Search NASA⌕ Search

SEARCH · Search NASA

Results for “Waste Forms”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 775 records · Page 43
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Synthesis of apatite phosphates containing Cs{sup +}, Sr{sup 2+} and RE{sup 3+} ions and chemical durability studies

Graphical abstract: New chemically durable hydroxy and fluoro apatite phosphate wasteforms CsLa{sub 0.8}Nd{sub 0.1}Sm{sub 0.1}Sr{sub 3}(PO{sub 4}){sub 3}OH and CsLa{sub 0.7}Nd{sub 0.1}Sm{sub 0.1}Eu{sub 0.1}Sr{sub 3}(PO{sub 4}){sub 3}F are fabricated. Display Omitted Highlights: ► Novel apatite phosphates CsLaSr{sub 3}P{sub 3}(PO{sub 4}){sub 3}OH and CsLaSr{sub 3}(PO{sub 4}){sub 3}F are reported. ► New hydroxyapatite wasteform CsLa{sub 0.8}Nd{sub 0.1}Sm{sub 0.1}Sr{sub 3}P{sub 3}O{sub 12}OH is fabricated. ► Fluoro analogue CsLa{sub 0.7}Nd{sub 0.1}Sm{sub 0.1}Eu{sub 0.1}Sr{sub 3}P{sub 3}O{sub 12}F incorporates ‘Eu{sup 3+}’. ► Negligible leachability of Cs and rare earth ions demonstrated. -- Abstract: Crystallization of Cs, Sr and selected rare earth ions in apatite phosphate structure was investigated. Compounds of the formula CsLaSr{sub 3}(PO{sub 4}){sub 3}OH (CSOH), CsLaSr{sub 3}(PO{sub 4}){sub 3}F (CSF) and the simulated waste forms of the compositions CsLa{sub 0.8}Nd{sub 0.1}Sm{sub 0.1}Sr{sub 3}(PO{sub 4}){sub 3}OH (CNSOH) and CsLa{sub 0.7}Nd{sub 0.1}Sm{sub 0.1}Eu{sub 0.1}Sr{sub 3}(PO{sub 4}){sub 3}F (CNSEF) were synthesized and characterized by powder XRD, TGA and SEM techniques. The leachability behaviour of the single phases was studied for a period of one month by MCC-5 test using powder specimen. The leachability of Cs and other rare earth ions was found to be negligible. The leach rates (LR{sub i}) of Sr and P were observed as follows: CSOH – LR{sub Sr} = 10{sup −3}–10{sup −4} g/m{sup 2}/d, LR{sub P} = 10{sup −4} g/m{sup 2}/d; CSF – LR{sub Sr} = 10{sup −3}–10{sup −5} g/m{sup 2}/d, LR{sub P} = 10{sup −4}–10{sup −5} g/m{sup 2}/d; CNSOH and CNSEF – LR{sub Sr} = 10{sup −3}–10{sup −4} g/m{sup 2}/d, LR{sub P} = 10{sup −4}–10{sup −5} g/m{sup 2}/d.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Structure and thermodynamics of calcium rare earth silicate oxyapatites, Ca 2 RE 8 (SiO 4 ) 6 O 2 (RE = Pr, Tb, Ho, Tm)

Calcium rare earth silicate oxyapatites, (Ca 2 RE 8 (SiO 4 ) 6 O 2 ), are of interest as components of glass-ceramic nuclear waste forms. To assess their long-term behavior in a geologic repository, it is essential to determine their structure and thermodynamic stability at relevant conditions. Here, in this work, we performed detailed structural and thermodynamic investigations on Ca 2 Pr 8 (SiO 4 ) 6 O 2 , Ca 2 Tb 8 (SiO 4 ) 6 O 2 , Ca 2 Ho 8 (SiO 4 ) 6 O 2 , and Ca 2 Tm 8 (SiO 4 ) 6 O 2 by high energy synchrotron powder X-ray diffraction combined with Rietveld analysis and high temperature oxide melt drop solution calorimetry. Enthalpies of formation from constituent oxides (ΔH f,ox ) were determined to be -765.1 ± 22.8 kJ/mol for Ca 2 Pr 8 (SiO 4 ) 6 O 2 ; -638.9 ± 20.5 kJ/mol for Ca 2 Tb 8 (SiO 4 ) 6 O 2 ; -643.3 ± 10.3 kJ/mol for Ca 2 Ho 8 (SiO 4 ) 6 O 2 ; and -403.2 ± 5.1 kJ/mol for Ca 2 Tm 8 (SiO 4 ) 6 O 2 . These thermodynamic parameters were used in assessing the thermochemical stability of these phases in the presence of water vapor from room temperature to 600 K, as encountered in the subsurface environments of a geological repository.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Analysis of Selected Legacy 85 Kr Samples

