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We are purusing an understand of the durability and materials processability of the low temperature sintering Bi-Si oxide Glass Composite Material (GCM)1 Waste Form for iodine capture materials. The chemical and physical controls over iodine release from candidate 129 I waste forms must be quantified to predict long-term waste form effectiveness.
Perovskite-based ceramic composites were developed as potential waste form materials for immobilizing cesium (Cs) and iodine (I) with high waste loadings and chemical durability. The perovskite Cs 3 Bi 2 I 9 has high Cs (22 wt%) and I (58 wt%) content, and thus can be used as a potential host phase to immobilize these critical radionuclides. In this work, the perovskite Cs 3 Bi 2 I 9 phase was synthesized by a cost effective solution-based approach, and was embedded into a highly durable hydroxyapatite matrix by spark plasma sintering to form dense ceramic composite waste forms. The chemical durabilities of the monolithic Cs 3 Bi 2 I 9 and Cs 3 Bi 2 I 9 -hydroxyapatite composite pellets were investigated by static and semi-dynamic leaching tests, respectively. Cs and I are incongruently released from the matrix for both pure Cs 3 Bi 2 I 9 and composite structures. The normalized Cs release rate is faster than that of I, which can be explained by the difference in the strengths between Cs-I and Bi-I bonds as well as the formation of insoluble micrometer-sized BiOI precipitates. The activation energies of elemental releases based on dissolution and diffusion-controlled mechanisms are determined with significantly higher energy barriers for dissolution from the composite versus that of the monolithic Cs 3 Bi 2 I 9 . The ceramic-based composite waste forms exhibit excellent chemical durabilities and waste loadings, commensurate with the state-of-the-art glass-bonded perovskite composites for I and Cs immobilization.
This is the final report of the NEUP project “Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms.” This project started on July 1, 2012 and was successfully completed on June 30, 2016. This report provides an overview of the main achievements, results and findings through the duration of the project. Additional details can be found in the main body of this report and in the individual Quarterly Reports and associated Deliverables of this project, which have been uploaded in PICS-NE. The objective of this research was to advance understanding and develop validated models on the effects of self-radiation from beta and alpha decay on the response of used nuclear fuel and nuclear waste forms during high-temperature interim storage and long-term permanent disposition. To achieve this objective, model used-fuel materials and model waste form materials were identified, fabricated, and studied.
Radiation damage was simulated in multiphase titanate-based ceramic waste forms using an ion accelerator to generate high energy alpha particles (He + ) and an ion implanter to generate 7 MeV gold (Au 3+ ) particles. X-ray diffraction and transmission electron microscopy were used to characterize the damaged surfaces and nearby regions. Simulated multiphase ceramic waste forms were prepared using two processing methods: spark plasma sintering and melt-processing. Both processing methods produced ceramics with similar phase assemblages consisting of hollandite-, zirconolite/pyrochlore-, and perovskite- type phases. The measured heavy ion (Au 3+ ) penetration depth was less in spark plasma sintered samples than in melt-processed samples. Furthermore, structural breakdown of the hollandite phase occurred under He + irradiation indicated by the presence of x-ray diffraction peaks belonging to TiO 2 , BaTiO 5 , and other hollandite related phases (Ba 2 Ti 9 O 20 ). The composition of the constituent hollandite phase affected the extent of damage induced by Au 3+ ions.
It has been demonstrated that monazite-type materials are excellent candidates for nuclear waste forms, and hence, their facile synthesis is of great importance for the needed sequestration of existing nuclear waste. The synthesis of monazite, LaPO 4 , requires inconveniently high temperatures near 1000°C and generally involves the conversion of the presynthesized rhabdophane, LaPO 4 •nH 2 O, to the LaPO 4 monazite phase. During this structure transformation, the rhabdophane converts irreversibly to the thermodynamically stable monoclinic monazite structure. A low-temperature (185° to 260°C) mild hydrothermal acid-promoted synthesis of monazite is described that can both transform presynthesized rhabdophane or assemble reagents to the monoclinic monazite structure. The pH dependence of this reaction is detailed, and its applicability to the LnPO 4 (Ln = La, Ce, Pr, Nd, Sm-Gd), Ca 0.5 Th 0.5 PO 4 , and Sr 0.5 Th 0.5 PO 4 systems is discussed. The crystal growth of Ca 0.5 Th 0.5 PO 4 and Sr 0.5 Th 0.5 PO 4 is described, and their crystal structures were reported. In situ x-ray diffraction studies, performed as a function of temperature, provide insight into the structure transformation process.
