Hazardous wastes in aquatic environments: Biological uptake and metabolism studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barber, J.; Apblett, A.; Ensley, H.

1996-05-02

The projects discussed in this article include the following: the uptake, accumulation, metabolism, toxicity and physiological effects of various environmentally-important contaminants, inorganic and organic, in several wetland species that are interrelated through food webs; and investigation of the potential for developing and linking chemical and biological methods of remediation so as to encapsulate bioaccummulated ions in stable wasteforms such as ceramics and/or zeolites. 24 refs.

Evaluation of final waste forms and recommendations for baseline alternatives to group and glass

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bleier, A.

1997-09-01

An assessment of final waste forms was made as part of the Federal Facilities Compliance Agreement/Development, Demonstration, Testing, and Evaluation (FFCA/DDT&E) Program because supplemental waste-form technologies are needed for the hazardous, radioactive, and mixed wastes of concern to the Department of Energy and the problematic wastes on the Oak Ridge Reservation. The principal objective was to identify a primary waste-form candidate as an alternative to grout (cement) and glass. The effort principally comprised a literature search, the goal of which was to establish a knowledge base regarding four areas: (1) the waste-form technologies based on grout and glass, (2) candidatemore » alternatives, (3) the wastes that need to be immobilized, and (4) the technical and regulatory constraints on the waste-from technologies. This report serves, in part, to meet this goal. Six families of materials emerged as relevant; inorganic, organic, vitrified, devitrified, ceramic, and metallic matrices. Multiple members of each family were assessed, emphasizing the materials-oriented factors and accounting for the fact that the two most prevalent types of wastes for the FFCA/DDT&E Program are aqueous liquids and inorganic sludges and solids. Presently, no individual matrix is sufficiently developed to permit its immediate implementation as a baseline alternative. Three thermoplastic materials, sulfur-polymer cement (inorganic), bitumen (organic), and polyethylene (organic), are the most technologically developed candidates. Each warrants further study, emphasizing the engineering and economic factors, but each also has limitations that regulate it to a status of short-term alternative. The crystallinity and flexible processing of sulfur provide sulfur-polymer cement with the highest potential for short-term success via encapsulation. Long-term immobilization demands chemical stabilization, which the thermoplastic matrices do not offer. Among the properties of the remaining

Radiation Stability of Benzyl Tributyl Ammonium Chloride towards Technetium-99 Extraction - 13016

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paviet-Hartmann, Patricia; Horkley, Jared; Campbell, Keri

2013-07-01

A closed nuclear fuel cycle combining new separation technologies along with generation III and generation IV reactors is a promising way to achieve a sustainable energy supply. But it is important to keep in mind that future recycling processes of used nuclear fuel (UNF) must minimize wastes, improve partitioning processes, and integrate waste considerations into processes. New separation processes are being developed worldwide to complement the actual industrialized PUREX process which selectively separates U(VI) and Pu(IV) from the raffinate. As an example, the UREX process has been developed in the United States to co-extract hexavalent uranium (U) and hepta-valent technetiummore » (Tc) by tri-n-butyl phosphate (TBP). Tc-99 is recognized to be one of the most abundant, long-lived radio-toxic isotopes in UNF (half-life, t{sub 1/2} = 2.13 x 10{sup 5} years), and as such, is targeted in UNF separation strategies for isolation and encapsulation in solid waste-forms for final disposal in a nuclear waste repository. Immobilization of Tc-99 by a durable solid waste-form is a challenge, and its fate in new advanced technology processes is of importance. It is essential to be able to quantify and locate 1) its occurrence in any new developed flowsheets, 2) its chemical form in the individual phases of a process, 3) its potential quantitative transfer in any waste streams, and consequently, 4) its quantitative separation for either potential transmutation to Ru-100 or isolation and encapsulation in solid waste-forms for ultimate disposal. In addition, as a result of an U(VI)-Tc(VII) co-extraction in a UREX-based process, Tc(VII) could be found in low level waste (LLW) streams. There is a need for the development of new extraction systems that would selectively extract Tc-99 from LLW streams and concentrate it for feed into high level waste (HLW) for either Tc-99 immobilization in metallic waste-forms (Tc-Zr alloys), and/or borosilicate-based waste glass. Studies have

Radionuclide Retention Mechanisms in Secondary Waste-Form Testing: Phase II

DOE Office of Scientific and Technical Information (OSTI.GOV)

Um, Wooyong; Valenta, Michelle M.; Chung, Chul-Woo

2011-09-26

This report describes the results from laboratory tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate candidate stabilization technologies that have the potential to successfully treat liquid secondary waste stream effluents produced by the Hanford Tank Waste Treatment and Immobilization Plant (WTP). WRPS is considering the design and construction of a Solidification Treatment Unit (STU) for the Effluent Treatment Facility (ETF) at Hanford. The ETF, a multi-waste, treatment-and-storage unit that has been permitted under the Resource Conservation and Recovery Act (RCRA), can accept dangerous, low-level, and mixed wastewaters for treatment. The STU needsmore » to be operational by 2018 to receive secondary liquid waste generated during operation of the WTP. The STU will provide the additional capacity needed for ETF to process the increased volume of secondary waste expected to be produced by WTP. This report on radionuclide retention mechanisms describes the testing and characterization results that improve understanding of radionuclide retention mechanisms, especially for pertechnetate, {sup 99}TcO{sub 4}{sup -} in four different waste forms: Cast Stone, DuraLith alkali aluminosilicate geopolymer, encapsulated fluidized bed steam reforming (FBSR) product, and Ceramicrete phosphate bonded ceramic. These data and results will be used to fill existing data gaps on the candidate technologies to support a decision-making process that will identify a subset of the candidate waste forms that are most promising and should undergo further performance testing.« less

Incorporation of thorium in the rhabdophane structure: Synthesis and characterization of Pr1-2xCaxThxPO4·nH2O solid solutions

NASA Astrophysics Data System (ADS)

Qin, Danwen; Mesbah, Adel; Gausse, Clémence; Szenknect, Stéphanie; Dacheux, Nicolas; Clavier, Nicolas

2017-08-01

Thorium incorporation in the rhabdophane structure as Pr1-2xCaxThxPO4·nH2O solid solutions was successfully achieved and resulted in the preparation of a low temperature precursor of the monazite-cheralite type Pr1-2xCaxThxPO4. The rhabdophane compounds are considered as potential neoformed phases in case of release of actinides from the phosphate-based ceramic wasteforms envisaged to host radionuclides in the back-end of the nuclear fuel cycle. A multiparametric study was thus undertaken to specify the wet chemistry conditions (starting stoichiometry, temperature, heating time) leading to single phase Pr1-2xCaxThxPO4·nH2O powdered samples. The excess of calcium appeared to be a prevailing factor with a suggested initial Ca:Th ratio being equal to 10. Similarly, the recommended heating time should exceed 4 days while the optimal temperature of synthesis is 110 °C. Under these conditions, the stability domain of Pr1-2xCaxThxPO4·nH2O ranged from x = 0.00 to x = 0.15. After heating at 1100 °C under air during 6 h, rhabdophane-type samples were fully converted into the highly durable Pr1-2xCaxThxPO4 cheralite ceramic wasteform.

Radiological performance assessment for the E-Area Vaults Disposal Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cook, J.R.

This report is the first revision to ``Radiological Performance Assessment for the E-Area Vaults Disposal Facility, Revision 0'', which was issued in April 1994 and received conditional DOE approval in September 1994. The title of this report has been changed to conform to the current name of the facility. The revision incorporates improved groundwater modeling methodology, which includes a large data base of site specific geotechnical data, and special Analyses on disposal of cement-based wasteforms and naval wastes, issued after publication of Revision 0.

One-dimensional Spatial Distributions of Gamma-ray Emitting Contaminants in Field Lysimeters Using a Collimated Gamma-ray Spectroscopy System.