Legacy samples composed of 85 Kr encapsulated in solid zeolite 5A material and five small metal tubes containing a mixture of the zeolite combined with a glass matrix resulting from hot isostatic pressing have been preserved. The samples were a result of krypton R&D encapsulation efforts in the late 1970s performed at the Idaho Chemical Processing Plant. These samples were shipped to Oak Ridge National Laboratory (ORNL) in mid-FY 2014. Upon receipt the outer shipping package was opened, and the inner package, removed and placed in a radiological hood. The individual capsules were double bagged as they were removed from the inner shipping pig and placed into individual glass sample bottles for further analysis. The five capsules were then x-ray imaged. Capsules 1 and 4 appear intact and to contain an amorphous mass within the capsules. Capsule 2 clearly shows the saw marks on the capsule and a quantity of loose pellet or bead-like material remaining in the capsule. Capsule 3 shows similar bead-like material within the intact capsule. Capsule 5 had been opened at an undetermined time in the past. The end of this capsule appears to have been cut off, and there are additional saw marks on the side of the capsule. X-ray tomography allowed the capsules to be viewed along the three axes. Of most interest was determining whether there was any residual material in the closed end of Capsule 5. The images confirmed the presence of residual material within this capsule. The material appears to be compacted but still retains some of the bead-like morphology. Based on the nondestructive analysis (NDA) results, a proposed path forward was formulated to advance this effort toward the original goals of understanding the effects of extended storage on the waste form and package. Based on the initial NDA and the fact that there are at least two breached samples, it was proposed that exploratory tests be conducted with the breached specimens before opening the three intact capsules. Portions of these would be analyzed to determine the fraction of krypton/xenon remaining in the matrix and the amount of rubidium remaining in the matrix. The inner surface of the breached capsules would be examined for corrosion. The materials contained in Capsules 2 and 5 have been examined. There appears to be a relatively uniform distribution of Kr and Rb throughout the pellets examined. The chemical composition of the pellets appears to be consistent with 5A molecular sieves. The material contained within Capsule 5 showed ~1 at. % lead. The origin of the Pb is currently indeterminate. X-ray diffraction analysis shows a significant shift from the 5A structure, most likely due to the Kr encapsulation / sintering process that occurred when the samples were made. The capsule walls were also examined and showed extensive corrosion throughout. Elemental mapping of the capsule material appeared consistent with carbon steel, while the weld material appeared consistent with a stainless steel. The interior surface of the capsule appeared to have a layer of material containing Al, Si, and Ca similar to the 5A molecular sieve. Analysis for Rb within the corrosion sites was inconclusive.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

