) 3 O (hexagonal, P6 3 /m). The high temperature structural behaviors and coefficients of thermal expansion were determined by in situ high temperature synchrotron X-ray diffraction (HT-XRD) implemented with Rietveld analysis and thermogravimetric analysis coupled with differential scanning calorimetry (TGA-DSC). A-Ce 2 Si 2 O 7 was found to be stable in N 2 and air up to ~1483 K with anisotropic thermal expansion along the a and c axes (α a = 12.3 × 10 -6 K -1 and α c = 12.4 × 10 -6 K -1 ). Ce 4.67 (SiO 4 ) 3 O had a slow partial oxidation between 533 K and 873 K to a new nonstoichiometric phase Ce 3+ 1.67- xCe4 + x Ce 3+ 3 (SiO 4 ) 3 O 1+0.5x , followed by a thermal decomposition to CeO 2 and SiO 2 at ~1000 K in air. By using high temperature oxide melt solution calorimetry at 973 K with lead borate as the solvent, the standard enthalpy of formation was determined for A-Ce 2 Si 2 O 7 (-3825.1 ± 6.0 kJ/mol) and Ce 4.67 (SiO 4 ) 3 O (-7391.3 ± 9.5 kJ/mol). Finally, these thermodynamic parameters were compared with those of CeO 2 , CeSiO 4 , and other silicate oxyapatites for examining their chemical stability in high temperature environments relevant for aeronautical applications, mineral formation, and nuclear fuel cycle.

36 MATERIALS SCIENCE↗

Influence of Metal Ion Complexation on the Radiolytic Longevity of Butyramide Extractants under Direct Dissolution Process Conditions

The direct dissolution of voloxidized used nuclear fuel (UNF) into an organic solution–comprised of diluent and specialized extractants–poses a promising alternative to the traditional liquid–liquid solvent extraction approach to reprocessing UNF. However, moving to direct dissolution removes the presence of a concentrated nitric acid aqueous phase, which has been shown to significantly influence the radiolytic longevity of extractants in conventional extraction flowsheets. Given the limited knowledge of radiation effects under direct dissolution conditions, here we present a time-resolved and dose-accumulation study on the impact of direct dissolution conditions on the radiolytic longevity of two candidate butyramide extractants, N,N-di(2-ethylhexyl) butyramide (DEHBA) and N,N-di(2-ethylhexyl)isobutyramide (DEHiBA), in pre-equilibrated n-dodecane solvent in the presence and absence of process-relevant metal ions, specifically, uranium and rhenium. Loss G(DEHBA) and G(DEHiBA) values were found to be comparable to each other, with an average of 0.37 ± 0.02 μmol J –1 , and to previous data from the γ irradiation of DEHBA and DEHiBA under conventional solvent extraction conditions. Rhenium, and by extension technetium, extraction had a modest decrease (~10%) in the overall radiolytic stability of DEHiBA only, despite >2× observed increases in chemical kinetic reactivity of the corresponding complexes with the n-dodecane radical cation. Uranium loading, on the other hand, significantly improved the lifetime of both ligands (>30%) under γ irradiation, with a greater stabilization observed for DEHBA over DEHiBA. The observed radioprotective effect afforded by uranium loading is fortuitous for the longevity of direct dissolution solvent.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Multifaceted Modularity: A Key for Stepwise Building of Hierarchical Complexity in Actinide Metal–Organic Frameworks

Growing necessity for efficient nuclear waste management is a driving force for development of alternative architectures towards fundamental understanding of mechanisms involved in actinide integration inside extended structures. In this manuscript, metal-organic frameworks (MOFs) were investigated as a model system for engineering radionuclide containing materials through utilization of unprecedented MOF modularity, which cannot be replicated in any other type of materials. Through the implementation of recent synthetic advances in the MOF field, hierarchical complexity of An-materials were built stepwise, which was only feasible due to preparation of the first examples of actinide-based frameworks with “unsaturated” metal nodes. The first successful attempts of solid-state metathesis and metal node extension in An-MOFs are reported, and the results of the former approach revealed drastic differences in chemical behavior of extended structures versus molecular species. Successful utilization of MOF modularity also allowed us to structurally characterize the first example of bimetallic An-An nodes. To the best of our knowledge, through combination of solid-state metathesis, guest incorporation, and capping linker installation, we were able to achieve the highest Th wt% in mono- and bi-actinide frameworks with minimal structural density. Overall, combination of a multistep synthetic approach with homogeneous actinide distribution and moderate solvothermal conditions could make MOFs an exceptionally powerful tool to address fundamental questions responsible for chemical behavior of An-based extended structures, and therefore, shed light on possible optimization of nuclear waste administration.

