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DOE OSTI2012

Small-Scale Melter Testing With Law Simulants To Assess The Impact Of Higher Temperature Melter Operations

About 50 million gallons of high-level mixed waste is currently in storage in underground tanks at The United States Department of Energy's (DOE's) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE's Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal. The tank waste will be separated into low- and high-activity fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High Level Waste (IHLW) products. The ILAW product will be disposed of in an engineered facility on the Hanford site while the IHL W product will be directed to the national deep geological disposal facility for high-level nuclear waste. The ILAW and IHLW products must meet a variety of requirements with respect to protection of the environment before they can be accepted for disposal.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

2011 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada National Security Site, Nye County, Nevada: Review of the Performance Assessments and Composite Analyses

The Maintenance Plan for the Performance Assessments and Composite Analyses for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site (National Security Technologies, LLC, 2007a) requires an annual review to assess the adequacy of the Performance Assessments (PAs) and Composite Analyses (CAs), with the results submitted annually to U.S. Department of Energy (DOE) Office of Environmental Management. The Disposal Authorization Statements for the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) also require that such reviews be made and that secondary or minor unresolved issues be tracked and addressed as part of the maintenance plan (DOE, 1999a; 2000).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Experimental Results for SimFuels

Assessing the performance of Spent (or Used) Nuclear Fuel (UNF) in geological repository requires quantification of time-dependent phenomena that may influence its behavior on a time-scale up to millions of years. A high-level waste repository environment will be a dynamic redox system because of the time-dependent generation of radiolytic oxidants and reductants and the corrosion of Fe-bearing canister materials. One major difference between used fuel and natural analogues, including unirradiated UO 2 , is the intense radiolytic field. The radiation emitted by used fuel can produce radiolysis products in the presence of water vapor or a thin-film of water that may increase the waste form degradation rate and change radionuclide behavior. To study UNF, we have been working on producing synthetic UO 2 ceramics, or SimFuels that can be used in testing and which will contain specific radionuclides or non-radioactive analogs so that we can test the impact of radiolysis on fuel corrosion without using actual spent fuel. Although, testing actual UNF would be ideal for understanding the long term behavior of UNF, it requires the use of hot cells and is extremely expensive. In this report, we discuss, factors influencing the preparation of SimFuels and the requirements for dopants to mimic the behavior of UNF. We have developed a reliable procedure for producing large grain UO 2 at moderate temperatures. This process will be applied to a series of different formulations.

Geological Disposal↗
DOE OSTI2012

Radiolysis Process Model

Assessing the performance of spent (used) nuclear fuel in geological repository requires quantification of time-dependent phenomena that may influence its behavior on a time-scale up to millions of years. A high-level waste repository environment will be a dynamic redox system because of the time-dependent generation of radiolytic oxidants and reductants and the corrosion of Fe-bearing canister materials. One major difference between used fuel and natural analogues, including unirradiated UO 2 , is the intense radiolytic field. The radiation emitted by used fuel can produce radiolysis products in the presence of water vapor or a thin-film of water (including OH• and H• radicals, O 2 -, eaq, H 2 O 2 , H 2 , and O 2 ) that may increase the waste form degradation rate and change radionuclide behavior. H 2 O 2 is the dominant oxidant for spent nuclear fuel in an O 2 depleted water environment, the most sensitive parameters have been identified with respect to predictions of a radiolysis model under typical conditions. As compared with the full model with about 100 reactions it was found that only 30-40 of the reactions are required to determine [H 2 O 2 ] to one part in 10–5 and to preserve most of the predictions for major species. This allows a systematic approach for model simplification and offers guidance in designing experiments for validation.