PubMed

Erdmann, Bryan J; Powell, Brian A; Kaplan, Daniel I; DeVol, Timothy A

2018-05-01

One-dimensional scans of gamma-ray emitting contaminants were conducted on lysimeters from the RadFLEX facility at the Savannah River Nationals Laboratory (SRNL). The lysimeters each contained a contamination source that was buried in SRNL soil. A source consisted of Cs, Co, Ba, and Eu incorporated either into a solid waste form (Portland cement and reducing grout) or applied to a filter paper for direct soil exposure. The lysimeters were exposed to natural environmental conditions for 3 to 4 y. The initial contaminant activities range from 4.0 to 9.0 MBq for the solid wasteforms and 0.25 to 0.47 MBq for the soil-incorporated source. The measurements were performed using a collimated high-purity germanium gamma-ray spectrometer with a spatial resolution of 2.5 mm. These scans showed downward mobility of Co and Ba when the radionuclides were incorporated directly into the SRNL soil. When radionuclides were incorporated into the solid waste forms positioned in the SRNL soil, Cs exhibited both upward and downward dispersion while the other radionuclides showed no movement. This dispersion was more significant for the Portland cement than the reducing grout wasteform. Europium-152 was the only radionuclide of those studied that showed no movement within the spatial resolution of the scanner from the original placement within the lysimeter. Understanding radionuclide movement in the environment is important for developing strategies for waste management and disposal.

Initial results of metal waste-form development activities at ANL-West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Keiser, D.D. Jr.; Westphal, B.R.; Herbst, R.S.

1997-12-01

Argonne National Laboratory (ANL) is developing a metal alloy to contain metallic waste constituent residual from the electrometallurgical treatment of spent nuclear fuel. This alloy will contain stainless steel (from stainless steel-clad fuel elements), {approximately} 15 wt% zirconium (from alloy fuel), fission products noble to the process (e.g., ruthenium, palladium, technetium, etc.), and minor amounts of actinides. The alloy will serve as a final waste form for these components and will be disposed of in a geologic repository. The alloy ingot is produced in an induction furnace situated in a hot cell using argon cover gas. This paper discusses resultsmore » from the melting campaigns that have been initiated at ANL-West to generate the metal waste form using actual process materials. In addition, metal waste form samples have been doped with technetium and selected actinides and are described in the context of how elements of interest partition between various phases in the alloy and how this distribution of elements in the alloy may affect the leaching behavior of the components in an aqueous environment.« less

BLT-EC (Breach, Leach and Transport-Equilibrium Chemistry) data input guide. A computer model for simulating release and coupled geochemical transport of contaminants from a subsurface disposal facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

MacKinnon, R.J.; Sullivan, T.M.; Kinsey, R.R.

1997-05-01

The BLT-EC computer code has been developed, implemented, and tested. BLT-EC is a two-dimensional finite element computer code capable of simulating the time-dependent release and reactive transport of aqueous phase species in a subsurface soil system. BLT-EC contains models to simulate the processes (container degradation, waste-form performance, transport, chemical reactions, and radioactive production and decay) most relevant to estimating the release and transport of contaminants from a subsurface disposal system. Water flow is provided through tabular input or auxiliary files. Container degradation considers localized failure due to pitting corrosion and general failure due to uniform surface degradation processes. Waste-form performancemore » considers release to be limited by one of four mechanisms: rinse with partitioning, diffusion, uniform surface degradation, and solubility. Transport considers the processes of advection, dispersion, diffusion, chemical reaction, radioactive production and decay, and sources (waste form releases). Chemical reactions accounted for include complexation, sorption, dissolution-precipitation, oxidation-reduction, and ion exchange. Radioactive production and decay in the waste form is simulated. To improve the usefulness of BLT-EC, a pre-processor, ECIN, which assists in the creation of chemistry input files, and a post-processor, BLTPLOT, which provides a visual display of the data have been developed. BLT-EC also includes an extensive database of thermodynamic data that is also accessible to ECIN. This document reviews the models implemented in BLT-EC and serves as a guide to creating input files and applying BLT-EC.« less

Plasma vitrification of waste materials

DOEpatents

McLaughlin, David F.; Dighe, Shyam V.; Gass, William R.

1997-01-01

This invention provides a process wherein hazardous or radioactive wastes in the form of liquids, slurries, or finely divided solids are mixed with finely divided glassformers (silica, alumina, soda, etc.) and injected directly into the plume of a non-transferred arc plasma torch. The extremely high temperatures and heat transfer rates makes it possible to convert the waste-glassformer mixture into a fully vitrified molten glass product in a matter of milliseconds. The molten product may then be collected in a crucible for casting into final wasteform geometry, quenching in water, or further holding time to improve homogeneity and eliminate bubbles.

Plasma vitrification of waste materials

DOEpatents

McLaughlin, D.F.; Dighe, S.V.; Gass, W.R.

1997-06-10

This invention provides a process wherein hazardous or radioactive wastes in the form of liquids, slurries, or finely divided solids are mixed with finely divided glassformers (silica, alumina, soda, etc.) and injected directly into the plume of a non-transferred arc plasma torch. The extremely high temperatures and heat transfer rates makes it possible to convert the waste-glassformer mixture into a fully vitrified molten glass product in a matter of milliseconds. The molten product may then be collected in a crucible for casting into final wasteform geometry, quenching in water, or further holding time to improve homogeneity and eliminate bubbles. 4 figs.

Cement waste-form development for ion-exchange resins at the Rocky Flats Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Veazey, G.W.; Ames, R.L.

1997-03-01

This report describes the development of a cement waste form to stabilize ion-exchange resins at Rocky Flats Environmental Technology Site (RFETS). These resins have an elevated potential for ignition due to inadequate wetness and contact with nitrates. The work focused on the preparation and performance evaluation of several Portland cement/resin formulations. The performance standards were chosen to address Waste Isolation Pilot Plant and Environmental Protection Agency Resource Conservation and Recovery Act requirements, compatibility with Rocky Flats equipment, and throughput efficiency. The work was performed with surrogate gel-type Dowex cation- and anion-exchange resins chosen to be representative of the resin inventorymore » at RFETS. Work was initiated with nonactinide resins to establish formulation ranges that would meet performance standards. Results were then verified and refined with actinide-containing resins. The final recommended formulation that passed all performance standards was determined to be a cement/water/resin (C/W/R) wt % ratio of 63/27/10 at a pH of 9 to 12. The recommendations include the acceptable compositional ranges for each component of the C/W/R ratio. Also included in this report are a recommended procedure, an equipment list, and observations/suggestions for implementation at RFETS. In addition, information is included that explains why denitration of the resin is unnecessary for stabilizing its ignitability potential.« less

Assessing the oxidation states and structural stability of the Ce analogue of brannerite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aluri, Esther Rani; Bachiu, Lisa M.; Grosvenor, Andrew P.

2017-07-04

The Ce-containing analogue of brannerite (ie, UTi2O6) was previously considered to be stoichiometric (ie, CeTi2O6); however, it has recently been determined that the material is O deficient. This oxygen-deficient material has been suggested to be charged balanced by the presence of a minor concentration of Ce3+ or by the A-site being cation deficient with the Ce oxidation state being 4+. A variety of Ti-containing oxides (including brannerite) have been investigated as potential nuclear wasteforms, and it is necessary to understand the electronic structure of a proposed nuclear wasteform material as well as how the structure responds to radiation from incorporatedmore » waste elements. The radiation resistance of a material can be simulated by ion implantation. The objective of this study was to confirm the Ce oxidation state in the cation- and oxygen-deficient material (ie, Ce0.94Ti2O6 - δ) and to determine how radiation damage affects this material. X-ray photoelectron spectroscopy (XPS) and X-ray absorption near-edge spectroscopy were used to study Ce0.94Ti2O6 - δ before and after being implanted with 2 MeV Au- ions. Analysis of the Ce 3d XPS spectra from the as-synthesized samples by using a previously developed fitting method has unequivocally shown that Ce adopts both 4+ (major) and 3+ (minor) oxidation states, which was confirmed by examination of magnetic susceptibility data. Analysis of XPS and X-ray absorption near-edge spectroscopy spectra from ion-implanted materials showed that both Ce and Ti were reduced because of radiation damage and that the local coordination environments of the cations are greatly affected by radiation damage.« less

Bubble Formation and Lattice Parameter Changes Resulting from He Irradiation of Defect-Fluorite Gd2Zr2O7

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taylor, Caitlin A.; Patel, Maulik K.; Aguiar, Jeffery A.