Actinide↗

Chalcogen-based aerogels as a multifunctional platform for remediation of radioactive iodine

Aerogels employing chalcogen-based (i.e., S, Se, and/or Te) structural units and interlinking metals are termed chalcogels and have many emerging applications. Here, chalcogels are discussed in the context of nuclear fuel reprocessing and radioactive waste remediation. Motivated by previous work on removal of heavy metals in aqueous solution, we explored the application of germanium sulfide chalcogels as a sorbent for gas-phase I2 based on Pearson's Hard/Soft Acid-Base (HSAB) principle. This work was driven by a significant need for high-efficiency sorbents for I-129, a long-lived isotope evolved during irradiated UO2 nuclear fuel reprocessing. These chalcogel compositions are shown to possess an affinity for iodine gas, I2(g), at various concentrations in air and the affinity is attributed to a strong chemical attraction between the chalcogen and I2(g), according to the HSAB principle. The high sorption efficiency is facilitated by the high porosity as well as the exceptionally large surface area of the chalcogels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Neutron Total Scattering Analysis of Materials for the Nuclear Fuel Cycle

Here, this paper reviews recent experimental efforts at the University of Tennessee and Oak Ridge National Laboratory to comprehensively characterize the structural details of materials relevant for the nuclear fuel cycle by employing advanced neutron scattering techniques. For the study of nuclear ceramics, neutron scattering offers distinct advantages over traditional laboratory or synchrotron X-ray diffraction, including enhanced sensitivity to elements with a low atomic mass, such as oxygen, nitrogen, and carbon. The key to these efforts is the recent advancement in the neutron scattering infrastructure at the high-flux diffractometers at the Spallation Neutron Source. The high neutron flux at these instruments enables neutron total scattering, a nondestructive bulk technique that simultaneously captures both short-range structural effects through pair distribution function analysis and long-range order through diffraction pattern analysis. This approach is particularly important for a comprehensive description of defective, disordered, or amorphous nuclear materials. The case studies presented here include analyses of the local defect structure in hyperstoichiometric uranium oxides and short-range order of ion-irradiated ceramics. This advanced analytical methodology will improve our understanding of the behavior of materials in extreme environments and contribute to the development of more resilient nuclear materials.

Neutron scattering↗

Westinghouse modular grinding process - improvement for follow on processes

In nuclear power plants (NPP) ion exchange (IX) resins are used in several systems for water treatment. The resins can be in bead or powdered form. For waste treatment of spent IX resins, two methods are basically used: Direct immobilization (e.g. with cement, bitumen, polymer or High Integrity Container (HIC)); Thermal treatment (e.g. drying, oxidation or pyrolysis). Bead resins have some properties (e.g. particle size and density) that can have negative impacts on following waste treatment processes. Negative impacts could be: Floatation of bead resins in cementation process; Sedimentation in pipeline during transportation; Poor compaction properties for Hot Resin Supercompaction (HRSC). Reducing the particle size of the bead resins can have beneficial effects enhancing further treatment processes and overcoming prior mentioned effects. Westinghouse Electric Company has developed a modular grinding process to crush/grind the bead resins. This modular process is designed for flexible use and enables a selective adjustment of particle size to tailor the grinding system to the customer needs. The system can be equipped with a crusher integrated in the process tank and if necessary a colloid mill. The crusher reduces the bead resins particle size and converts the bead resins to a pump able suspension with lower sedimentation properties. With the colloid mill the resins can be ground to a powder. Compared to existing grinding systems this equipment is designed to minimize radiation exposure of the worker during operation and maintenance. Using the crushed and/or ground bead resins has several beneficial effects like facilitating cementation process and recipe development, enhancing oxidation of resins, improving the Hot Resin Supercompaction volume reduction performance. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Nanoscale TiO 2 coating improves water stability of Cs 2 SnCl 6

To improve the stability of Cs 2 SnCl 6 under aqueous/moisture environments, we applied a concept of artificial passivation by depositing a protective TiO 2 coating of 10 nm on the surface of Cs 2 SnCl 6 . Static leaching experiments results indicate that the initial release rates of Cs + and Cl - are decreased by 20–30 times with TiO 2 coating, suggesting its possibility to improve the short-term water/environmental stability of Cs 2 SnCl 6 . An amorphous-to-crystalline phase transition in TiO 2 film was observed, possibly resulting in degradation of Cs 2 SnCl 6 . However, the crystalline TiO 2 film still remains after 21 days water exposure and can still act as an effective passivation layer to reduce the release rates of Cs + and Cl - by as much as about 17 and 7 times, respectively, relative to static leaching without artificial coatings. Therefore, the water/environmental stability of metal halide perovskite Cs 2 SnCl 6 , which is a highly soluble molecular salt, can be enhanced by the nanoscale TiO 2 coating as an artificial passivation film.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Progress toward bridging from atomistic to continuum modeling to predict nuclear waste glass dissolution.