Hydrogen Peroxide↗
DOE OSTI2013

Technetium Removal Using Tc-Goethite Coprecipitation

This report describes the results from laboratory tests performed at Pacific Northwest National Laboratory for the U.S. Department of Energy (DOE) EM-31 Support Program (EMSP) subtask, “Low temperature waste forms coupled with technetium removal using an alternative immobilization process such as Fe(II) treated-goethite precipitation” to increase our understanding of 99Tc long-term stability in goethite mineral form and the process that controls the 99Tc(VII) reduction and removal by the final Fe (oxy)hydroxide forms. The overall objectives of this task were to 1) evaluate the transformation process of Fe (oxy)hydroxide solids to the more crystalline goethite (α-FeOOH) mineral for 99 Tc removal and 2) determine the mechanism that limits 99 Tc(IV) reoxidation in Fe(II)-treated 99 Tc-goethite mineral and 3) evaluate whether there is a long-term 99 Tcoxidation state change for Tc sequestered in the iron solids.

Goethite↗
DOE OSTI2013

Demonstrate the removal efficiency and capacity of MOF materials for krypton recovery

Metal organic framework materials (MOFs) were developed and tested in support of the U.S. Department of Energy Office of Nuclear Energy, Fuel Cycle Technology Separations and Waste Forms Campaign. Specifically, materials are being developed for the removal of xenon (Xe) and krypton (Kr) from gaseous products of nuclear fuel reprocessing unit operations. Two metal organic framework structures were investigated in greater detail to demonstrate the removal efficiency and capacity of MOF materials for krypton recovery. Our two bed breakthrough measurements on NiDOBDC and FMOFCu indicate these materials can capture and separate parts per million levels of Xe and Kr from air. The removal efficiency and adsorption capacity for Kr on these two MOFs were further increased upon removal of Xe upfront.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Initial proof-of-principle for near room temperature Xe and Kr separation from air with MOFs

Materials were developed and tested in support of the U.S. Department of Energy, Office of Nuclear Energy, Fuel Cycle Technology Separations and Waste Forms Campaign. Specifically, materials are being developed for the removal of Xenon and krypton from gaseous products of nuclear fuel reprocessing unit operations. During FY 2012, Three Metal organic framework (MOF) structures were investigated in greater detail for the removal and storage of Xe and Kr from air at room temperature. Our breakthrough measurements on Nickel based MOF could capture and separate parts per million levels of Xe from Air (40 ppm Kr, 78% N2, 21% O2, 0.9% Ar, 0.03% CO2). Similarly, the selectivity can be changed from Xe > Kr to Xe < Kr simply by changing the temperature in another MOF. Also for the first time we estimated the cost of the metal organic frameworks in bulk.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Composition Driven, Statistical Review of Cast Stone Screening Matrix Results

Recent studies in support of Cast Stone development for the Hanford Site were designed to screen the effects of mix components on properties and performance of the waste form. Dry blend components, including portland cement, blast furnace slag, and Class F fly ash, sourced from both southeastern and northwestern suppliers, were included in the screening matrix. Multiple decontaminated salt solutions (simulated) were combined with the dry blend components at various liquid to solid ratios following a statistically designed matrix of compositions. Properties of the resulting mixes and performance of the cured materials were then measured and related to the dry blend components and salt solutions that were used for each mix. At the conclusion of the screening matrix work, SRNL felt that additional insight into the study might be obtained by relating the properties and performance of the materials to the composition of each Cast Stone mix, as opposed to just the dry blend – salt solution combination. This memorandum provides the results of a statistically driven review of the screening matrix data as a function of the composition of each mix. It is meant to augment the results obtained by reviewing the screening matrix data as a function of the dry blend and salt solution combination.