2016-08-15

Pyrochlores have long been considered as potential candidates for advanced ceramic waste-forms for the immobilization of radioactive waste nuclides. This work provides evidence that Gd2Zr2O7, often considered the most radiation tolerant pyrochlore, could be susceptible to radiation damage in the form of bubble nucleation at the highest He doses expected over geological time. Ion irradiations were utilized to experimentally simulate the radiation damage and He accumulation produced by ..alpha..-decay. Samples were pre-damaged using 7 MeV Au3+ to induce the pyrochlore to defect-fluorite phase transformation, which would occur due to ..alpha..-recoil damage within several hundred years of storage in a Gd2Zr2O7more » waste-form. These samples were then implanted to various He concentrations in order to study the long-term effects of He accumulation. Helium bubbles 1-3 nm in diameter were observed in TEM at a concentration of 4.6 at.% He. Some bubbles remained isolated, while others formed chains 10-30 nm in length parallel to the surface. GIXRD measurements showed lattice swelling after irradiating pristine Gd2Zr2O7 with 7 MeV Au3+ to a fluence of 2.2 x 1015 Au/cm2. An increase in lattice swelling was also measured after 2.2 x 1015 Au/cm2 + 2 x 1015 He/cm2 and 2.2 x 1015 Au/cm2 + 2 x 1016 He/cm2. A decrease in lattice swelling was measured after irradiation with 2.2 x 1015 Au/cm2 + 2 x 1017 He/cm2, the fluence where bubbles and bubble chains were observed in TEM. Bubble chains are thought to form in order to reduce lattice strain normal to the surface, which is produced by the Au and He irradiation damage.« less

Reactive spark plasma synthesis of CaZrTi2O7 zirconolite ceramics for plutonium disposition

NASA Astrophysics Data System (ADS)

Sun, Shi-Kuan; Stennett, Martin C.; Corkhill, Claire L.; Hyatt, Neil C.

2018-03-01

Near single phase zirconolite ceramics, prototypically CaZrTi2O7, were fabricated by reactive spark plasma sintering (RSPS), from commercially available CaTiO3, ZrO2 and TiO2 reagents, after processing at 1200 °C for only 1 h. Ceramics were of theoretical density and formed with a controlled mean grain size of 1.9 ± 0.6 μm. The reducing conditions of RSPS afforded the presence of paramagnetic Ti3+, as demonstrated by EPR spectroscopy. Overall, this study demonstrates the potential for RSPS to be a disruptive technology for disposition of surplus separated plutonium stockpiles in ceramic wasteforms, given its inherent advantage of near net shape products and rapid throughput.

Crystal Chemical Substitution at Ca and La Sites in CaLa4(SiO4)3O To Design the Composition Ca1- xM xLa4-xRE x(SiO4)3O for Nuclear Waste Immobilization and Its Influence on the Thermal Expansion Behavior.

PubMed

Ravikumar, Ramya; Gopal, Buvaneswari; Jena, Hrudananda

2018-06-04

The oxysilicate apatite host CaLa 4 (SiO 4 ) 3 O has been explored for immobilization of radioactive nuclides. Divalent ion, trivalent rare earth ion, and combined ionic substitutions in the silicate oxyapatite were carried out to optimize the simulated wasteform composition. The phases were characterized by powder X-ray diffraction, FT-IR, TGA, SEM-EDS, and HT-XRD techniques. The results revealed the effect of ionic substitutions on the structure and thermal expansion behavior. The investigation resulted in the formulation of simulated wasteforms such as La 3.4 Ce 0.1 Pr 0.1 Nd 0.1 Sm 0.1 Gd 0.1 Y 0.1 (SiO 4 ) 3 O (WF-1) and Ca 0.8 Sr 0.1 Pb 0.1 La 3.4 Ce 0.1 Pr 0.1 Nd 0.1 Sm 0.1 Gd 0.1 Y 0.1 (SiO 4 ) 3 O (WF-2). In comparison to the average axial thermal expansion coefficients of the hexagonal unit cell of the parent CaLa 4 (SiO 4 ) 3 O measured in the temperature range 298-1073 K (α' a = 9.74 × 10 -6 K -1 and α' c = 10.10 × 10 -6 K -1 ), rare earth ion substitution decreases the thermal expansion coefficients, as in the case of La 3.4 Ce 0.1 Pr 0.1 Nd 0.1 Sm 0.1 Gd 0.1 Y 0.1 (SiO 4 ) 3 O (α' a = 8.67 × 10 -6 K -1 and α' c = 7.94 × 10 -6 K -1 ). However, the phase Ca 0.8 Sr 0.1 Pb 0.1 La 3.4 Ce 0.1 Pr 0.1 Nd 0.1 Sm 0.1 Gd 0.1 Y 0.1 (SiO 4 ) 3 O shows an increase in the values of thermal expansion coefficients: α' a = 11.74 × 10 -6 K -1 and α' c = 11.70 × 10 -6 K -1 .

Development of Risk Insights for Regulatory Review of a Near-Surface Disposal Facility for Radioactive Waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Esh, D.W.; Ridge, A.C.; Thaggard, M.

2006-07-01

key risk drivers and risk limiters of the SDF. Review emphasis was placed on those aspects of the disposal system that were expected to drive performance: the physical and chemical performance of the cementitious wasteform and concrete vaults. Refinement of the modeling of the degradation and release from the cementitious wasteform had a significant effect on the predicted dose to a member of the public. (authors)« less

Glass Property Data and Models for Estimating High-Level Waste Glass Volume

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vienna, John D.; Fluegel, Alexander; Kim, Dong-Sang

2009-10-05

This report describes recent efforts to develop glass property models that can be used to help estimate the volume of high-level waste (HLW) glass that will result from vitrification of Hanford tank waste. The compositions of acceptable and processable HLW glasses need to be optimized to minimize the waste-form volume and, hence, to save cost. A database of properties and associated compositions for simulated waste glasses was collected for developing property-composition models. This database, although not comprehensive, represents a large fraction of data on waste-glass compositions and properties that were available at the time of this report. Glass property-composition modelsmore » were fit to subsets of the database for several key glass properties. These models apply to a significantly broader composition space than those previously publised. These models should be considered for interim use in calculating properties of Hanford waste glasses.« less

Hanford waste-form release and sediment interaction: A status report with rationale and recommendations for additional studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Serne, R.J.; Wood, M.I.

1990-05-01

This report documents the currently available geochemical data base for release and retardation for actual Hanford Site materials (wastes and/or sediments). The report also recommends specific laboratory tests and presents the rationale for the recommendations. The purpose of this document is threefold: to summarize currently available information, to provide a strategy for generating additional data, and to provide recommendations on specific data collection methods and tests matrices. This report outlines a data collection approach that relies on feedback from performance analyses to ascertain when adequate data have been collected. The data collection scheme emphasizes laboratory testing based on empiricism. 196more » refs., 4 figs., 36 tabs.« less

The Pressure-Induced Structural Response of A2Hf2O7 (A=Y, Sm, Eu, Gd, Dy, Yb) Compounds from 0.1-50 GPa

NASA Astrophysics Data System (ADS)

Turner, K. M.; Rittman, D.; Heymach, R.; Turner, M.; Tracy, C.; Mao, W. L.; Ewing, R. C.

2016-12-01

A2B2O7 (A, B= cations) compounds have structures that make their properties conducive to many applications; for example they are a proposed waste-form for actinides generated in the nuclear fuel cycle. This interest in part is due to their structural responses to extreme environments of high P, T, or under intense irradiation. Depending on their cationic radius ratio, ra/rb, A2B2O7 compounds either crystallize as pyrochlore (ra/rb=1.46-1.7) or "defect fluorite" (ra/rb>1.46). The structure types are similar: they are derivatives of ideal fluorite with two cations and 1/8 missing anions. In pyrochlore, the cations and anion vacancy are ordered. In "defect fluorite"-structured oxides, the cations and anion vacancies are random. A2B2O7 compounds rarely amorphize in extreme environments. Rather, they disorder and undergo phase transitions; this resistance to amorphization contributes to the durability of this potential actinide waste-form. Under high-pressure, A2B2O7 compounds are known to disorder or form a cottunite-like phase. Their radius ratio affects their response to extreme environments; "defect fluorite" type compounds tend to disorder, and pyrochlore type compounds tend to form the cottunite-like phase. We have examined six A2Hf2O7 compounds (A=Y, Sm, Eu, Gd, Dy, Yb) in situ to 50 GPa. By keeping the B-site constant (Hf), we examined the effect of a changing radius ratio on the pressure-induced structural response of hafnates. We used symmetric DACs, ruby fluorescence, stainless steel gaskets, and methanol: ethanol (4:1 by volume) pressure medium. We characterized these materials with in situ Raman spectroscopy at Stanford University, and synchrotron X-Ray Diffraction (XRD) at APS 16 BM-D and ALS 12.2.2. The compounds were pyrochlore structured (Sm, Eu, Gd) and "defect-fluorite" structured (Y, Dy, Yb) hafnates . These compounds undergo a slow phase transition to a high-pressure cotunnite-like phase between 18-30 GPa. They undergo disordering of their cation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunn, Darrell; Poinssot, Christophe; Begg, Bruce