Abstract As part of a broader study of ceramic nuclear waste‐forms, four different lanthanide titanates were fabricated; La 0.1 Sm 0.1 Gd 0.1 Tb 0.1 Dy 0.3 Ho 0.1 Er 0.2 YbTiO 5 , Sm 0.3 Gd 0.3 Dy 0.3 Yb 1.1 TiO 5 , Sm 0.1 Gd 0.4 Dy 0.4 Yb 1.1 TiO 5 , and Sm 0.2 Gd 0.2 Dy 0.2 Yb 1.4 TiO 5 . The aim was to produce single‐phase novel materials with cubic symmetry, capable of incorporating a wide variety of cations and with acceptable radiation tolerance. The chemistry flexibility and radiation tolerance are some of the major desirable properties for nuclear waste‐form materials. By using multiple lanthanides the average lanthanide radius can be controlled and consequently the structure, along with properties such as radiation tolerance. The radiation tolerance was assessed using in situ 1 MeV krypton irradiation and transmission electron microscopy characterization. Those materials for which cubic symmetry was achieved displayed better radiation tolerance; a greater critical fluence of ions ( F c ) was required for the crystalline to amorphous transition, and a lower temperature was required to maintain crystallinity ( T c ) during irradiation.
Herein, we describe a first-of-a-kind study where lead tellurite (Pb-Te-O), iron phosphate (Fe-P-O), and lanthanide aluminoborosilicate (LABS) waste forms are evaluated for treating and/or immobilizing waste streams containing high rare-earth (RE) content, e.g., RECl 3 , REOCl, and/or REO x . Results from this study show that NdOCl can be immobilized in 78TeO 2 -22PbO glass at up to 10 mass%. Higher loadings of up to 15 mass% were demonstrated, but with a small fraction of crystallization. Two Fe-P-O glasses produced from either RECl 3 (RE = Nd,Ce) or NdOCl revealed products with very similar properties, demonstrating that the phosphate process can be used to chlorinate a variety of RE-containing waste streams. Two LABS glasses produced from Nd 2 O 3 or NdOCl showed similar properties, but the LABS-NdOCl glass did have some residual Cl present. Characterization is provided for all waste forms including chemical analysis, X-ray diffraction, and bulk density.
The attached report documents sample preparation and x-ray diffraction results for a series of cement and blended cement matrices prepared with either water or a 4.4 M Na salt solution. The objective of the study was to provide initial phase characterization for the Cementitious Barriers Partnership reference case cementitious salt waste form. This information can be used to: 1) generate a base line for the evolution of the waste form as a function of time and conditions, 2) potentially to design new binders based on mineralogy of the binder, 3) understand and predict anion and cation leaching behavior of contaminants of concern, and 4) predict performance of the waste forms for which phase solubility and thermodynamic data are available.
This study explores the corrosion interactions between a metallic canister material, stainless steel (SS) 316, and an I-bearing ceramic waste form, lead vanado-iodoapatite (I-APT, Pb 9.85 (VO 4 ) 6 I 1.7 ), in a chloride solution. Crevice corrosion of the SS in close proximity to the I-APT resulted in the development of an aggressive environment at the interface of the two materials, which was acidic and enriched in Cl – anions. I-APT also corroded in the crevice region, primarily through ion-exchange between the I – ions from the I-APT matrix and anions from the environment. The enrichment of Cl – anions within the occluded crevice space as the result of SS crevice corrosion enhanced the corrosion of I-APT. The release of iodine from this apatite waste form could be accelerated owing to this mechanism. This is evidenced by a depletion of iodine from the I-APT matrix and a large amount of Cl-bearing precipitates on the surfaces of both SS and I-APT. On the other hand, the corrosion of I-APT leads to the precipitation of a V- and Pb-rich layer, which inhibits the localized corrosion of SS to an extent. This study advances the understanding of the near-field corrosion interactions between metallic canisters and ceramic waste forms.