Edwards, T. B.↗
DOE OSTI2014

Expanded Analysis of Hot Isostatic Pressed Iodine-Loaded Silver-Exchanged Mordenite

Reduced silver-exchanged mordenite (Ag0Z) is being evaluated as a potential material to control the release of radioactive iodine that is released during the reprocessing of used nuclear fuel into the plant off-gas streams. The purpose of this study was to determine if hot pressing could directly convert this iodine loaded sorbent into a waste form suitable for long-term disposition. The minimal pretreatment required for production of pressed pellets makes hot pressing a technically and economically desirable process. Initial scoping studies utilized hot uniaxial pressing (HUPing) to prepare samples of non-iodine-loaded reduced silver exchanged mordenite (Ag0Z). The resulting samples were very fragile due to the low pressure (~ 28 MPa) used. It was recommended that hot isostatic pressing (HIPing), performed at higher temperatures and pressures, be investigated. HIPing was carried out in two phases, with a third and final phase currently underway. Phase I evaluated the effects of pressure and temperature conditions on the manufacture of a pressed sample. The base material was an engineered form of silver zeolite. Six samples of Ag0Z and two samples of I-Ag0Z were pressed. It was found that HIPing produced a pressed pellet of high density. Analysis of each pressed pellet by scanning electron microscopy-energy dispersive spectrophotometry (SEM-EDS) and X-ray diffraction (XRD) demonstrated that under the conditions used for pressing, the majority of the material transforms into an amorphous structure. The only crystalline phase observed in the pressed Ag0Z material was SiO2. For the samples loaded with iodine (I-Ag0Z) iodine was present as AgI clusters at low temperatures, and transformed into AgIO4 at high temperatures. Surface mapping and EDS demonstrate segregation between silver iodide phases and silicon dioxide phases. Based on the results of the Phase I study, an expanded test matrix was developed to examine the effects of multiple source materials, compositional variations, and an expanded temperature range. Each sample was analyzed with the approach used in Phase I. In all cases, there is nothing in the SEM or XRD analyses that indicates creation of any AgI-containing silicon phase, with the samples being found to be largely amorphous. Phase III of this study has been initiated and is the final phase of scoping tests. It will expand upon the test matrix completed in Phase II and will examine the durability of the pressed pellets through product consistency testing (PCT) studies. Transformation of the component material into a well-characterized iodine-containing mineral phase would be desirable. This would limit the additional experimental testing and modeling required to determine the long-term stability of the pressed pellet, as much of that information has already been learned for several common iodine-containing minerals. However, this is not an absolute requirement, especially if pellets produced by hot isostatic pressing can be demonstrated through initial PCT studies to retain iodine well despite their amorphous composition.

hot isostatic pressing↗
DOE OSTI2015

Complete Iodine Loading of NO-Aged Ag 0 -Functionalized Silica Aerogel (Milestone Report)

In off-gas treatment systems within a nuclear fuel reprocessing plant, capture materials will be exposed to a gas stream for extended periods during their lifetime. This gas stream may be at elevated temperature and could contain water, NO x gas, or a variety of other constituents. For this reason, it is important to understand the effects of long-term exposure, or aging, on proposed capture materials. One material under consideration for iodine sequestration is silver-functionalized silica aerogel (Ag 0 -aerogel). The aim of this study was to determine the effect of extended exposure at 150°C to an air stream containing NO on the iodine capture capacity of Ag 0 -aerogel. Ag 0 -aerogel was provided by the Pacific Northwest National Laboratory (PNNL), which manufactures the material at a lab scale. Prior to aging, the material has an iodine loading capacity of approximately 290 mg I/g Ag 0 -aerogel. Previous studies have aged the material in a dry air stream or in a moist air stream for up to 6 months. Both tests resulted in a 22% loss in iodine capacity. Aging the material in a static 2% NO 2 environment for up to 2 months results in a 15% loss of iodine capacity.3 In this study, exposure of Ag 0 -aerogel to 1% NO at 150°C for 2 months produced a loss of 43% in iodine loading capacity. This is largest loss observed for aerogel aging studies to date. The performance of Ag 0 -aerogel in this study was compared to the performance of reduced silver mordenite (Ag 0 Z) in similar studies. Ag 0 Z is a zeolite mineral considered to be the current standard technology for iodine removal from off-gas streams of a potential US used fuel processing plant. In an aging study exposing Ag 0 Z to 1% NO for 2 months, an iodine capacity loss of over 80% was observed. This corresponds to a silver utilization of 13.5% for 2 month NO-aged Ag 0 Z, compared to 57% silver utilization for 2 month NO-aged aerogel. While iodine loading capacity and silver utilization are critical parameters in evaluating these materials, other properties must also be considered when selecting the appropriate material (e.g., relative material densities and potential waste form production technology). The resistance of Ag 0 -aerogel to NO is promising, and investigations of this material for use in iodine capture should continue to be pursued.