Management of nuclear waste remains an important international topic that includes reprocessing of commercial nuclear fuel, waste-form design and development, storage and disposal packaging, the process of repository site selection, system design, and performance assessment. Requirements to manage and dispose of materials from the production of nuclear weapons, and the renewed interest in nuclear power, in particular through the Generation IV Forum and the Advanced Fuel Cycle Initiative, can be expected to increase the need for scientific advances in waste management. A broad range of scientific and engineering disciplines is necessary to provide safe and effective solutions and address complexmore » issues. This volume offers an interdisciplinary perspective on materials-related issues associated with nuclear waste management programs. Invited and contributed papers cover a wide range of topics including studies on: spent fuel; performance assessment and models; waste forms for low- and intermediate-level waste; ceramic and glass waste forms for plutonium and high-level waste; radionuclides; containers and engineered barriers; disposal environments and site characteristics; and partitioning and transmutation.« less

Evolution in performance assessment modeling as a result of regulatory review

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rowat, J.H.; Dolinar, G.M.; Stephens, M.E.

1995-12-31

AECL is planning to build the IRUS (Intrusion Resistant Underground Structure) facility for near-surface disposal of LLRW. The PSAR (preliminary safety assessment report) was subject to an initial regulatory review during mid-1992. The regulatory authority provided comments on many aspects of the safety assessment documentation including a number of questions on specific PA (Performance Assessment) modelling assumptions. As a result of these comments as well as a separate detailed review of the IRUS disposal concept, changes were made to the conceptual and mathematical models. The original disposal concept included a non-sorbing vault backfill, with a strong reliance on the wasteformmore » as a barrier. This concept was altered to decrease reliance on the wasteform by replacing the original backfill with a sand/clinoptilolite mix, which is a better sorber of metal cations. This change lead to changes in the PA models which in turn altered the safety case for the facility. This, and other changes that impacted performance assessment modelling are the subject of this paper.« less

Combined effects of radiation damage and He accumulation on bubble nucleation in Gd2Ti2O7

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taylor, Caitlin A.; Patel, Maulik K.; Aguiar, Jeffery A.

2016-10-01

Pyrochlores have long been considered as host phases for long-term immobilization of radioactive waste nuclides that would undergo ..alpha..-decay for hundreds of thousands of years. This work utilizes ion-beam irradiations to examine the combined effects of radiation damage and He accumulation on bubble formation in Gd2Ti2O7 over relevant waste-form timescales. Helium bubbles are not observed in pre-damaged Gd2Ti2O7 implanted with 2 x 1016 He/cm2, even after post-implantation irradiations with 7 MeV Au3+ at 300, 500, and 700 K. However, He bubbles with average diameters of 1.5 nm and 2.1 nm are observed in pre-damaged (amorphous) Gd2Ti2O7 and pristine Gd2Ti2O7, respectively,more » after implantation of 2 x 1017 He/cm2. The critical He concentration for bubble nucleation in Gd2Ti2O7 is estimated to be 6 at.% He.« less

Waste-form development for conversion to portland cement at Los Alamos National Laboratory (LANL) Technical Area 55 (TA-55)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Veazey, G.W.; Schake, A.R.; Shalek, P.D.

1996-10-01

The process used at TA-55 to cement transuranic (TRU) waste has experienced several problems with the gypsum-based cement currently being used. Specifically, the waste form could not reliably pass the Waste Isolation Pilot Plant (WIPP) prohibition for free liquid and the Environmental Protection Agency (EPA)-Toxicity Characteristic Leaching Procedure (TCLP) standard for chromium. This report describes the project to develop a portland cement-based waste form that ensures compliance to these standards, as well as other performance standards consisting of homogeneous mixing, moderate hydration temperature, timely initial set, and structural durability. Testing was conducted using the two most common waste streams requiringmore » cementation as of February 1994, lean residue (LR)- and oxalate filtrate (OX)-based evaporator bottoms (EV). A formulation with a pH of 10.3 to 12.1 and a minimum cement-to-liquid (C/L) ratio of 0.80 kg/l for OX-based EV and 0.94 kg/L for LR-based EV was found to pass the performance standards chosen for this project. The implementation of the portland process should result in a yearly cost savings for raw materials of approximately $27,000 over the gypsum process.« less

Memo, "Incorporation of HLW Glass Shell V2.0 into the Flowsheets," to ED Lee, CCN: 184905, October 20, 2009

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gimpel, Rodney F.; Kruger, Albert A.

2013-12-18

Efforts are being made to increase the efficiency and decrease the cost of vitrifying radioactive waste stored in tanks at the U.S. Department of Energy Hanford Site. The compositions of acceptable and processable high-level waste (HL W) glasses need to be optimized to minimize the waste-form volume and, hence, to reduce cost. A database of glass properties of waste glass and associated simulated waste glasses was collected and documented in PNNL 18501, Glass Property Data and Models for Estimating High-Level Waste Glass Volume and glass property models were curve-fitted to the glass compositions. A routine was developed that estimates HLmore » W glass volumes using the following glass property models: II Nepheline, II One-Percent Crystal Temperature (T1%), II Viscosity (11) II Product Consistency Tests (PCT) for boron, sodium, and lithium, and II Liquidus Temperature (TL). The routine, commonly called the HL W Glass Shell, is presented in this document. In addition to the use of the glass property models, glass composition constraints and rules, as recommend in PNNL 18501 and in other documents (as referenced in this report) were incorporated. This new version of the HL W Glass Shell should generally estimate higher waste loading in the HL W glass than previous versions.« less

Characterisation of Ba(OH){sub 2}–Na{sub 2}SO{sub 4}–blast furnace slag cement-like composites for the immobilisation of sulfate bearing nuclear wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mobasher, Neda; Bernal, Susan A.; Hussain, Oday H.

2014-12-15

Soluble sulfate ions in nuclear waste can have detrimental effects on cementitious wasteforms and disposal facilities based on Portland cement. As an alternative, Ba(OH){sub 2}–Na{sub 2}SO{sub 4}–blast furnace slag composites are studied for immobilisation of sulfate-bearing nuclear wastes. Calcium aluminosilicate hydrate (C–A–S–H) with some barium substitution is the main binder phase, with barium also present in the low solubility salts BaSO{sub 4} and BaCO{sub 3}, along with Ba-substituted calcium sulfoaluminate hydrates, and a hydrotalcite-type layered double hydroxide. This reaction product assemblage indicates that Ba(OH){sub 2} and Na{sub 2}SO{sub 4} act as alkaline activators and control the reaction of the slagmore » in addition to forming insoluble BaSO{sub 4}, and this restricts sulfate availability for further reaction as long as sufficient Ba(OH){sub 2} is added. An increased content of Ba(OH){sub 2} promotes a higher degree of reaction, and the formation of a highly cross-linked C–A–S–H gel. These Ba(OH){sub 2}–Na{sub 2}SO{sub 4}–blast furnace slag composite binders could be effective in the immobilisation of sulfate-bearing nuclear wastes.« less

Ageing of a phosphate ceramic used to immobilize chloride contaminated actinide waste

NASA Astrophysics Data System (ADS)

Metcalfe, B. L.; Donald, I. W.; Fong, S. K.; Gerrard, L. A.; Strachan, D. M.; Scheele, R. D.

2009-03-01

A process for the immobilization of intermediate level waste containing a significant quantity of chloride using Ca3(PO4)2 as the host material has been developed. Waste ions are incorporated into two phosphate-based phases, chlorapatite [Ca5(PO4)3Cl] and spodiosite [Ca2(PO4)Cl]. Non-active trials performed using Sm as the actinide surrogate demonstrated the durability of these phases in aqueous solution. Trials of the process, in which actinide-doped materials were used, were performed at PNNL which confirmed the wasteform resistant to aqueous leaching. Initial leach trials conducted on 239Pu/241Am loaded ceramic at 313 K/28 days gave normalized mass losses of 1.2 × 10-5 g m-2 and 2.7 × 10-3 g m-2 for Pu and Cl, respectively. In order to assess the response of the phases to radiation-induced damage, accelerated ageing trials were performed on samples in which the 239Pu was replaced with 238Pu. No changes to the crystalline structure of the waste were detected in the XRD spectra after the samples had experienced an α radiation fluence of 4 × 1018 g-1. Leach trials showed that there was an increase in the P and Ca release rates but no change in the Pu release rate.

Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

DOEpatents

Boatner, Lynn A.; Sales, Brian C.

1989-01-01

Lead-iron phosphate glasses containing a high level of Fe.sub.2 O.sub.3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90.degree. C., with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10.sup.2 to 10.sup.3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe.sub.2 O.sub.3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800.degree. C., since they exhibit very low melt viscosities in the 800.degree. to 1050.degree. C. temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550.degree. C. and are not adversely affected by large doses of gamma radiation in H.sub.2 O at 135.degree. C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms.

Ageing of a phosphate ceramic used to immobilize chloride contaminated actinide waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Metcalfe, Brian L.; Donald, Ian W.; Fong, Shirley K.

2009-03-31

AWE has developed a process for the immobilization of ILW waste containing a significant quantity of chloride using Ca3(PO4)2 as the host material. Waste ions are incorporated into two phosphate based phases, chlorapatite, Ca5(PO4)3Cl, and spodiosite, Ca2(PO4)Cl. Non-active trials performed at AWE using samarium as the actinide surrogate demonstrated the durability of these phases in aqueous solution. Trials of the process using actinide-doped material were performed at PNNL which confirmed the immobilized wasteform resistant to aqueous leaching. Initial leach trials conducted on 239Pu /241Am loaded ceramic at 40°C/28 days gave normalized mass losses of 1.2 x 10-5 g.m-2 and 2.7more » x 10-3 g.m-2 for Pu and Cl respectively. In order to assess the response of the phases to radiation-induced damage, accelerated ageing trials were performed on samples in which the 239Pu was replaced by 238Pu. No changes to the crystalline structure of the waste were detected using XRD after the samples had experienced a radiation dose of 4 x 1018 α.g-1. Leach trials showed that there had been an increase in the P and Ca release rates but no change in the Pu release rate.« less

Microbial degradation of isosaccharinic acid at high pH

PubMed Central

Bassil, Naji M; Bryan, Nicholas; Lloyd, Jonathan R

2015-01-01

Intermediate-level radioactive waste (ILW), which dominates the radioactive waste inventory in the United Kingdom on a volumetric basis, is proposed to be disposed of via a multibarrier deep geological disposal facility (GDF). ILW is a heterogeneous wasteform that contains substantial amounts of cellulosic material encased in concrete. Upon resaturation of the facility with groundwater, alkali conditions will dominate and will lead to the chemical degradation of cellulose, producing a substantial amount of organic co-contaminants, particularly isosaccharinic acid (ISA). ISA can form soluble complexes with radionuclides, thereby mobilising them and posing a potential threat to the surrounding environment or ‘far field'. Alkaliphilic microorganisms sampled from a legacy lime working site, which is an analogue for an ILW-GDF, were able to degrade ISA and couple this degradation to the reduction of electron acceptors that will dominate as the GDF progresses from an aerobic ‘open phase' through nitrate- and Fe(III)-reducing conditions post closure. Furthermore, pyrosequencing analyses showed that bacterial diversity declined as the reduction potential of the electron acceptor decreased and that more specialised organisms dominated under anaerobic conditions. These results imply that the microbial attenuation of ISA and comparable organic complexants, initially present or formed in situ, may play a role in reducing the mobility of radionuclides from an ILW-GDF, facilitating the reduction of undue pessimism in the long-term performance assessment of such facilities. PMID:25062127

The effects of BaSO₄ loading on OPC cementing system for encapsulation of BaSO₄ scale from oil and gas industry.

PubMed

Hussein, O; Utton, C; Ojovan, M; Kinoshita, H

2013-10-15

The BaSO4 scales obtained from piping decontamination from oil and gas industries are most often classified as low level radioactive waste. These wastes could be immobilised by stable cement matrix to provide higher safety of handling, transportation, storage and disposal. However, the information available for the effects of the basic formulation such as waste loading on the fundamental properties is still limited. The present study investigated the effect of BaSO4 loading and water content on the properties of OPC-BaSO4 systems containing fine BaSO4 powder and coarse granules. The BaSO4 with different particle size had a marked effect on the compressive strength due to their different effects on hydration products formed. Introduction of fine BaSO4 powder resulted in an increased formation of CaCO3 in the system, which significantly contributed to the compressive strength of the products. Amount of water was important to control the CaCO3 formation, and water to cement ratio of 0.53 was found to be a good level to maintain a low porosity of the products both for fine BaSO4 powder and coarse BaSO4 granule. BaSO4 loading of up to 60 wt% has been achieved satisfying the minimum compressive strength of 5 MPa required for the radioactive wasteforms. Copyright © 2013 Elsevier B.V. All rights reserved.

Stabilization/solidification of radioactive salt waste by using xSiO2-yAl2O3-zP2O5 (SAP) material at molten salt state.

PubMed

Park, Hwan-Seo; Kim, In-Tae; Cho, Yong-Zun; Eun, Hee-Chul; Lee, Han-Soo

2008-12-15

The molten salt waste from the pyroprocess is one of the problematic wastes to directly apply a conventional process such as vitrification or ceramization. This study suggested a novel method using a reactive material for metal chlorides at a molten temperature of salt waste, and then converting them into manageable product at a high temperature. The inorganic composite, SAP (SiO2-Al2O3-P2O5), synthesized by a conventional sol-gel process has three or four distinctive domains that are bonded sequentially, Si-O-Si-O-A-O-P-O-P. The P-rich phase in the SAP composite is unstable for producing a series of reactive sites when in contact with a molten LiCl salt. After the reaction, metal aluminosilicate, metal aluminophosphate, metal phosphates and gaseous chlorines are generated. From this process, the volatile salt waste is stabilized and it is possible to apply a high temperature process. The reaction products were fabricated successfully by using a borosilicate glass with an arbitrary composition as a chemical binder. There was a low possibility for the valorization of radionuclides up to 1200 degrees C, based on the result of the thermo gravimetric analysis. The Cs and Sr leach rates by the PCT-A method were about 1 x 10(-3) g/(m2 day). For the final disposal of the problematic salt waste, this approach suggested the design concept of an effective stabilizer for metal chlorides and revealed the chemical route to the fabrication of monolithic wasteform by using a composite as an example. Using this method, we could obtain a higher disposal efficiency and lower waste volume, compared with the present immobilization methods.

Effect of Iron and Carbonation of the Diffusion of Iodine and Rhenium in Waste Encasement Concrete and Soil Fill Material under Hydraulically Unsaturated Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wellman, Dawn M.; Parker, Kent E.; Powers, Laura

2008-07-31

Assessing long-term performance of Category 3 cement wasteforms and accurate prediction for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e. sorption or precipitation). A set of sediment-concrete half-cell diffusion experiments was conducted under unsaturated conditions (4% and 7% by weight moisture content) using carbonated and non-carbonated concrete-soil half-cells. Results indicate the behavior of rhenium and iodine release was comparable within a given half-cell test. Diffusivity in soil is a function of moisture content; a 3% increase in moisture content affords a one to two order of magnitude increase in diffusivity. Release of iodine and rheniummore » was 1 to 3 orders of magnitude less from non-carbonated, relative to carbonated, concrete monoliths. Inclusion of iron in non-carbonate monoliths resulted in the lowest concrete diffusivity values for both iodine and rhenium. This suggests that in the presence of iron, iodine and rhenium are converted to reduced species, which are less soluble and better retained within the concrete monolith. The release of iodine and rhenium was greatest from iron-bearing, carbonated concrete monoliths, suggesting carbonation negates the effect of iron on the retention of iodine and rhenium within concrete monoliths. This is likely due to enhanced formation of microcracks in the presence of iron, which provide preferential paths for contaminant migration. Although the release of iodine and rhenium were greatest from carbonated concrete monoliths containing iron, the migration of iodine and rhenium within a given half-cell is dependent on the moisture content, soil diffusivity, and diffusing species.« less

Biogenic Hydroxyapatite: A New Material for the Preservation and Restoration of the Built Environment.