K-Basins Sludge Treatment and Packaging at the Hanford Site - 13585

Highly radioactive sludge resulting from the storage of degraded spent nuclear fuel has been consolidated in Engineered Containers (ECs) in the 105-K West Storage Basin located on the Hanford site near the Columbia River in Washington State. CH2M Hill Plateau Remediation Company (CHPRC) is proceeding with a project to retrieve the sludge, place it in Sludge Transport and Storage Containers (STSCs) and store those filled containers within the T Plant Canyon facility on the Hanford Site Central Plateau (Phase 1). Retrieval and transfer of the sludge material will enable removal of the 105-K West Basin and allow remediation of the subsurface contamination plumes under the basin. The U.S. Department of Energy (DOE) plans to treat and dispose of this K Basins sludge (Phase 2) as Remote Handled Transuranic Waste (RH TRU) at the Waste Isolation Pilot Plant (WIPP) located in New Mexico. The K Basin sludge currently contains uranium metal which reacts with water present in the stored slurry, generating hydrogen and other byproducts. The established transportation and disposal requirements require the transformation of the K Basins sludge to a chemically stable, liquid-free, packaged waste form. The Treatment and Packaging Project includes removal of the containerised sludge from T Plant, the treatment of the sludge as required, and packaging of all the sludge into a form that is certifiable for transportation to and disposal at WIPP. Completion of this scope will require construction and operation of a Sludge Treatment and Packaging Facility (STPF), which could be either a completely new facility or a modification of an existing Hanford Site facility. A Technology Evaluation and Alternatives Analysis (TEAA) for the STP Phase 2 was completed in 2011. A Request for Technology Information (RFI) had been issued in October 2009 to solicit candidate technologies for use in Phase 2. The RFI also included a preliminary definition of Phase 2 functions and requirements. Potentially applicable technologies were identified through a commercial procurement process, technical workshops, and review of the numerous previous sludge treatment technology studies. The identified technology approaches were screened using the criteria established in the Decision Plan, and focused bench top feasibility testing was conducted. Engineering evaluations of the costs, schedules, and technical maturity were developed and evaluated. Recommendations were developed based on technical evaluations. The criteria used in the evaluation process were as follows: (1) Safety, (2) Regulatory/stakeholder acceptance, (3) Technical maturity, (4) Operability and maintainability, (5) Life cycle cost and schedule, (6) Potential for beneficial integration with ongoing STP-Phase 1 activities, and (7) Integration with Site-wide RH-TRU processing/packaging, planning, schedule, and approach. The TEAA recommended Warm Water Oxidation (WWO) as the baseline treatment technology and two risk reduction enhancement options for further consideration during development of the process - size reduction and chemical oxidation (Fenton's reagent). The enhancement options would potentially allow a useful reduction in the total operating time required to process the K Basins sludge. The U.S. Department of Energy's Richland Field Office (DOE-RL) has approved this recommended technical approach. The baseline process can be broken down into the following main process steps: (1) STSC transfer from T Plant to the Sludge Treatment and Packaging Facility (STPF). (2) Retrieval of sludge from the STSCs and transfer to the Receipt and Reaction Tank (RRT). (3) Preparation for immobilization by oxidation using heated water (i.e., WWO) for those batches that require it and concentration by evaporating water at about atmospheric pressure in the RRT. (4) Immobilization by using additives to eliminate free liquids and packaging of the treated sludge into drums. (5) Inspection and handling of the filled drums prior to transfer to a separate storage and shipping facility. (6) Handling of vapor, condensate, and other waste streams generated by the process. Each of these steps is discussed in the paper, together with the current state of progress in developing the technology and requirements for continued development. A schematic of the recommended baseline WWO treatment process is given below. (authors)

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

An Initial Study of Uranium Morphology in Fused LiCl-KCl

Molten salt electrorefining is one method by which reusable constituents within used nuclear fuel can be recovered for recycle and especially radiotoxic isotopes removed for incorporation into tailored waste forms. In these processes, irradiated metallic uranium or uranium alloys are dissolved at an electrochemical anode and crystals of purified uranium are collected at a cathode. The morphology of the deposited crystals can impact a variety of important factors, including the adhesion characteristics of the deposit to the cathode surface and the amount of salt which is macroscopically and microscopically occluded with the deposit. In addition, future applications such as direct creation of nuclear fuels by electrochemical deposition might be enabled if the science and fundamentals of uranium deposition morphology were thoroughly understood. A reasonable amount of experience has been collected from uranium electrorefining within the range of typical plant conditions. However, only limited laboratory study has been performed to define and understand the range of uranium crystal morphologies under repeatable laboratory conditions. The present study was an initial step to define these morphologies and transitions. The deposit from each deposition potential was primarily composed of a different morphology with little crossover of that morphology into the other deposition potentials. The spacing of these experiments (overpotentials of 25, 150 and 450 mV) is wide enough that additional dominant morphologies could exist. In addition, while the majority of the deposit from each potential was comprised of a particular morphology, minor morphologies were also found within the deposit. This may indicate the presence of an additional dimension of complexity, namely uranium concentration. Additional experimental and theoretical studies would be valuable to elucidate this subject. Additional results will be presented and discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Predicting Saturated Hydraulic Conductivity over Time for Degrading Saltstone Vault Concrete - 20376