Krypton-85 (Kr-85) is a volatile radionuclide released during the reprocessing of used nuclear fuel (UNF). Prior to discharge of process off-gas streams to the environment, removal of Kr-85 is required to meet emission regulations. Once removed from the off-gas streams, Kr (containing stable Kr in addition to Kr-85) must be placed in storage until the Kr-85 has decayed to levels such that its potential release would meet regulatory guidelines. The current method of storage is high-pressure metal canisters. Historically, concerns have been expressed that the decay product of Kr-85, Rubidium-85 (Rb-85), could cause corrosion and weakening in the canisters that would ultimately lead to premature failure and release of Kr-85. Alternate methods and materials, such as matrix isolation in a vapor deposited metal and immobilization in zeolites, have been investigated as storage approaches for Kr. However, even these storage methods may become compromised due to the chemical effects of Rb-85 arising from the decay of Kr-85. A review was conducted of the available literature to examine topics relevant to the effects of Rb on the long-term storage of Kr-85, examine the source terms in the context of this concern, and suggest a path forward to resolve outstanding knowledge gaps. In the course of this analysis, several knowledge gaps were identified. Examples include: o The use of solid sorbents to potentially permit more Kr-85-bearing gas to be loaded into canisters, either at the pressures used in the current method or at significantly lower pressures. o A cost-to-benefit analysis is needed to ascertain the practicality of manufacturing waste forms for Kr-85 storage with solid sorbents. o The effects of radiation and chemical transmutation in solid sorbents or matrices for this application are virtually unknown. o Fundamental data are needed on corrosion rates and mechanisms as functions of Rb concentration, storage temperature, etc., for various storage approaches (e.g., as compressed gas or encapsulated in a getter material) and storage container materials. The present analysis focused on the available information regarding storage of Kr-85 in high-pressure containers. It was judged unlikely that such containers would be compromised by corrosion from the Rb decay product, especially if the Kr recovery operations result in a gas composition with very low oxygen content. However, this conclusion should be revisited based on preliminary data gathered from analysis of legacy Kr waste forms. These legacy waste forms are a result of experiments conducted at Idaho National Laboratory over 30 y ago and have been stored for more than three Kr-85 half-lives. Evaluation of these forms indicates significant corrosion, the cause of which has not been conclusively determined. (authors)
In 2010, five samples were reclaimed at the Idaho National Laboratory (INL). These samples were generated by R&D performed in the 1970’s relating to the long-term storage of 85 Kr and presented a unique opportunity to study a potential 85 Kr waste form after over three half-lives have elapsed. The container material of two of the reclaimed 85 Kr legacy waste form samples was examined by optical microscopy and by SEM-EDS. Substantial corrosion was observed throughout each sample container. The bulk of the container material was identified as carbon steel, while stainless steel was identified as the weld material used in container manufacture and/or sealing. The material used for Kr encapsulation was found adhered to the walls of each container and had a chemical composition consistent with zeolite minerals. Kr was found coincident with the adhered material, demonstrating that the encapsulation method used in the original experiments was effective in the long-term retention of Kr. In conclusion, the sectioning, mounting, and initial analysis of these legacy samples is an important step in the study of all five legacy 85 Kr waste forms and the information gained from the analysis of Samples 2 and 5 will be used to guide future efforts on these remaining samples.
