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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Decontamination and Melting of Reactor Upper Internals - 18536

In a BWR, the upper reactor internals, i.e. the steam separators and the steam dryer converts the hot water to a dry steam for the turbines. These components have due to the distance to the reactor core low to very low levels of neutron activation, if any. The contamination of these components is usually significant and consists of both loose contamination and contamination bound in oxide layers. The upper reactor internals on a BWR are quite voluminous in relation to their weight, has a complex geometry and are therefore time consuming to cut in pieces that allows high density packing in waste containers for ILW. The incentives for decontamination of the components down to levels that allow melting for volume reduction are therefore high. A performed full-scale treatment project on a steam dryer confirms the significant volume reduction. The steam dryer was decayed prior to segmentation and melting for volume reduction. Cyclife/EDF has, since significant volume reduction is within reach if the inventory of gamma emitting nuclides and by then the dose rate can be reduced, taken benefit of chemical decontamination studies and an analyze of potential treatment results. The analyze confirm that high decontamination factors can be achieved and that the acceptance criteria for melting can be met, with margin. Full scale decontamination results achieved by AREVA NP and others do align fairly well with the test results. Even without the option of clearance/recycling of the metals, due to high residual activity, a high or very high volume reduction factors can be achieved by melting as well as a potential reclassification of the metal in a disposal perspective. One key aspect for success is to keep the volume of the secondary waste from the decontamination down and to make sure that the waste form fulfils the requirements for disposal. (authors)

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

Nitrate and Nitrite at Hanford – From Tanks to Natural Attenuation

Washington River Protection Solutions, LLC (WRPS) contracted Pacific Northwest National Laboratory (PNNL) in support of their maturation testing and analysis to develop and deploy an ambient temperature solidification/stabilization process for low-activity waste (LAW) (referred to as “grout” in this report). The intent of this effort is to support a sample-and-send regulatory and processing strategy in the Hanford 200 West Area, specifically considering grout waste forms disposal at Hanford in the Integrated Disposal Facility (IDF). Currently the disposal of grouted LAW is not an option because the IDF is not permitted to receive this waste form. The material presented in this report is intended to help inform the U.S. Department of Energy (DOE), the site operating contractors, regulatory agencies, and stakeholders of the future implications of IDF disposal of a grouted LAW waste form, specifically for the behavior of nitrate (NO 3 - ) and nitrite (NO 2 - ) at the Hanford. The only calculation of nitrate release from a hypothetical grouted LAW inventory in the IDF showed that the overall release was slightly above the compliance limit of 45 mg/L nitrate and 3.32 mg/L nitrate after 1000 years (Asmussen et al. 2019). As such, uncertainty around the impact of nitrate and nitrite release from the IDF was identified in the recent National Academies of Sciences study of Hanford supplemental LAW by both the national laboratory team and Hanford stakeholders (Bates et al. 2023). However, these calculations (and others for the IDF involving secondary waste grout) may be conservative as no nitrate and nitrite retention in the grout waste form was assumed, nor any attenuation in the subsurface. There is evidence from the literature of processes that can attenuate nitrate migration both in the waste form and in the Hanford subsurface. If found to be technically defensible and likely to occur in the IDF and associated subsurface, then these processes should be included in modeling of the IDF to represent the behavior of nitrate and nitrite more accurately. Doing so would remove unnecessary conservatism in the modeling projections and reduce the uncertainty in assessments of facility compliance of a grouted LAW waste form in the IDF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Immobilization of Organic Liquid Radioactive Waste (LWR) in a Sulfur Composite Matrix at the BN-350 Fast Neutron Reactor Site, Kazakhstan - 18031