36 MATERIALS SCIENCE↗
DOE OSTI2015

High Temperature Gas-cooled Reactor (HTGR) Graphite Pebble Fuel: Review of Technologies for Reprocessing

This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

High Temperature Gas-cooled Reactor (HTGR) Graphite Pebble Fuel: Review of Technologies for Reprocessing

This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding and chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Office of River Protection Advanced Low-Activity Waste Glass Research and Development Plan

The U.S. Department of Energy Office of River Protection (ORP) has initiated and leads an integrated Advanced Waste Glass (AWG) program to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product performance requirements. The integrated ORP program is focused on providing a technical, science-based foundation for making key decisions regarding the successful operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) facilities in the context of an optimized River Protection Project (RPP) flowsheet. The fundamental data stemming from this program will support development of advanced glass formulations, key product performance and process control models, and tactical processing strategies to ensure safe and successful operations for both the low-activity waste (LAW) and high-level waste vitrification facilities. These activities will be conducted with the objective of improving the overall RPP mission by enhancing flexibility and reducing cost and schedule.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Summary of ALSEP Test Loop Solvent Irradiation Testing

Separating the minor actinide elements (americium and curium) from the fission product lanthanides is an important step in closing the nuclear fuel cycle. Isolating the minor actinides will allow transmuting them to short lived or stable isotopes in fast reactors, thereby reducing the long-term hazard associated with these elements. The Actinide Lanthanide Separation Process (ALSEP) is being developed by the DOE-NE Material Recovery and Waste Form Development Campaign to accomplish this separation with a single process. To develop a fundamental understanding of the solvent degradation mechanisms for the ALSEP Process, testing was performed in the INL Radiolysis/Hydrolysis Test Loop for the extraction section of the ALSEP flowsheet. This work culminated in the completion of the level two milestone (M2FT-16IN030102021) "Complete ALSEP test loop solvent irradiation test.” This report summarizes the testing performed and the impact of radiation on the ALSEP Process performance as a function of dose.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Evaluation of ALSEPO

The level three milestone (M3FT-17IN030102013) "Evaluation of ALSEPO” has been completed. Researchers at INL are studying the impacts of gamma radiolysis on the Actinide Lanthanide Separation Process (ALSEP) flowsheet. The ALSEP process is being developed by the DOE-NE Material Recovery and Waste Form Development Campaign. The process solvent used a combination of the solvating extractant N,N,N’,N’-tetra(2-ethylhexyl)diglycolamide and the cation exchanging extractant 2-ethylhexylphosphonic acid mono-2-ethylehxyl ester dissolved in an aliphatic diluent. The actinides and lanthanides are co-extracted from nitric acid solution and the actinides are selectively stripped using a carboxylic acid buffered aminopolycarboxylic acid solution. The results irradiation studies previously performed at the INL on another diglycolamide-based solvent extraction system containing 1-octanol as a phase modifier suggested that the octanol may be serving to inhibit the radiolytic degradation of the ALSEP extractants. Although the mechanism of this radioprotection was not determined, the addition of octanol to the ALSEP solvent was proposed. The goal of this research program was to establish the mechanism responsible for the observed increase in the radiation hardiness of the ALSEP solvent upon the addition of 1-octanol. In addition, other possible enhancements of solvent behavior such as possible increased resistance to third phase formation and enhanced extraction/stripping kinetics were examined. The addition of 1-octanol to the ALSEP formulation at either 1 % or 5 % (v/v) does reduce the rate of degradation of the solvent when exposed to gamma radiation. Unfortunately, the addition of the polar phase modifier essentially destroyed the An/Ln separation usually achievable with ALSEP chemistry. This loss of An/Ln separation ability is presumably do to the disruption of HEH[EHP] dimer formation due to the addition of the polar alcohol. The addition of octanol does not appear to have any significant impact upon the rate of removal of either americium or europium radiotracer from the loaded organic solvents. Based upon the distribution ratio data obtained for the 1% octanol and 5 % octanol ALSEPO solvent systems and the lack of any enhancement in the rate of stripping of the actinides and lanthanides from the loaded solvent, further efforts to reformulate the ALSEPO solvent were abandoned.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Milestone report - M4FT-15OR03120210 - Multi-lab Analytical Plan for Analysis of Legacy 85 Kr Samples