Experiments are being conducted to evaluate the early release of technetium (Tc) from fresh salt-stone waste forms using dynamic flow-through tests. Salt-stone is a chemically reducing, grout-based waste form used by the U.S. Department of Energy (DOE) to dispose of low-activity salt solutions arising from high-level radioactive waste tank closure activities at the Savannah River Site (SRS). A key safety function of salt-stone is maintaining a low-flow, chemically reducing environment that will attenuate the release of redox-sensitive radioelements such as Tc. Salt-stone also provides a physical barrier due to its low intrinsic permeability, which limits both water and oxygen ingress. These release experiments, conducted to support the U.S. Nuclear Regulatory Commission's monitoring role during tank closure, are designed to provide independent information on how well salt-stone will perform its safety function. The dynamic leach tests were conducted in a Hassler-type core-holder permeameter system that allows an accelerated flow rate of simulated SRS groundwater through an intact cylinder of salt-stone spiked with Tc-99 at a concentration of approximately 2,100 Bq/g (3 ppm). The first salt-stone specimen to be tested (Specimen 2) was prepared with Grade 100 ground granulated blast-furnace slag and aged for 119 days in a temperature-controlled, low-oxygen, high-humidity environment. Before test completion, 9.5 pore volumes (PV) of simulated SRS groundwater were passed through the salt-stone cylinder. After a few PV, apparent saturated hydraulic conductivity (K{sub sat}) values calculated from flow data reached a steady-state value of approximately 1.0 x 10{sup -9} cm/s, which is within the range of previously reported data. Tc-99 concentrations in leachate peaked early during the leach test at 2.0 x 10{sup -7} M and then declined to approximately 4 x 10{sup -8} M after 9.5 PV. The leachate Tc-99 concentration after elution of 9.5 PV is broadly consistent with a Tc solution concentration controlled by the solubility of reduced Tc solids. The range of Tc-99 concentrations is similar to the range reported for a column test that was sponsored by DOE, but the time pattern of release differs. The second salt-stone specimen tested (Specimen 3) was aged for 372 days and it further differed from the first specimen in that its Grade 100 slag component had been finely ground using a mortar and pestle in an attempt to make it more reactive. This second test yielded Tc release behavior broadly similar to that observed during the first test, except that the early Tc peak (at 1.2 x 10{sup -6} M) was six times higher and the concentration after 6.7 PV (8.8 x 10{sup -8} M) was approximately two times higher than in the first test. In addition, the steady-state K{sub sat} was an order of magnitude lower than that of the first specimen. During the second test, the injection pump was intentionally turned off for a period of 11 days (i.e., approximate residence time for 1 PV) and a Tc spike was observed when pumping resumed, suggesting that longer residence times may promote higher Tc effluent concentrations. During a second intentional shut-down of the pump for 25 days (approximate residence time for 2 PV), another Tc spike was observed. An unplanned excursion to higher confining pressure physically affected the specimen, temporarily lowering the apparent K{sub sat} followed by a subsequent increase in both K{sub sat} and Tc effluent concentration. One hypothesis is that the transient spike in confining pressure and resulting salt-stone compression and later decompression may have induced Tc release from otherwise inactive pores. (authors)
A Direct Feed Low Activity Waste (DFLAW) flowsheet has been initiated to begin immobilization of Hanford tank wastes. In the DFLAW configuration, a pretreatment system will segregate the high level and the low activity wastes (LAW). The LAW will be immobilized in the Waste Treatment Plant (WTP) and disposed of in Hanford's Integrated Disposal Facility (IDF). To obtain authorization to dispose of wastes in the IDF, a Performance Assessment (PA) must be completed in accordance with DOE Order 435.1. The PA shall include calculations of potential doses to representative future members of the public and potential releases from the facility to provide a reasonable expectation that performance objectives will not be exceeded as a result of operation and closure of the facility. To support the development of the PA, data from ILAW glass testing are needed to model the long term release of contaminants over the compliance period and verify that impacts to groundwater quality will be within performance objective limits. Washington River Protection Solutions is conducting an ILAW glass testing program to produce the required data. This program is a combined effort involving National Laboratories with known expertise in glass waste form research. The work involves accelerated testing to develop parameters that can be used in computer simulations of glass corrosion and contaminant release in the IDF environment. The results of this work will be compiled in an ILAW glass waste form performance data package for the IDF PA. (authors)
The UREX+1 process is under consideration for the separation of transuranic elements from spent nuclear fuel. The first steps of this process extract the fission product technicium-99 ( 99 Tc) into an organic phase containing tributylphosphate together with uranium. Treatment of this stream requires the separation of Tc from U and placement into a suitable waste storage form. A potential candidate waste form involves immobilizing the Tc as an alloy with either excess metallic zirconium or stainless steel. Although Tc-Zr alloys seem to be promising waste forms, alternative materials must be investigated. Innovative studies related to the synthesis and behavior of a different class of Tc materials will increase the scientific knowledge related to development of Tc waste forms. These studies will also provide a better understanding of the behavior of 99 Tc in repository conditions. A literature survey has selected promising alternative waste forms for further study: technetium metallic alloys, nitrides, oxides, sulfides, and pertechnetate salts. The goals of this project are to 1) synthesize and structurally characterize relevant technetium materials that may be considered as waste forms, 2) investigate material behavior in solution under different conditions of temperature, electrochemical potential, and radiation, and 3) predict the long-term behavior of these materials.