The immobilization treatment and final disposition of all forms of organic radioactive liquids, such as pump oil, spent solvent, tri-butyl-phosphate (TBP) and scintillation fluids, remain a significant challenge for nuclear waste generators and national regulators. The majority of nuclear nations have instituted regulations that require some form of encapsulation of treated organic liquid, notwithstanding LLW organic liquids that can be safely incinerated. In most cases cement is the designated material for encapsulation; however, cement can pose certain problems with the containment of treated organic liquids. In 2013, with funding from the U.S. Department of Energy - National Nuclear Security Administration through its Global Initiative for Proliferation Prevention Program (GIPP), a proven U.S. technology for the treatment of radioactive organic liquids was introduced at the BN-350 fast neutron reactor facility in Aktau, Kazakhstan. As the facility was undergoing full decommissioning, one hundred and sixty cubic meters of contaminated pump oil (including Cs-137, Sr-90 and U-235/238) contained in large above ground tanks had to be treated for final disposal. As a result of the research conducted over a two-year period, a new 2-step treatment process was invented by combining the U.S. solidification technology with a newly developed inorganic filler and molecular sulfur, referred to as radiation-resistant composite (RRC) to fully encapsulate the waste oil. The outcome of extensive test program indicates that the final form can reach up to 60 wt.% with a compressive strength of about 100 kg/cm{sup 2}. Preliminary conclusions indicate that RRC outperforms cement as an encapsulation form for final waste treatment which meets U.S. Nuclear Regulatory requirements for final disposal. This paper examines the preliminary experimental work that was conducted in Kazakhstan with the U.S. polymer technology and radioactive pump oil, and the subsequent secondary process of combining the solidified material into a homogeneous final form utilizing RRC. Topics covered in the paper include: - Kazakhstan's regulatory requirements for attaining an achievable waste form for acceptance, which led to the RRC invention; - Technical objectives for the application of polymer technology and RRC materials to radioactive liquid waste; - Polymer solidification with radioactive oil and the result; step one of the RRC process; - Description of RRC materials, their origin and chemical characteristics; - Process of combining the solidified polymer mass with RRC materials into a solid, homogeneous matrix; step two of the RRC process; - Experimental results including leaching data, waste loading estimates and heat process calculations to create the matrix; - Cost advantages of the RRC process over cementation; - Limitations of the RRC process; - Next steps to verify conformance of the RRC technology to the U.S. requirements. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Competitive TcO4-, IO3-, and CrO42- Incorporation into Ettringite

Ettringite is a naturally occurring mineral found in cementitious matrices that is known for its ability to incorporate environmentally mobile oxyanion contaminants. To better assess this immobilization mechanism for contaminants within cementitious waste forms intended for nuclear waste storage, this work explores how mixed oxyanion contaminants compete for ettringite incorporation and influence the evolving mineralogy. Ettringite was precipitated in the presence of TcO4-, IO3-, and/or CrO42-, known contaminants of concern to nuclear waste treatment, over pre-determined precipitation periods. Solution analyses quantified contaminant removal and the collected solid was characterized using bulk and microprobe XRD coupled with PDF and microprobe XRF analyses. Results suggest that =96% IO3- is removed from solution, regardless of ettringite precipitation time or the presence of TcO4- or CrO42-. However, TcO4- removal remained <20%, was not significantly improved with longer ettringite precipitation times, and significantly decreased in the presence of IO3-. When IO3- is comingled with CrO42, gypsum is formed as a secondary mineral phase, which results in oxyanion partitioning among mineral phases, e.g., IO3- incorporation into ettringite and CrO42- incorporation into gypsum via SO42- substitution. Results from this work exemplify the importance of competitive immobilization when assessing waste form performance and environmental risk of contaminant release.

Gillispie, Elizabeth C.↗
DOE OSTI2012

Crystalline Ceramic Waste Forms: Report Detailing Data Collection In Support Of Potential FY13 Pilot Scale Melter Test