Legacy samples composed of 85 Kr encapsulated in solid zeolite 5A material and five small metal tubes containing a mixture of the zeolite combined with a glass matrix resulting from hot isostatic pressing have been preserved. The samples were a result of krypton R&D encapsulation efforts in the late 1970s performed at the Idaho Chemical Processing Plant. These samples were shipped to Oak Ridge National Laboratory (ORNL) in mid-FY 2014. Upon receipt the outer shipping package was opened and the inner package removed and placed in a radiation hood. The individual capsules were double bagged as they were removed from the inner shipping pig and placed into individual glass sample bottles for further analysis. The five capsules were then x-ray imaged. Capsules 1 and 4 appear to contain an amorphous mass within the capsules. Capsule 2 clearly shows the saw marks on the capsule and a quantity of loose pellet or bead-like material remaining in the capsule. Capsule 3 shows similar bead-like material within the intact capsule. Capsule 5 was previously opened. The end of this capsule appears to have been cut off and there are additional saw marks on the side of the capsule. X-ray tomography allowed the capsules to be viewed along the three axes. Of most interest was determining whether there was any residual material in the closed end of Capsule 5. The images confirmed the presence of residual material within this capsule. The material appears to be compacted but still retains some of the bead-like morphology. A conference call was held on February 19, 2015, with participation from ORNL, Idaho National Laboratory, Sandia National Laboratories, and Pacific Northwest National Laboratory Sigma Team members. The nondestructive analysis (NDA) results were discussed, and a proposed path forward was formulated to advance this effort toward the original goals of understanding the effects of extended storage on the waste form and package. Based on the initial NDA and the fact that there are at least two breached samples it was proposed that exploratory tests be conducted with the breached specimens before opening the three intact capsules. Portions of these would be analyzed to determine the fraction of krypton/xenon remaining in the matrix and the amount of rubidium remaining in the matrix. The inner surface of the breached capsules would be examined for corrosion.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

NTRD Technical Monthly October FY18

Topics covered in this Nuclear Technology and Research Development (NTRD) monthly update include: Advanced Fuels Campaign, Material Recovery and Waste Forms Development, MPACT Campaign, Fuel Cycle Options Campaign, Joint Fuel Cycle Study Activities, and AFCI-HQ Program Support.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Preliminary Desorption studies for HZ-PAN and AgZ-PAN

Adsorption of Kr and Xe on HZ-PAN and AgZ-PAN has been the studied extensively. Prior research focused on sorbent capacity and breakthrough of Kr and Xe, with desorption cycles completed in order to close material balances and prepare sorbents for further adsorption testing. This series of tests was designed to determine preliminary information about desorption to guide further research to optimize desorption cycles for each sorbent. A series of desorption tests were run using thermal and pressure swing operations. For AgZ-PAN, these tests indicated that both heat and carrier gas are likely necessary to removed adsorbed gases. However, vacuum pressure and heat may be sufficient to desorb gases from HZ-PAN, which should result in a cleaner Kr stream during desorption. Desorption optimization for both sorbents merits further study in order to make final waste form volume as small as possible.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