PubMed

Turner, Ronald J; Renshaw, Joanna C; Hamilton, Andrea

2017-09-20

Ordinary Portland cement (OPC) is by weight the world's most produced man-made material and is used in a variety of applications in environments ranging from buildings, to nuclear wasteforms, and within the human body. In this paper, we present for the first time the direct deposition of biogenic hydroxyapatite onto the surface of OPC in a synergistic process which uses the composition of the cement substrate. This hydroxyapatite is very similar to that found in nature, having a similar crystallite size, iron and carbonate substitution, and a semi-crystalline structure. Hydroxyapatites with such a structure are known to be mechanically stronger and more biocompatible than synthetic or biomimetic hydroxyapatites. The formation of this biogenic hydroxyapatite coating therefore has significance in a range of contexts. In medicine, hydroxyapatite coatings are linked to improved biocompatibility of ceramic implant materials. In the built environment, hydroxyapatite coatings have been proposed for the consolidation and protection of sculptural materials such as marble and limestone, with biogenic hydroxyapatites having reduced solubility compared to synthetic apatites. Hydroxyapatites have also been established as effective for the adsorption and remediation of environmental contaminants such as radionuclides and heavy metals. We identify that in addition to providing a biofilm scaffold for nucleation, the metabolic activity of Pseudomonas fluorescens increases the pH of the growth medium to a suitable level for hydroxyapatite formation. The generated ammonia reacts with phosphate in the growth medium, producing ammonium phosphates which are a precursor to the formation of hydroxyapatite under conditions of ambient temperature and pressure. Subsequently, this biogenic deposition process takes place in a simple reaction system under mild chemical conditions and is cheap and easy to apply to fragile biological or architectural surfaces.

TECHNICAL ASSESSMENT OF BULK VITRIFICATION PROCESS & PRODUCT FOR TANK WASTE TREATMENT AT THE DEPARTMENT OF ENERGY HANFORD SITE

DOE Office of Scientific and Technical Information (OSTI.GOV)

SCHAUS, P.S.

At the U.S. Department of Energy (DOE) Hanford Site, the Waste Treatment Plant (WTP) is being constructed to immobilize both high-level waste (IUW) for disposal in a national repository and low-activity waste (LAW) for onsite, near-surface disposal. The schedule-controlling step for the WTP Project is vitrification of the large volume of LAW, current capacity of the WTP (as planned) would require 50 years to treat the Hanford tank waste, if the entire LAW volume were to be processed through the WTP. To reduce the time and cost for treatment of Hanford Tank Waste, and as required by the Tank Wastemore » Remediation System Environmental Impact Statement Record of Decision and the Hanford Federal Facility Consent Agreement (Tn-Party Agreement), DOE plans to supplement the LAW treatment capacity of the WTP. Since 2002, DOE, in cooperation with the Environmental Protection Agency and State of Washington Department of Ecology has been evaluating technologies that could provide safe and effective supplemental treatment of LAW. Current efforts at Hanford are intended to provide additional information to aid a joint agency decision on which technology will be used to supplement the WTP. A Research, Development and Demonstration permit has been issued by the State of Washington to build and (for a limited time) operate a Demonstration Bulk Vitrification System (DBVS) facility to provide information for the decision on a supplemental treatment technology for up to 50% of the LAW. In the Bulk Vitrification (BV) process, LAW, soil, and glass-forming chemicals are mixed, dried, and placed in a refractory-lined box, Electric current, supplied through two graphite electrodes in the box, melts the waste feed, producing a durable glass waste-form. Although recent modifications to the process have resulted in significant improvements, there are continuing technical concerns.« less

Mineralogical textural and compositional data on the alteration of basaltic glass from Kilauea, Hawaii to 300 degrees C: Insights to the corrosion of a borosilicate glass waste-form. [Yucca Mountain Project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, D.K.

1990-01-01

Mineralogical, textural and compositional data accompanying greenschist facies metamorphism (to 300{degrees}C) of basalts of the East Rift Zone (ERZ), Kilauea, Hawaii may be evaluated relative to published and experimental results for the surface corrosion of borosilicate glass. The ERZ alteration sequence is dominated by intermittent palagonite, interlayered smectite-chlorite, chlorite, and actinolite-epidote-anhydrite. Alteration is best developed in fractures and vesicles where surface reaction layers root on the glass matrix forming rinds in excess of 100 microns thick. Fractures control fluid circulation and the alteration sequence. Proximal to the glass surface, palagonite, Fe-Ti oxides and clays replace fresh glass as the surfacemore » reaction layer migrates inwards; away from the surface, amphibole, anhydrite, quartz and calcite crystallize from hydrothermal fluids in contact with the glass. The texture and composition of basaltic glass surfaces are similar to those of a SRL-165 glass leached statically for sixty days at 150 {degrees}C. While the ERZ reservoir is a complex open system, conservative comparisons between the alteration of ERZ and synthetic borosilicate glass are warranted. 31 refs., 2 figs.« less

The pressure-induced structural response of rare earth hafnate and stannate pyrochlore from 0.1-50 GPa

NASA Astrophysics Data System (ADS)

Turner, K. M.; Rittman, D.; Heymach, R.; Turner, M.; Tracy, C.; Mao, W. L.; Ewing, R. C.

2017-12-01

Complex oxides with the pyrochlore (A2B2O7) and defect-fluorite ((A,B)4O7) structure-types undergo structural transformations under high-pressure. These compounds are under consideration for applications including as a proposed waste-form for actinides generated in the nuclear fuel cycle. High-pressure transformations in rare earth hafnates (A2Hf2O7, A=Sm, Eu, Gd, Dy, Y, Yb) and stannates (A2Sn2O7, A=Nd, Gd, Er) were investigated to 50 GPa by in situ Raman spectroscopy and synchrotron x-ray diffraction (XRD). Rare-earth hafnates form the pyrochlore structure for A=La-Tb and the defect-fluorite structure for A=Dy-Lu. Lanthanide stannates form the pyrochlore structure. Raman spectra revealed that at ambient pressure all compositions have pyrochlore-type short-range order. Stannate compositions show a larger degree of pyrochlore-type short-range ordering relative to hafnates. In situ high-pressure synchrotron XRD showed that rare earth hafnates and stannates underwent a pressure-induced phase transition to a cotunnite-like (Pnma) structure that begins between 18-25 GPa in hafnates and between 30-33 GPa in stannates. The phase transition is not complete at 50 GPa, and upon decompression, XRD indicates that all compositions transform to defect-fluorite with an amorphous component. In situ Raman spectroscopy showed that disordering in stannates and hafnates occurs gradually upon compression. Pyrochlore-structured hafnates retain short-range order to a higher pressure (30 GPa vs. <10 GPa) than defect-fluorite-structured hafnates. Hafnates and stannates decompressed from 50 GPa show Raman spectra consistent with weberite-type structures, also reported in irradiated stannates. The second-order Birch-Murnaghan equation of state fit gives a bulk modulus of 250 GPa for hafnate compositions with the pyrochlore structure, and 400 GPa for hafnate compositions with the defect-fluorite structure. Stannates have a lower bulk modulus relative to hafnates (between 80-150 GPa

Monitoring Radionuclide Transport and Spatial Distribution with a 1D Gamma-Ray Scanner

NASA Astrophysics Data System (ADS)

Dozier, R.; Erdmann, B.; Sams, A.; Barber, K.; DeVol, T. A.; Moysey, S. M.; Powell, B. A.