Hot isostatic pressing (HIP) is being considered as a process to directly convert 129 I-bearing materials to a durable radiological waste form. Zeolites, specifically silver-exchanged mordenite (AgZ), have been studied extensively as potential radioactive iodine sorbents and will contain 129 I as chemisorbed AgI, a relatively insoluble compound with a high melting point. Oak Ridge National Laboratory has conducted several recent studies on the HIP of both I–AgZ and other iodine-bearing zeolites, resulting in an inventory of samples prepared with a suite of zeolites with varied iodine incorporation methods, pressing conditions, and sizes. These samples have been made available to researchers at other national laboratories in support of test method development to assess the stability of heterogenous iodine-bearing waste forms. The work described here generated four large-format HIPed I–AgZ samples. These samples are 1.5 in. diameter and 2 in. length and contain 49 g of I–AgZ. Two samples were prepared at an iodine loading of 64 mg I/g I– AgZ, and two samples were prepared at an iodine loading of 135 mg I/g I–AgZ. These two sets of duplicate samples will be made available upon request in support of iodine waste form durability method development efforts.
This paper studied the release of iodine from lead vanado-iodoapatite (I-APT, Pb 9.85 (VO 4 ) 6 I 1.7 ), a potential nuclear waste form for the radioactive waste element of I-129, which can be enhanced when crevice corrosion of stainless steel (SS) occurring nearby. Reference corrosion studies of I-APT were performed in different bulk solutions including DI water, 0.6 M and 6 M NaCl, and 0.1 M HNO 3 without metal crevice corrosion interactions. The localized enrichment of Cl – , one of the major consequences of SS crevice corrosion, was found to be the decisive factor that led to the enhanced release of iodine. A surface alteration layer consisting of a mixture of nanocrystalline I-APT and Cl-rich apatite (Cl-APT) formed on I-APT surface. Meanwhile, large Cl-APT crystals formed at the crevice mouth on the I-APT surface. This study reveals a new near-field corrosion mechanism for ceramic waste forms when they are exposed to aggressive local corrosive conditions created by the electrochemical reactions of nearby metals. The insight gained in this study could be beneficial for a more accurate prediction of waste form degradation.
Technetium-99 (Tc, t{sub 1/2} = 2.13 x 10{sup 5} years) is a challenge from a nuclear waste perspective and is one of the most abundant, long-lived radioisotopes found in used nuclear fuel (UNF). Within the Hanford Tank Waste Treatment and Immobilization Plant, technetium volatilises at typical glass melting temperature, is captured in the off-gas treatment system and recycled back into the feed to eventually increase Tc-loadings of the glass. In this research we are providing a technical solution for continuously depleting the Tc-content of the glass melter feed by precipitating Tc from the off-gas effluents and effectively immobilizing {sup 99}Tc as durable ceramic waste form. We have developed a wet-chemical coprecipitation-based synthesis route to fabricate gram-size Tc-based pyrochlores with the model composition Sm{sub 2}(Tc{sub 0.5-x}Ru{sub x}Ti{sub 0.5}){sub 2}O{sub 7}. These ceramic Tc waste forms exhibit better technetium retention and corrosion resistance in ASTM C1220-10 based testing compared with technetium-containing LAWE4-type borosilicate glass, combined with optimal waste loading. (authors)
Decontaminated salt solution from Tank 50 at the Savannah River Site (SRS) is known to contain dissolved mercury species which are present as hydroxides and oxides. These species are reported in analytical measurements as the organic cation [CH 3 Hg] + and the inorganic cation Hg 2+ . The salt solution is mixed with cementitious reagents, blast-furnace slag (BFS) and thermally beneficiated Class F fly ash in the Saltstone Processing Facility and pumped into engineered Saltstone Disposal Units, (SDUs), to produce a solidified waste form. The solidified waste form, saltstone, encapsulates and chemically stabilizes the mercury and other contaminants, sequestering them from release into the environment. With the discovery of methylmercury in the salt solution, testing was performed to determine the extent to which saltstone prepared with methylmercury (MeHg) retains mercury in the Toxicity Characteristic Leaching Procedure (TCLP). Results from this testing indicated that the amount of mercury extracted remained below the regulatory limit, and that the saltstone formulation with a greater proportion of BFS performed slightly better at retaining Hg in the solid.