At the Savannah River Site (SRS), low-level liquid tank waste is mixed with cementitious materials to produce a solid, monolithic waste form referred to as saltstone. The liquid saltstone mixture is pumped into massive concrete structures referred to as saltstone disposal units (SDUs) for curing and long-term storage. SDUs are vaults made of high performance concrete forming a barrier between saltstone and the surrounding environment. Predicting the long-term performance of SDU vault concrete, as well as the saltstone material itself, is critical to understanding the potential releases of radionuclides to the surrounding environment as described in the Saltstone Disposal Facility (SDF) performance assessment (PA). The hydraulic conductivity profile for SDU vault concrete was conservatively assumed to degrade linearly from the initial saturated hydraulic conductivity of the intact concrete over a long period of time. The initial hydraulic properties of saturated SDU vault concrete have been measured, and the final hydraulic properties of degraded vault concrete are assumed to be the same as the porous backfill material that will surround the vault after closure. The assumed final condition of the vault concrete is also conservative in that it provides a greater estimate of water penetration through the vault than would generally be expected. Therefore, insights that provide more realistic but conservative estimates as to how degradation processes may change the hydraulic properties of the SDU vault concrete, the time dependent relationship of the vault concrete hydraulic properties to degradation (and thus time), and the final hydraulic properties of the vault concrete provides useful additions to the SDF PA. This paper focuses on the assumption that the saltstone vault concrete saturated hydraulic conductivity would decrease linearly as a function of degradation time; this assumption resulted from a worst-case, weighted (by relative layer thickness), arithmetic average of intact and backfill material hydraulic conductivities representing flow parallel to a layered system. This research indicates that the effective saturated hydraulic conductivity of the degraded SDU vault concrete should be based on the relative thickness-weighted geometric mean of the intact and degraded layers hydraulic conductivities. This alternative, which is supported extensively in the relevant literature and evaluated in this paper, leads to a more realistic and defensible estimate of the effective saturated hydraulic conductivity in degraded vault concrete while remaining conservative. The impact of using the more realistic estimate of the effective saturated hydraulic conductivity for the degraded concrete decreases the effective saturated hydraulic conductivity by up to several orders of magnitude for early times (i.e., when the SDU vault concrete is assumed degraded via sulfate attack and carbonation). This revised estimate of the effective saturated hydraulic conductivity has been implemented in the recently revised SDF PA. (authors)