2016-12-01

Understanding radionuclide movement in the environment is important for informing strategies for radioactive waste management and disposal. A 1-dimensional (1D) gamma-ray emission scanning system was developed to investigate radionuclide transport behavior within soils. Two case studies illustrate the use of the system for non-destructively monitoring transport processes within a soil column. The first case study explores the system capabilities for simultaneously detecting technetium-99m (99mTc), iodine-131 (131I), and sodium-22 (22Na) moving through a column (length = 14.1 cm, diameter = 3.8 cm) packed with soil from the Department of Energy's Savannah River Site. A sodium iodide (NaI) detector was placed at 4 cm above the influent and a Bismuth germanate (BGO) detector at about 10 cm above the influent. The NaI detector results show 99mTc, 131I, and 22Na having similar breakthrough curves with the tail of 99mTc being lower than that of 131I and 22Na. NaCl tracer results compliment the gamma-ray emission measurements. These results are promising because we are able to monitor movement of the isotopes in the column in real-time. In the second case study, the 1D gamma scanner was used to quantify radionuclide mobility within a lysimeter (length = 51 cm, diameter = 10 cm). A cementitious waste form containing cobalt-60 (60Co), barium-133 (133Ba), cesium-137 (137Cs), and europium-152 (152Eu), with the amount of each contained in the cement ranging from 3 to 8.5 MBq, was placed at the midpoint of the lysimeter. The lysimeter was then exposed to natural rainfall and environmental conditions and effluent samples were collected and quantified on a quarterly basis. Following 3.3 years of exposure, the radionuclide distribution in the lysimeter was quantified with a 0.64 cm collimated high-purity germanium gamma-ray spectrometer. Diffusion of 137Cs away from the cementitious wasteform was observed. No movement was seen for 133Ba, 60Co, or 152Eu within the detection limits

Evaluation of Technetium Getters to Improve the Performance of Cast Stone

DOE Office of Scientific and Technical Information (OSTI.GOV)

Neeway, James J.; Qafoku, Nikolla P.; Serne, R. Jeffrey

2015-11-01

Cast Stone has been selected as the preferred waste form for solidification of aqueous secondary liquid effluents from the Hanford Tank Waste Treatment and Immobilization Plant (WTP) process condensates and low-activity waste (LAW) melter off-gas caustic scrubber effluents. Cast Stone is also being evaluated as a supplemental immobilization technology to provide the necessary LAW treatment capacity to complete the Hanford tank waste cleanup mission in a timely and cost effective manner. One of the major radionuclides that Cast Stone has the potential to immobilize is technetium (Tc). The mechanism for immobilization is through the reduction of the highly mobile Tc(VII)more » species to the less mobile Tc(IV) species by the blast furnace slag (BFS) used in the Cast Stone formulation. Technetium immobilization through this method would be beneficial because Tc is one of the most difficult contaminants to address at the U.S. Department of Energy (DOE) Hanford Site due to its complex chemical behavior in tank waste, limited incorporation in mid- to high-temperature immobilization processes (vitrification, steam reformation, etc.), and high mobility in subsurface environments. In fact, the Tank Closure and Waste Management Environmental Impact Statement for the Hanford Site, Richland, Washington (TC&WM EIS) identifies technetium-99 ( 99Tc) as one of the radioactive tank waste components contributing the most to the environmental impact associated with the cleanup of the Hanford Site. The TC&WM EIS, along with an earlier supplemental waste-form risk assessment, used a diffusion-limited release model to estimate the release of different contaminants from the WTP process waste forms. In both of these predictive modeling exercises, where effective diffusivities based on grout performance data available at the time, groundwater at the 100-m down-gradient well exceeded the allowable maximum permissible concentrations for 99Tc. (900 pCi/L). Recent relatively short-term (63

NRC Consultation and Monitoring at the Savannah River Site: Focusing Reviews of Two Different Disposal Actions - 12181

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ridge, A. Christianne; Barr, Cynthia S.; Pinkston, Karen E.

Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA) requires the U.S. Department of Energy (DOE) to consult with the U.S. Nuclear Regulatory Commission (NRC) for certain non-high level waste determinations. The NDAA also requires NRC to monitor DOE's disposal actions related to those determinations. In Fiscal Year 2011, the NRC staff reviewed DOE performance assessments for tank closure at the F-Tank Farm (FTF) Facility and salt waste disposal at the Saltstone Disposal Facility (SDF) at the Savannah River Site (SRS) as part of consultation and monitoring, respectively. Differences in inventories, waste forms,more » and key barriers led to different areas of focus in the NRC reviews of these two activities at the SRS. Because of the key role of chemically reducing grouts in both applications, the evaluation of chemical barriers was significant to both reviews. However, radionuclide solubility in precipitated metal oxides is expected to play a significant role in FTF performance whereas release of several key radionuclides from the SDF is controlled by sorption or precipitation within the cementitious wasteform itself. Similarly, both reviews included an evaluation of physical barriers to flow, but differences in the physical configurations of the waste led to differences in the reviews. For example, NRC's review of the FTF focused on the modeled degradation of carbon steel tank liners while the staff's review of the SDF performance included a detailed evaluation of the physical degradation of the saltstone wasteform and infiltration-limiting closure cap. Because of the long time periods considered (i.e., tens of thousands of years), the NRC reviews of both facilities included detailed evaluation of the engineered chemical and physical barriers. The NRC staff reviews of residual waste disposal in the FTF and salt waste disposal in the SDF focused on physical barriers to flow and chemical barriers to radionuclide

Ion Exchange Distribution Coefficient Tests and Computer Modeling at High Ionic Strength Supporting Technetium Removal Resin Maturation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nash, Charles A.; Hamm, L. Larry; Smith, Frank G.

2014-12-19

final wasteform. This work examined the impact of high ionic strength, high density, and high viscosity if higher concentration LAW feed solution is used. Perrhenate (ReO 4 -) has been shown to be a good nonradioactive surrogate for pertechnetate in laboratory testing for this ion exchange resin, and the performance bias is well established. Equilibrium contact testing with 7.8 M [Na +] average simulant concentrations indicated that the SuperLig ® 639 resin average perrhenate distribution coefficient was 368 mL/g at a 100:1 phase ratio. Although this indicates good performance at high ionic strength, an equilibrium test cannot examine the impact of liquid viscosity, which impacts the diffusivity of ions and therefore the loading kinetics. To get an understanding of the effect of diffusivity, modeling was performed, which will be followed up with column tests in the future.« less

Performance assessment methodology and preliminary results for low-level radioactive waste disposal in Taiwan.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arnold, Bill Walter; Chang, Fu-lin; Mattie, Patrick D.

2006-02-01

the disposal system. Final performance assessment analyses will be used in the regulatory process of licensing a site. The SNL/INER team has developed a performance assessment methodology that is used to simulate processes associated with the potential release of radionuclides to evaluate these sites. The following software codes are utilized in the performance assessment methodology: GoldSim (to implement a probabilistic analysis that will explicitly address uncertainties); the NRC's Breach, Leach, and Transport - Multiple Species (BLT-MS) code (to simulate waste-container degradation, waste-form leaching, and transport through the host rock); the Finite Element Heat and Mass Transfer code (FEHM) (to simulate groundwater flow and estimate flow velocities); the Hydrologic Evaluation of Landfill performance Model (HELP) code (to evaluate infiltration through the disposal cover); the AMBER code (to evaluate human health exposures); and the NRC's Disposal Unit Source Term -- Multiple Species (DUST-MS) code (to screen applicable radionuclides). Preliminary results of the evaluations of the two disposal concept sites are presented.« less

99Tc and Re incorporated into metal oxide polyoxometalates: oxidation state stability elucidated by electrochemistry and theory.

PubMed

McGregor, Donna; Burton-Pye, Benjamin P; Mbomekalle, Israel M; Aparicio, Pablo A; Romo, Susanna; López, Xavier; Poblet, Josep M; Francesconi, Lynn C

2012-08-20

The radioactive element technetium-99 ((99)Tc, half-life = 2.1 × 10(5) years, β(-) of 253 keV), is a major byproduct of (235)U fission in the nuclear fuel cycle. (99)Tc is also found in radioactive waste tanks and in the environment at National Lab sites and fuel reprocessing centers. Separation and storage of the long-lived (99)Tc in an appropriate and stable waste-form is an important issue that needs to be addressed. Considering metal oxide solid-state materials as potential storage matrixes for Tc, we are examining the redox speciation of Tc on the molecular level using polyoxometalates (POMs) as models. In this study we investigate the electrochemistry of Tc complexes of the monovacant Wells-Dawson isomers, α(1)-P(2)W(17)O(61)(10-) (α1) and α(2)-P(2)W(17)O(61)(10-) (α2) to identify features of metal oxide materials that can stabilize the immobile Tc(IV) oxidation state accessed from the synthesized Tc(V)O species and to interrogate other possible oxidation states available to Tc within these materials. The experimental results are consistent with density functional theory (DFT) calculations. Electrochemistry of K(7-n)H(n)[Tc(V)O(α(1)-P(2)W(17)O(61))] (Tc(V)O-α1), K(7-n)H(n)[Tc(V)O(α(2)-P(2)W(17)O(61))] (Tc(V)O-α2) and their rhenium analogues as a function of pH show that the Tc-containing derivatives are always more readily reduced than their Re analogues. Both Tc and Re are reduced more readily in the lacunary α1 site as compared to the α2 site. The DFT calculations elucidate that the highest oxidation state attainable for Re is VII while, under the same electrochemistry conditions, the highest oxidation state for Tc is VI. The M(V)→ M(IV) reduction processes for Tc(V)O-α1 are not pH dependent or only slightly pH dependent suggesting that protonation does not accompany reduction of this species unlike the M(V)O-α2 (M = (99)Tc, Re) and Re(V)O-α1 where M(V/IV) reduction process must occur hand in hand with protonation of the terminal M═O to

SLURRY MIX EVAPORATOR BATCH ACCEPTABILITY AND TEST CASES OF THE PRODUCT COMPOSITION CONTROL SYSTEM WITH THORIUM AS A REPORTABLE ELEMENT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwards, T.