This report summarizes research performed for the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Subcontinuum and Upscaling Task. The work conducted focused on developing a roadmap to include molecular scale, mechanistic information in continuum-scale models of nuclear waste glass dissolution. This information is derived from molecular-scale modeling efforts that are validated through comparison with experimental data. In addition to developing a master plan to incorporate a subcontinuum mechanistic understanding of glass dissolution into continuum models, methods were developed to generate constitutive dissolution rate expressions from quantum calculations, force field models were selected to generate multicomponent glass structures and gel layers, classical molecular modeling was used to study diffusion through nanopores analogous to those in the interfacial gel layer, and a micro-continuum model (K{mu}C) was developed to study coupled diffusion and reaction at the glass-gel-solution interface.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Technetium and Iodine Getters to Improve Cast Stone Performance

To determine the effectiveness of the various getter materials prior to their solidification in Cast Stone, a series of batch sorption experiments was performed at Pacific Northwest National Laboratory. To quantify the effectiveness of the removal of Tc(VII) and I(I) from solution by getters, the distribution coefficient, Kd (mL/g), was calculated. Testing involved placing getter material in contact with spiked waste solutions at a 1:100 solid-to-solution ratio for periods up to 45 days with periodic solution sampling. One Tc getter was also tested at a 1:10 solid-to-solution ratio. Two different solution media, 18.2 MΩ deionized water (DI H2O) and a 7.8 M Na LAW simulant, were used in the batch sorption tests. Each test was conducted at room temperature in an anoxic chamber containing N2 with a small amount of H2 (0.7%) to maintain anoxic conditions. Each getter-solution combination was run in duplicate. Three Tc- and I-doping concentrations were used separately in aliquots of both the 18.2 MΩ DI H2O and a 7.8 M Na LAW waste simulant. The 1× concentration was developed based on Hanford Tank Waste Operations Simulator (HTWOS) model runs to support the River Protection Project System Plan Revision 6. The other two concentrations were 5× and 10× of the HTWOS values. The Tc and I tests were run separately (i.e., the solutions did not contain both solutes). Sampling of the solid-solution mixtures occurred nominally after 0.2, 1, 3, 6, 9, 12, 15 days and ~35 to 45 days. Seven getter materials were tested for Tc and five materials were tested for I. The seven Tc getters were blast furnace slag 1 (BFS1) (northwest source), BFS2 (southeast source), Sn(II)-treated apatite, Sn(II) chloride, nano tin phosphate, KMS (a potassium-metal-sulfide), and tin hydroxapatite. The five iodine getters were layered bismuth hydroxide (LBH), argentite mineral, synthetic argentite, silver-treated carbon, and silver-treated zeolite. The Tc Kd values measured from experiments conducted using the 7.8 M Na LAW simulant (the simulant selected to represent LAW) for the first 15 days for four Tc getters (BFS1, BFS2, Sn(II)-treated apatite, and Sn(II) chloride) show no, to a very small, capacity to remove Tc from the LAW simulant. For the Tc-getter experiments in the 7.8 M LAW simulant, the majority of the effluent samples show very small drops in Tc concentrations for the 35-day compared to the 15-day samplings. However, the Tc concentration in the simulant blanks also dropped slightly during this period, so the effect of the getter contacting LAW simulant at 35 days compared to 15 days is minimal; except that the BFS1 1:10 test shows a slow but steady decrease in Tc concentration in the LAW simulant supernatant from the beginning to the 35 day contact at which point about 20% of the original Tc has been removed from solution. Lastly, the KMS getter gives the highest Kd value for Tc at 35 days where Kd values have increased to 104 mL/g. When considering the different I getters reacting with the 7.8 M LAW simulant, two getters are much more effective than the others: Ag zeolite and Syn Arg. The other getters have calculated iodide distribution coefficients that show very limited effectiveness in the caustic conditions created by the LAW simulant. These are preliminary results that will need more detailed analyses including both pre- and post-batch sorption getter solid-phase characterization using state-of-the-art instrumentation such as synchrotron X ray absorption spectroscopy, which can delineate the oxidation state of the Tc and likely iodine species as well as some of the getters key major components, sulfur and iron in the BFS, and tin and sulfur in the tin-bearing and sulfur-bearing getters. This report also describes future experimental studies to be performed to better elucidate the mechanisms controlling the Tc and I sequestration processes in the various getters and leach tests of getter-bearing Cast Stone monoliths.