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

Dynamic Leaching Method for Intact Salt-stone Samples - 18495

At the Savannah River Site (SRS) low activity salt solution is stabilized via encapsulation within a grout termed salt-stone. Salt-stone is emplaced into large (multi-million liter) concrete storage containers referred to as Salt-stone Disposal Units (SDUs). Technetium-99 (Tc-99) and iodine-129 (I-129) are long-lived radionuclides contained in the low activity salt waste and subsequently incorporated into the grout waste form: both radionuclides are contributors to the potential long-term radiation exposure of the public and environment surrounding the SDUs. With respect to Tc-99, in the reducing, high pH environment within the grout, Tc-99 is expected to be relatively immobile since it exists in a reduced Tc(IV) oxidation state in the form of sparingly soluble hydrated oxides (TcO{sub 2}.xH{sub 2}O) or sulfides (TcS{sub x}). However, in the presence of O{sub 2} (associated with the future infiltration of air or oxygenated ground waters into the salt-stone monolith) it is possible for redox-sensitive Tc(IV) to transition into highly soluble (and mobile) Tc(VII) species, such as pertechnetate (TcO{sub 4}{sup -}), which are more readily transported to the surrounding environment. In contrast, I-129 is highly soluble and its leaching behavior is not anticipated to be significantly impacted by potential changes in redox environment or pH within salt-stone. Traditional approaches to quantifying the leaching behavior of Tc-99 and I-29 from cementitious matrices have involved partitioning experiments using size-reduced (crushed/ground) samples, and determination of the radionuclide fraction immobilized by the cementitious solids. Such experiments create artificially high solid-liquid contact areas that likely result in higher Tc-99 and I-129 leachate concentrations than would be expected for intact, monolithic samples. In the current study a new technique, termed the Dynamic Leaching Method (DLM), is being used to initially investigate the Tc-99 leaching behavior from monolithic salt-stone samples; the system will be used at a later date to evaluate the leaching characteristics of I-129. The data derived using this technique is intended to inform the SRS Salt-stone Disposal Facility (SDF) Performance Assessment (PA) which models the long-term transport of radionuclides from the SDUs to the environment. The DLM utilizes a flexible-wall permeameter to achieve saturated leaching under an elevated hydraulic gradient to simulate and accelerate the future transport of groundwater through salt-stone. Leachates are regularly analyzed with respect to radionuclide concentrations, key constituent concentrations (e.g., Al, Ca, Mg, Na, Si, CO{sub 3}{sup 2-}, NO{sub 2}{sup -}, NO{sub 3}{sup -}, SO{sub 4}{sup 2-}), pH, and reduction potential (E{sub h}) in order to develop an understanding of how chemical changes associated with the pore solution influence grout weathering and radionuclide leaching behavior. This paper provides a brief update with respect to data derived from preliminary Tc-99 leaching evaluations as the DLM was being developed, but the paper is predominantly focused on new sample data in which the DLM technique, pre- and post-DLM salt-stone sample characterization, and leachate analysis protocols have been optimized. (authors)

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

Chemical Complexity Controls Energy Dissipation and Defect Evolution

The development of metallic alloys is arguably one of the oldest sciences, dating back at least 3,000 years. Most research and applications have been focused on alloys with comprised multiple phases with one or two dominant elements, to which the addition of alloying elements in low concentrations leads to various performance improvements and changes in radiation resistance. In sharp contrast to traditional alloys, recent success in the synthesis of single phase concentrated solid solution alloys (SP-CSAs) has opened up new frontiers in materials research. In these alloys, a random arrangement of multiple elemental species on a lattice (fcc or bcc) results in unique site-to-site lattice distortions and local disordered chemical environments. Intense radiation in nuclear fission and fusion energy power systems, nuclear waste forms, high-energy accelerators and space exploration transfers energy to the electrons and atoms that make up the material, and thereby produces defects that ultimately compromise material strength and lifetime. A grand challenge in materials research is to understand complex electronic correlations and non-equilibrium atomic interactions, and how such intrinsic properties and dynamic processes affect energy transfer and defect evolution in irradiated materials. Since SP-CSAs possess unique links between intrinsic material properties (can be altered by alloy complexity), energy dissipation and various defect dynamic processes; they are ideal systems to fill knowledge gaps between electronic- /atomic-level interactions and radiation resistance mechanisms. In our work, we show that chemical disorder and compositional complexity in SP-CSAs have an enormous impact on defect dynamics through substantial modification of energy dissipation pathways. Based on a closely integrated computational and experimental study using a novel set of Ni-based SP-CSAs, we have explicitly demonstrated that increasing chemical disorder can lead to a substantial reduction in the electron mean free path and electrical and thermal conductivity. These reductions have a significant impact on energy dissipation and consequentially on defect evolution during ion irradiation. Considerable enhancement in radiation resistance with increasing chemical complexity from pure nickel to binary, and to more complex quaternary solid solutions, is observed under ion irradiation. Contrary to conventional alloys with low solute concentration but multiple phases, energy dissipation and defect evolution at the level of electrons and atoms in SP- CSA systems with extreme compositional disorder is an unexplored frontier in materials science. The integrated experimental/modeling effort provides new insights into defect dynamics at the level of atoms and electrons, and an innovative path forward towards solving a long-standing challenge in structural materials. Increasing chemical complexity achieved orders-of-magnitude reductions in electron mean free paths and electron and thermal conductivities of alloys of NiCo, NiFe and NiCoFeCr, compared to Ni, and suppressed damage accumulation under ion irradiation. This discovery provides insights to address the grand challenge of understanding complex electronic correlations and non-equilibrium atomic interactions, and the effect of these intrinsic properties and dynamic processes on energy transfer and defect evolution in irradiated materials. Understanding how material properties and defect dynamics can be tailored by alloy complexity may pave the way for new design principles of functional alloys. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2018