2010-10-07

The Defense Waste Processing Facility (DWPF), which is operated by Savannah River Remediation, LLC (SRR), has recently begun processing Sludge Batch 6 (SB6) by combining it with Frit 418 at a nominal waste loading (WL) of 36%. A unique feature of the SB6/Frit 418 glass system, as compared to the previous glass systems processed in DWPF, is that thorium will be a reportable element (i.e., concentrations of elemental thorium in the final glass product greater than 0.5 weight percent (wt%)) for the resulting wasteform. Several activities were initiated based upon this unique aspect of SB6. One of these was anmore » investigation into the impact of thorium on the models utilized in DWPF's Product Composition and Control System (PCCS). While the PCCS is described in more detail below, for now note that it is utilized by Waste Solidification Engineering (WSE) to evaluate the acceptability of each batch of material in the Slurry Mix Evaporator (SME) before this material is passed on to the melter. The evaluation employs models that predict properties associated with processability and product quality from the composition of vitrified samples of the SME material. The investigation of the impact of thorium on these models was conducted by Peeler and Edwards [1] and led to a recommendation that DWPF can process the SB6/Frit 418 glass system with ThO{sub 2} concentrations up to 1.8 wt% in glass. Questions also arose regarding the handling of thorium in the SME batch acceptability process as documented by Brown, Postles, and Edwards [2]. Specifically, that document is the technical bases of PCCS, and while Peeler and Edwards confirmed the reliability of the models, there is a need to confirm that the current implementation of DWPF's PCCS appropriately handles thorium as a reportable element. Realization of this need led to a Technical Task Request (TTR) prepared by Bricker [3] that identified some specific SME-related activities that the Savannah River National Laboratory

Sintered bentonite ceramics for the immobilization of cesium- and strontium-bearing radioactive waste

NASA Astrophysics Data System (ADS)

Ortega, Luis Humberto

were also tested. The final solid product was a hard dense ceramic with a density that varied from 2.12 g/cm3 for a 19% waste loading with a 1200°C sintering temperature to 3.03 g/cm 3 with a 29% waste loading and sintered at 1100°C. Differential Scanning Calorimetry and Thermal Gravimetric Analysis (DSC-TGA) of the loaded bentonite displayed mass loss steps which were consistent with water losses in pure bentonite. Water losses were complete after dehydroxylation at ˜650°C. No mass losses were evident beyond the dehydroxylation. The ceramic melts at temperatures greater than 1300°C. Light flash analysis found heat capacities of the ceramic to be comparable to those of strontium and barium feldspars as well as pollucite. Thermal conductivity improved with higher sintering temperatures, attributed to lower porosity. Porosity was minimized in 1200°C sinterings. Ceramics with waste loadings less than 25 wt% displayed slump, the lowest waste loading, 15 wt% bloated at a 1200°C sintering. Waste loading above 25 wt% produced smooth uniform ceramics when sintered >1100°C. Sintered bentonite may provide a simple alternative to vitrification and other engineered radioactive waste-forms.

RESULTS OF THE FY09 ENHANCED DOE HIGH LEVEL WASTE MELTER THROUGHPUT STUDIES AT SRNL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, F.; Edwards, T.

2010-06-23

High-level waste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). The objective of this task is to develop data, assess property models, and refine or develop the necessary models to support increased WL of HLW at SRS. It is a continuationmore » of the studies initiated in FY07, but is under the specific guidance of a Task Change Request (TCR)/Work Authorization received from DOE headquarters (Project Number RV071301). Using the data generated in FY07, FY08 and historical data, two test matrices (60 glasses total) were developed at the Savannah River National Laboratory (SRNL) in order to generate data in broader compositional regions. These glasses were fabricated and characterized using chemical composition analysis, X-ray Diffraction (XRD), viscosity, liquidus temperature (TL) measurement and durability as defined by the Product Consistency Test (PCT). The results of this study are summarized below: (1) In general, the current durability model predicts the durabilities of higher waste loading glasses quite well. A few of the glasses exhibited poorer durability than predicted. (2) Some of the glasses exhibited anomalous behavior with respect to durability (normalized leachate for boron (NL [B])). The quenched samples of FY09EM21-02, -07 and -21 contained no nepheline or other wasteform affecting crystals, but have unacceptable NL [B] values (> 10 g/L). The ccc sample of FY09EM21-07 has a NL [B] value that is more than one half the value of the quenched sample. These glasses also have lower concentrations of Al{sub 2}O{sub 3} and SiO{sub 2}. (3) Five of the ccc samples (EM-13, -14, -15

Defining And Characterizing Sample Representativeness For DWPF Melter Feed Samples

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shine, E. P.; Poirier, M. R.

2013-10-29

statisticians used carefully thought out designs that systematically and economically provided plans for data collection from the DWPF process. Key shared features of the sampling designs used at DWPF and the Gy sampling methodology were the specification of a standard for sample representativeness, an investigation that produced data from the process to study the sampling function, and a decision framework used to assess whether the specification was met based on the data. Without going into detail with regard to the seven errors identified by Pierre Gy, as excellent summaries are readily available such as Pitard [1989] and Smith [2001], SRS engineers understood, for example, that samplers can be biased (Gy's extraction error), and developed plans to mitigate those biases. Experiments that compared installed samplers with more representative samples obtained directly from the tank may not have resulted in systematically partitioning sampling errors into the now well-known error categories of Gy, but did provide overall information on the suitability of sampling systems. Most of the designs in this report are related to the DWPF vessels, not the large SRS Tank Farm tanks. Samples from the DWPF Slurry Mix Evaporator (SME), which contains the feed to the DWPF melter, are characterized using standardized analytical methods with known uncertainty. The analytical error is combined with the established error from sampling and processing in DWPF to determine the melter feed composition. This composition is used with the known uncertainty of the models in the Product Composition Control System (PCCS) to ensure that the wasteform that is produced is comfortably within the acceptable processing and product performance region. Having the advantage of many years of processing that meets the waste glass product acceptance criteria, the DWPF process has provided a considerable amount of data about itself in addition to the data from many special studies. Demonstrating representative sampling

New Fragment Separation Technology for Superheavy Element Research

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shaughnessy, D A; Moody, K J; Henderson, R A

2008-01-28

This project consisted of three major research areas: (1) development of a solid Pu ceramic target for the MASHA separator, (2) chemical separation of nuclear decay products, and (3) production of new isotopes and elements through nuclear reactions. There have been 16 publications as a result of this project, and this collection of papers summarizes our accomplishments in each of the three areas of research listed above. The MASHA (Mass Analyzer for Super-Heavy Atoms) separator is being constructed at the U400 Cyclotron at the Flerov Laboratory of Nuclear Reactions in Dubna, Russia. The purpose of the separator is to physicallymore » separate the products from nuclear reactions based on their isotopic masses rather than their decay characteristics. The separator was designed to have a separation between isotopic masses of {+-}0.25 amu, which would enable the mass of element 114 isotopes to be measured with outstanding resolution, thereby confirming their discovery. In order to increase the production rate of element 114 nuclides produced via the {sup 244}Pu+{sup 48}Ca reaction, a new target technology was required. Instead of a traditional thin actinide target, the MASHA separator required a thick, ceramic-based Pu target that was thick enough to increase element 114 production while still being porous enough to allow reaction products to migrate out of the target and travel through the separator to the detector array located at the back end. In collaboration with UNLV, we began work on development of the Pu target for MASHA. Using waste-form synthesis technology, we began by creating zirconia-based matrices that would form a ceramic with plutonium oxide. We used samarium oxide as a surrogate for Pu and created ceramics that had varying amounts of the starting materials in order to establish trends in material density and porosity. The results from this work are described in more detail in Refs. [1,4,10]. Unfortunately, work on MASHA was delayed in Russia because