Tc getters, I getters, Cast Stone, Waste Forms↗

Prioritized List of Research Needs to support MRWFD Case Study Flowsheet Advancement

In FY-13, a case study evaluation was performed of full recycle technologies for both the processing of light-water reactor (LWR) used nuclear fuels as well as fast reactor (FR) fuel in the full recycle option. This effort focused on the identification of the case study processes and the initial preparation of material balance flowsheets for the identified technologies. In identifying the case study flowsheets, it was decided that two cases would be developed: one which identifies the flowsheet as currently developed and another near-term target flowsheet which identifies the flowsheet as envisioned within two years, pending the results of ongoing research. The case study focus is on homogeneous aqueous recycle of the U/TRU resulting from the processing of LWR fuel as feed for metal fuel fabrication. The metal fuel is utilized in a sodium-cooled fast reactor, and the used fast reactor fuel is processed using electrochemical separations. The recovered U/TRU from electrochemical separations is recycled to fuel fabrication and the fast reactor. Waste streams from the aqueous and electrochemical processing are treated and prepared for disposition. Off-gas from the separations and waste processing are also treated. As part of the FY-13 effort, preliminary process unknowns and research needs to advance the near-term target flowsheets were identified. In FY-14, these research needs were updated, expanded and prioritized. This report again updates the prioritized list of research needs based upon results to date in FY-15. The research needs are listed for each of the main portions of the flowsheet: 1) Aqueous headend, 2) Headend tritium pretreatment off-gas, 3) Aqueous U/Pu/Np recovery, 4) Aqueous TRU product solidification, 5) Aqueous actinide/lanthanide separation, 6) Aqueous off-gas treatment, 7) Aqueous HLW management, 8) Treatment of aqueous process wastes, 9) E-chem actinide separations, 10) E-chem off-gas, 11) E-chem HLW management. The identified research needs were prioritized within each of these areas. No effort was made to perform an overall prioritization. This information will be used by the MRWFD Campaign leadership in research planning for FY-16. Additionally, this information will be incorporated into the next version of the Case Study Report scheduled to be issued September 2015.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

OPC Paste Samples Exposed To Aggressive Solutions. Cementitious Barriers Partnership

The study presented in this report focused on a low-activity wasteform containing a high-pH pore solution with a significant level of sulfate. The purpose of the study was to improve understanding of the complex concrete/wasteform reactive transport problem, in particular, the role of pH in sulfate attack. Paste samples prepared at three different water-to-cement ratios were tested. The mixtures were prepared with ASTM Type I cement, without additional admixtures. The samples were exposed to two different sodium sulfate contact solutions. The first solution was prepared at 0.15M Na 2 SO 4 . The second solution also incorporated 0.5M NaOH, to mimic the high pH conditions found in Saltstone. The data collected indicated that, in Na 2 SO 4 solution, damage occurs to the pastes. In the case of the high-pH sulfate solution (Na 2 SO 4 + NaOH), no signs of damage were observed on any of the paste mixtures. These results indicate that the high sulfate content found in the wasteform pore solution will not necessarily lead to severe damage to concrete. Good-quality mixtures could thus prove durable over the long term, and act as an effective barrier to prevent radionuclides from reaching the environment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Fuel Cycle Technologies 2016 Achievements Report

The U.S. Department of Energy (DOE), Office of Nuclear Energy (DOE- NE) established the Fuel Cycle Technologies (FCT) program to address challenges such as growing stockpiles of used nuclear fuel (UNF) and high-level waste (HLW), and an urgency to enhance accident tolerance of the existing reactor fleet. Resolution of these challenges involves developing systems that reduce waste while improving resource utilization and safety. Towards this goal, the FCT program implemented a research and development (R&D) science-based approach, integratesing theory and experiment with high performance computing modeling and simulation. The FCT program establishes and strengthens collaborations with experts at international agencies, universities, and industry to develop the nuclear fuel cycle of the future. DOE-NE national laboratories and their international counterparts provide extensive expertise and experience and access to critical R&D facilities. Universities provide a wellspring of innovative ideas while industry provides the venue to integrate the expertise, experience, and ideas into commercial practice.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Updated Medical Waste Management Plans and Emergency Action Plans for the Lawrence Livermore National Laboratory Main Site

Attached are the updated Medical Waste Management Plans and Emergency Action Plans for Lawrence Livermore National Laboratory’s (LLNL) Biosafety Level 2 (BSL-2) facilities and Biosafety Level 3 (BSL-3) facility. A redline version of each document has also been provided. Minor changes such as facility personnel, contact information, and titles for those that support onsite Biosafety Labs at LLNL have been made to the document. It should be noted that the most recent versions of the Medical Waste Large Quantity Generator Registration Application Form and Medical Waste Management Plan were incorporated and sourced from the Alameda County Department of Environmental Health website.

54 ENVIRONMENTAL SCIENCES↗