The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be successfully produced from a melting and crystallization process. The objective of this report is to summarize the data collection in support of future melter demonstration testing for crystalline ceramic waste forms. The waste stream used as the basis for the development and testing is a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. The principal difficulties encountered during processing of the “reference ceramic” waste form by a melt and crystallization process were the incomplete incorporation of Cs into the hollandite phase and the presence of secondary Cs-Mo non-durable phases. In the single phase hollandite system, these issues were addressed in this study by refining the compositions to include Cr as a transition metal element and the use of Ti/TiO 2 buffer to maintain reducing conditions. Initial viscosity studies of ceramic waste forms indicated that the pour spout must be maintained above 1400°C to avoid flow blockages due to crystallization. In-situ electron irradiations simulate radiolysis effects indicated hollandite undergoes a crystalline to amorphous transition after a radiation dose of 10 13 Gy which corresponds to approximately 1000 years at anticipated doses (2 10 10 -2 10 11 Gy). Dual-beam ion irradiations employing light ion beam (such as 5 MeV alpha) and heavy ion beam (such as 100 keV Kr) studies indicate that reference ceramic waste forms are radiation tolerant to the α-particles and β-particles, but are susceptible to a crystalline to amorphous transition under recoil nuclei effects. A path forward for refining the processing steps needed to form the targeted phase assemblages is outlined in this report. Processing modifications including melting in a reducing atmosphere with the use of Ti/TiO 2 buffers, and the addition of Cr to the transition metal additives to facilitate Cs-incorporation in the hollandite phase. In addition to melt processing, alternative fabrication routes are being considered including Spark Plasma Sintering (SPS) and Hot Isostatic Pressing (HIP).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Polyacrylonitrile Composites of Ag–Al–Si–O Aerogels and Xerogels as Iodine and Iodide Sorbents

Mechanically robust engineered sorbents were constructed by embedding Ag-loaded aerogels or xerogels in a polyacrylonitrile (PAN) matrix at 30 or 50 mass% loadings of gel in PAN. Here, the process for making the base gels as well as the composite materials is described herein, a technology that is very flexible for making hybrid sorbents using various inorganic or organic active sorbents. The PAN-gel composite materials were evaluated for iodine uptake in the form of gaseous iodine [I 2(g) ] at room temperature and at 150°C as well as iodide (I – ) at room temperature from simulated liquid nuclear wastes. The composites show high I 2(g) uptake and high I – uptake (= 93 mass%) in simulated liquid secondary waste spiked with 5 mg L -1 I – . These results demonstrate the utility of such a composite form for iodine removal from contaminated gas and liquid streams.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Environmental Geochemistry and Acid Mine Drainage Evaluation of an Abandoned Coal Waste Pile at the Alborz-Sharghi Coal Washing Plant, NE Iran

In this paper, an abandoned waste coal pile, which is resulted from Alborz-Sharghi coal washing plant, NE of Iran was mineralogically and geochemically characterized to evaluate pyrite oxidation, acid mine drainage (AMD) generation, and trace element mobility. After digging ten trenches and vertical sampling, a quantitative method including the atomic absorption test, and the quality-based methods including optical study were carried out for determination of pyrite fractions in the waste pile. The geochemical results revealed that the fraction of remaining pyrite increased with depth, indicating that pyrite oxidation is limited to the shallower depths of the pile which were confirmed by variations of sulfate, pH, EC, and carbonate with depth of the pile. To evaluate the trend of trace elements and mineralogical constituents of the waste particles, the samples were analyzed by using XRD, ICP-MS, and ICP-OES methods. The results showed the secondary and neutralizing minerals comprising gypsum have been formed below the oxidation zone. Besides, positive values of net neutralization potential indicated that AMD generation has not taken in the waste pile. In addition, variations of trace elements with depth reveal that Pb and Zn exhibited increasing trends from pile surface toward the bottom sampling trenches while another of them such as Cu and Ni had decreasing trends with increasing depth of the waste pile.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Test Bed: Treatability of Hanford Tank Waste - 18501