A Novel Mobile Examination and Remediation Facility for On-Site Remediation of Dry Storage Systems - 18389

Most of the spent nuclear fuel (SNF) in dry storage in the United States is safely and securely stored in dual-purpose (i.e., storage and transportation) canisters [1]. However, some of the SNF is stored in storage-only (i.e., non-transportable) packages, or packages that may become nonconforming for continued storage or transportation for a variety of reasons. Currently, the capability to remediate dry storage systems (DSSs) for SNF that are not transportable or that need to be repaired to maintain confinement boundary integrity during storage service does not exist without on-site access to a spent fuel pool. Hence, there is a potential future need for a mobile facility with remediation capabilities that can be used at multiple storage installations. These capabilities would permit needed repairs to an existing DSS to support continued storage or for the dry transfer of fuel to a new conveyance suitable for transport, if necessary. This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. A novel design concept for a Mobile Examination and Remediation Facility (MERF) is proposed with functions and capabilities to remediate a DSS. The proposed MERF would be a multi-purpose facility, remotely operated and computer-controlled with semi-automated equipment to allow trained operators to safely perform remediation activities while keeping worker radiation dose as low as reasonably achievable. It would include a shielded rotating platform with support towers and would use commercial off- the-shelf examination and repair equipment. A slip ring arrangement underneath the rotating platform would allow for the power supply without impeding the continuous rotation and would simplify cable management. The support towers would be mounted on the rotating platform, and each tower would have vertical and horizontal positioning capability for axial and radial tool location. The rotating platform would be attached to the main MERF structure. Capabilities would include visual, surface, and volumetric examinations to fully characterize physical conditions on canister surface before and after a repair; repair techniques, e.g., laser beam welding, friction stir welding, or cold spray, to perform needed remediation activities; and associated equipment for surface preparation such as cutting and grinding equipment. The proposed design would be simple, modular, robust, economically viable, and easily transportable and would have a small footprint. The modular design would present several advantages including compatibility with a variety of different DSSs, on-site assembly, and transportability; it also would be built in interchangeable sections to facilitate decontamination. Further, shielding would be supplemented through interface with the existing transfer cask and storage overpack. Robustness and canister stability would be achieved using pneumatic holders and steel support structures designed with vibration absorbers to reduce impacts and/or a 'pendulum' effect during a seismic event. Thus, the proposed MERF would be designed to: (i) minimize handling of an affected DSS to the extent practicable, (ii) provide radiation shielding and minimize dose to operators and the public during remediation activities, and (iii) enable examination, repair, and repackaging. This paper describes the proposed MERF design concept and capabilities, main components, equipment, different modes of operation, and operating procedures. Structural and safety assessments would need to be performed before final design, licensing, and deployment. The MERF is evaluated with respect to regulatory impact (first of a kind, prior experience), operational impact (safety, security, staffing, ongoing operations, constraints), worker dose (low, medium, high), cost (low, medium, high), schedule (duration, time sequence), and technology readiness level. (authors)

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

Effect of Curing Temperature and Composition on the Performance Properties of Saltstone - 12473

When freshly mixed Saltstone is placed into a vault at the Savannah River Site, the heat of hydration produced during curing causes the temperature of the grout to increase. It is not uncommon for the temperature of the grout to increase by 30 to 50 deg. C above the starting temperature depending on the vault size and location, environmental conditions and pour schedule. For example, temperatures measured by thermocouples in one cell have exceeded 70 deg. C. The elevated temperatures, once reached, can extend for weeks or even months. Recent studies at Savannah River National Laboratory (SRNL) have shown that curing simulated Saltstone mixes at elevated temperatures can lead to reduced compressive strength and dynamic Young's modulus as well as increased total porosity when compared to the same grout mixes cured at 20 deg. C. Samples cured at 60 deg. C can have hydraulic conductivities approximately 600 times greater than samples cured at 20 deg. C. Previous studies have shown that the 60 deg. C cured mix had a dynamic Young's modulus that was half that of the mix cured at 20 deg. C. However, high temperature curing may not always have negative effects on the performance properties of the grout. These results indicate that higher curing temperatures can significantly alter the performance properties that are inputs for the Saltstone Performance Assessment. The Saltstone formulation (mix design) must produce a grout waste form that meets both placement and performance properties. A statistically designed set of mixes was developed to determine key process and compositional factors that affect the performance properties of Saltstone grout. A total of 27 mixes were batched and tested. There are eight baseline mixes that contain high and low values of aluminate concentration, varying w/p ratio, and varying fly ash content in the premix. Each of the eight mixes was cured at 22, 40 and 60 deg. C. The three additional reference mixes batched at the beginning, middle and end of testing showed good reproducibility throughout the study. The results of varying the operational and compositional factors on the performance properties of Saltstone grout are: - The Young's modulus is inversely affected by the curing temperature of the grout. - The aluminate content of the salt solution increases the Young's modulus of the Saltstone for samples cured at 20 and 40 deg. C. - All grouts cured at 60 deg. C have low Young's moduli and the composition of the grout mix did not have any positive effects on the performance properties after curing at this temperature. - The fly ash content in the premix had not discernible effect on the Young's modulus. - The water to premix ratio does not affect the Young's modulus. - Grouts made with the high aluminate salt solution have lower hydraulic conductivities than those made with the low aluminate simulant. - Temperature has an inverse effect on the hydraulic conductivity. - In general, increasing the water to premix ratio and the fly ash content will increase the hydraulic conductivity. - The porosity and cured density of the Saltstone grout are primarily a factor of the water to premix ratio of the grout formulation. The correlation between hydraulic conductivity and Young's modulus investigated in previous studies was evaluated further in this report. The current data as well as data from previous studies was fitted by a second degree polynomial function with a 77.1 % R-squared value. This model can be used to estimate the hydraulic conductivity of a sample based its Young's modulus for a limited range (E values up to 7 GPa). More data points should be included to further define the relationship between these performance properties are high Young's moduli and low hydraulic conductivities. The Young's modulus should still be measured on grouts formulated in future studies to be used as screening method for samples to be sent for hydraulic conductivity measurements. Saltstone grout has a complicated microstructure that is affected by the composition of the salt solution, the dry feed mixture composition, and the conditions under which it cures. The performance properties investigated are interrelated, which can make it difficult to determine individual relationships between various factors. Porosity is a large factor in determining both the Young's modulus and the hydraulic conductivity of the samples. Although the cured density is a straightforward measurement, it is not a significant performance property that needs further investigation. The porosity measurement provides similar and more meaningful data in relation to other performance properties. Based on the results of this study, the aluminate content of the salt solution does have an effect on the Young's modulus and the hydraulic conductivity and its effects should continue to be investigated. However, a new projection of aluminate contents should be used since projections change over time. The variability of fly ash in the premix did not prove to be a contributing factor that effects the performance properties of Saltstone and does not need to be included in future studies unless significant deviations from the baseline mix are proposed. The water to premix ratio has a significant effect on the performance properties of interest and the effects of varying this factor need further investigation. It is recommended that smaller ranges of w/p are investigated in order to provide data that is more relevant to the Saltstone being placed in the vaults. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Iron polyphosphate glasses for waste immobilization

Abstract Iron polyphosphate glasses resist attack by water, can dissolve large concentrations of species that are otherwise insoluble in borosilicate glass melts, and can be processed at relatively low temperatures (1000‐1200°C), and so are viable hosts for vitrifying hazardous and radioactive wastes. The properties of iron polyphosphate glasses depend on the distributions of phosphate anions and the nature of the bonds between those anions and various metal polyhedra, and quantitative information can be obtained about those structures using a variety of spectroscopic, diffraction, and chromatographic techniques. This structural information helps explain compositional trends in properties, including dissolution rates, electrical conductivity, thermal properties and crystallization tendency. Studies of waste forms made with low activity and high level wastes are reviewed and related back to an understanding of the structures and properties of simpler iron polyphosphate glasses.