In an effort to advance the Low Activity Waste Pretreatment System technology readiness level at the Hanford Site, the effectiveness of a critical technology element, sRF ion-exchange resin, was demonstrated in the selective removal of Cs-137 from 11.5 L of composited Hanford supernate tank waste. Prior to ion-exchange, solids filtration was performed on the tank waste composite. Between the time of filtrate collection and filtrate processing on the ion-exchange columns, additional solids had formed. These secondary solids resulted from the combination of supernates high in fluoride content with supernates high in phosphate content and the relatively short blend time. Samples were collected before, during and after the ion-exchange process for characterization of a full suite of constituents (inorganic, organic, and radionuclides) to aid in the classification of the cesium-removed waste for shipping, receiving, treatment, and disposal determinations. Following chemical analysis, the remaining 10.8 L (2.9 gal) of cesium-depleted waste was shipped to an off-site vendor for grout-encapsulation and eventual disposal at the Waste Control Specialists Federal Waste Disposal Facility in Andrews, Texas as low-level waste. (authors)

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

Demonstration of SuperLig{sup R} 605 and SuperLig{sup R} 644 in a Regeneration Flow Sheet to Remove Radioactive Strontium and Cesium from Seawater - 16236

Remediation techniques are needed for efficient removal of radioactive strontium and cesium from seawater. Molecular Recognition Technology, using selective SuperLig{sup R} resins, demonstrates a clean chemistry approach to solving this problem. Molecular Recognition Technology offers high levels of separation of chemically pure metal contaminants, whose subsequent disposal is compatible with simple waste forms (grout and glass). SuperLig{sup R} resins can be regenerated with common reagents and used many times, limited by a dose of 1 E+9 Rad, which assures low volumes of secondary solid waste. The metal selectivity of the SuperLig{sup R} resin ensures that the solid waste is of low residual activity. These properties of sustainability are illustrated in a project that demonstrated the ability of two SuperLig{sup R} resins to selectively remove Cs and Sr from a seawater mimic representative of the contaminated seawater in Fukushima, Dai'ichi, Japan harbor. Removal of radioactive Sr-90 and Cs-134/Cs-137 from seawater using conventional ion exchange or a complexing agent is complicated by the presence of large excesses of competing ions such as Na, K, Ca, and Mg, which can be co-extracted. These competing ions, if co-extracted, reduce the capacity of a single pass resin with the consequent increase in volume of spent resin or decrease in capacity of a regenerable resin. The successful demonstration reported in this study has shown the ability of two proprietary resins, SuperLig{sup R} 605 and SuperLig{sup R} 644, to remove Sr and Cs, respectively, with high selectivity from simulated solutions containing these metal ions together with major constituents of seawater at their normal concentrations. This effort was part of a demonstration project to investigate methods for removal of Sr-90 and Cs-134/Cs-137 from 160,000 cubic meters of seawater impounded at Fukushima, Dai'ichi, Japan harbor. SuperLig{sup R} resins can be regenerated by elution with small volumes of simple acid solutions. SuperLig{sup R} resins have a high tolerance to radiation, with their performance not being degraded by doses of 1.0 E+9 Rad. Estimates, based on the results obtained, are that Cs and Sr levels in the volume of seawater in the Fukushima, Dai'ichi, Japan harbor could be reduced to desired discharge limits in 200 days with some 30 cubic meters of the two SuperLig{sup R} resins and that decontamination factors of 10 or more could be achieved. Results of the demonstration, together with advantages of the Molecular Recognition Technology process over conventional procedures for these separations are presented and discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Process optimization of caustic scrubber and iodine-129 immobilization in sodalite-based waste forms

Caustic scrubbers are proposed to remove vaporized iodine from nuclear fuel reprocessing plants. Resulting caustic slurries of radioactive iodine ( 129 I) and other anions can be captured and immobilized into the cage structures of sodalite-type minerals. Here, mixed anion sodalites were hydrothermally synthesized from a simulated caustic scrubber solution containing hydroxide/water, carbonate, chlorine, bromine, iodine, nitrate, and nitrite ions with added kaolinite. Experiments were conducted with twenty-five different sets of process conditions. Resulting powders were characterized by quantitative X-ray diffraction (XRD) and multi-functional thermal analysis, including gravimetry, calorimetry, and mass spectrometry. Using statistical analysis, process variables were tailored to maximize conversion of kaolinite into cage silicates, immobilize caustic scrubber anions, and minimize amorphous content. Further, chemical analysis showed the incorporation of all the targeted ions into the solids. Detailed investigation of a batch characterized after washing (creating secondary waste) and one without washing but subsequent calcination is given. This work contributes to defining the constraints required for subsequent consolidation process steps of the waste form to avoid volatilization.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Potential of hybrid functionalized meso-porous materials for the separation and immobilization of radionuclides

Functionalized meso-porous materials are a class of hybrid organic-inorganic material in which a meso-porous metal oxide framework is functionalized with multifunctional organic molecules. These molecules may contain one or more anchor groups that form strong bonds to the pore surfaces of the metal oxide framework and free functional groups that can impart and or modify the functionality of the material such as for binding metal ions in solution. Such materials have been extensively studied over the past decade and are of particular interest in absorption applications because of the tremendous versatility in choosing the composition and architecture of the metal oxide framework and the nature of the functional organic molecule as well as the efficient mass transfer that can occur through a well-designed hierarchically porous network. A sorbent for nuclear applications would have to be highly selective for particular radio nuclides, it would need to be hydrolytically and radiolytically stable, and it would have to possess reasonable capacity and fast kinetics. The sorbent would also have to be available in a form suitable for use in a column. Finally, it would also be desirable if once saturated with radio nuclides, the sorbent could be recycled or converted directly into a ceramic or glass waste form suitable for direct repository disposal or even converted directly into a material that could be used as a transmutation target. Such a cradle-to- grave strategy could have many benefits in so far as process efficiency and the generation of secondary wastes are concerned.This paper will provide an overview of work done on all of the above mentioned aspects of the development of functionalized meso-porous adsorbent materials for the selective separation of lanthanides and actinides and discuss the prospects for future implementation of a cradle-to-grave strategy with such materials. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Closed Fuel Cycle Waste Treatment Strategy

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) as a general strategy and associated Waste Treatment Baseline Study (WTBS). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to both a nominal case study and near-term target reprocessing flowsheets. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include: • Tritium separated from either a low volume gas stream or a high volume water stream. The recommended waste form is low-water cement in high integrity containers (HICs). • Iodine-129 separated from off-gas streams in aqueous processing. There are a range of potentially suitable waste forms. As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals. • Carbon-14 separated from LWR fuel treatment off-gases and immobilized as a CaCO3 in a cement waste form. • Krypton-85 separated from LWR and SFR fuel treatment off-gases and stored as a compressed gas. • An aqueous reprocessing high-level waste (HLW) raffinate waste which is immobilized by the vitrification process in one of three forms: a single phase borosilicate glass, a borosilicate based glass ceramic, or a multi-phased crystalline synthetic rock (Synroc). • An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel that is either included in the borosilicate HLW glass or is immobilized in the form of a metal alloy in the case of glass ceramics or Synroc. • Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware that are washed and super-compacted for disposal or as an alternative with high promise the purification and reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations. • Electrochemical process salt HLW which is incorporated into a glass bonded Sodalite waste form known as the ceramic waste form (CWF). • Electrochemical process UDS and SS cladding hulls which are melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from the reference mass balances are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1992

The Tharsis Montes, Mars - Comparison of volcanic and modified landforms

The three Tharsis Montes shield volcanos, Arsia Mons, Pavonis Mons, and Ascraeus Mons, have broad similarities that have been recognized since the Mariner 9 reconnaissance in 1972. Upon closer examination the volcanos are seen to have significant differences that are due to individual volcanic histories. All three volcanos exhibit the following characteristics: gentle (less than 5 deg) flank slopes, entrants in the northwestern and southeastern flanks that were the source for lavas extending away from each shield, summit caldera(s), and enigmatic lobe-shaped features extending over the plains to the west of each volcano. The three volcanos display different degrees of circumferential graben and trough development in the summit regions, complexity of preserved caldera collapse events, secondary summit-region volcanic construction, and erosion on the lower western flanks due to mass wasting and the processes that formed the large lobe-shaped features. All three lobe-shaped features start at elevations of 10 to 11 km and terminate at 6 km. The complex morphology of the lobe deposits appear to involve some form of catastrophic mass movement followed by effusive and perhaps pyroclastic volcanism.

Zimbelman, James R.↗
DOE OSTI2020

Estimation of Sorption Behavior of Europium to Calcium Silicate Hydrate by Thermal Analysis - 20172

High-level radioactive waste generated from reprocessing processes will be disposed in a geological disposal that is deep underground. Calcium silicate hydrate is formed as a secondary mineral in the repository via the reaction of calcium ions leaching from the cement and the silicic acid dissolved from the host rocks. In this study, the sorption behavior of europium, which is also known as a chemical analog of americium, on calcium silicate hydrate was examined. The solid phase was analyzed via thermogravimetry- differential thermal analysis, differential scanning calorimetry, Raman spectroscopy, fluorescence spectroscopy, and fluorescence lifetime analysis. To understand the behavior of calcium silicate hydrate under conditions saturated with groundwater, the study samples were analyzed using the above methods, except for deferential scanning calorimetry, without a drying process. The results of the differential scanning calorimetry analyses showed that the temperatures for the endothermic peaks of calcium silicate hydrate for the lower Ca/Si ratios were larger than those for the higher Ca/Si ratios for coexistent conditions of europium ions. Hence, the relationships of the temperatures of the endothermic peaks and the Ca/Si ratios were reversed by the addition of europium ions. These study results suggest that calcium silicate hydrate can immobilize europium ions, thus preventing the production of colloidal forms. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

The Effects of Secondary Mineral Precipitates on 90Sr Mobility at the Hanford Site, USA

The effects of secondary precipitates on 90Sr transport at the Hanford Site were investigated using quartz column experiments with simulated caustic tank waste leachates (STWL). Significantly enhanced retardation of Sr transport was observed in the column contacted with STWL due to Sr sorption and co-precipitation with neo-formed nitratecancrinite. However, the column results also suggest that neo-formed secondary precipitates could behave like native mobile colloids that can enhance Sr transport. Initially immobilized Sr within secondary precipitates could remobilize given a change in the porewater background conditions. The mobility of the neo-formed Sr-bearing precipitates increased with increased solution flow rate. In the field, porewater contents and flow rates can be changed by snowmelt (or storm water) events or artificial infiltration. The increased porewater flow rate caused by these events could affect the mobility of 90Sr-containing secondary precipitates, which can be a potential source for facilitated Sr transport in Hanford Site subsurface environments.

Effects Secondary Mineral Precipitates Mobility Ha↗
DOE OSTI2019

Effect of solution chemistry on the iodine release from iodoapatite in aqueous environments

To ensure the safe disposal of nuclear waste, understanding the release process of radionuclides retained in the nuclear waste forms is of vital importance. Iodoapatite Pb 9.85 (VO 4 ) 6 I 1.7 , a potential waste form for iodine-129, was selected as a model system for ceramic waste forms in this study to understand the effect of aqueous species on iodine release. Semi-dynamic leaching tests were conducted on bulk samples in cap-sealed Teflon vessels with 0.1 mol/L NaCl, Na 2 CO 3 , Na 3 PO 4 , and Na 2 SO 4 solutions under 90 °C, fixed sample surface area to solution volume ratio of 5/m, and periodic replacement of leaching solutions. The reacted solutions were then analyzed by Inductively Coupled Plasma-Mass Spectrometry and Inductively Coupled Plasma-Optical Emission Spectrometry; the leached surfaces were characterized by X-ray diffraction, scanning electron microscopy, and infrared spectroscopy. The finding shows that, compared to deionized water, the ion-rich solutions enhanced the iodine release as a result of the increased ionic strength, reduced activity coefficient of dissolved species, and increased solution pH. Surface reactions can lead to the formations of secondary phases by ion-exchange and precipitation. These findings indicate that an ion-rich environment in the geological repository can be detrimental to the disposal safety of the nuclear waste form.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Closed Fuel Cycle Waste Treatment Strategy

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