Brow, Richard K.↗
DOE OSTI2019

Environmental Assessment for Use of DOE-Owned High-Assay Low-Enriched Uranium Stored at Idaho National Laboratory

The U.S. Department of Energy (DOE) proposes to make available about 10 Metric Tons (MT) of High-Assay Low-Enriched Uranium (HALEU ) feedstock produced through the electrometallurgical treatment (EMT) process, and other small quantities of HALEU stored at Idaho National Laboratory (INL) for research development & demonstration in support of the commercial nuclear industry and government agencies, including use in advanced reactors. HALEU is a term applied to uranium that is enriched in the uranium-235 (U-235) isotope to a value that is 5% to 20% of the total uranium. Private sector advanced nuclear reactor designs and advanced nuclear fuel designs call for use of HALEU, but currently no commercial facility manufactures HALEU. DOE proposes to expand the fuel fabrication capability at INL to produce HALEU fuel at INL from 10 MT of HALEU feedstock to meet near-term needs. The production requires expansion of the fuel fabrication capability, including the purchase of new equipment and use of facilities at INL’s Materials and Fuels Complex (MFC) and the Idaho Nuclear Technology and Engineering Center (INTEC). DOE has made no decision on the specific use of the fuel. The fuel could be used in advanced reactor design or for the purpose of research and development. DOE would work with other Federal agencies and commercial vendors to determine use of HALEU fuel. Use of this HALEU, including its potential use in a future nuclear reactor, will be analyzed in future National Environmental Policy Act (NEPA) documents by the appropriate agency. The EMT process in operation at the INL converts sodium-bonded spent nuclear fuel into waste forms suitable for disposal as high-level waste and a HALEU product that is unsuitable for diversion to nuclear weapons but could be either reused in fuel or disposed of as waste. The Final Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel, DOE/EIS-0306, DOE 2000) discussed the EMT process, but did not make a decision on the disposition or use of the HALEU product from the EMT process. This Environmental Assessment (EA) addresses the HALEU product as HALEU feedstock in production of HALEU fuel.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2019

Performance Test and Evaluation on Polymer Form of Carbonized Waste - 19164

A carbonized waste is a material obtained by treating combustible wastes with superheated steam for volume reduction. A surrogate-carbonized waste was solidified to a monolith type of polymer form by using epoxy resin because it was obtained in powder form. The polymer form of carbonized waste was tested in accordance with standardized methods to evaluate its performance in terms of its mechanical integrity, durability, leaching properties, and microbiological stability for final disposal in Korea. Performance test results demonstrate that the polymer form of carbonized waste has excellent mechanical integrity and retains its structural integrity after thermal cycling, irradiation and 90-day water immersion. Leachability indices for cesium, cobalt, and strontium of the polymer form of carbonized waste were better than those required of a final disposal site in Korea. In addition, no culture growth of fungi or bacteria and no biodegradation were observed on the polymer form of carbonized waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Review of Mercury in the SRS Tank Waste Storage and Treatment Processes for Applicability to Hanford Waste Processing

The Savannah River Site (SRS) and Hanford produced large quantities of plutonium and other radioisotopes. The separations processes recovered plutonium and produced large quantities of High- Level Waste (HLW) that accumulated at each site in large, underground storage tanks. SRS has been retrieving the stored HLW, pretreating the waste for immobilization, and immobilizing the waste in glass or grout waste forms since the mid-1990s. Treatment of the waste at Hanford is expected to start in the early 2020s. Mercury is present in both the SRS and Hanford tank waste. SRS has recently completed an evaluation of mercury speciation throughout the tank waste treatment and immobilization flowsheet. This study was reviewed to determine what information from the study can be applied to the Hanford tank waste treatment program.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

In-House Alternative for Chemical Analyses of Rhenium and Sulfur in Glass

Accurate measurements of rhenium (Re) and sulfur (S) concentrations in solid samples are of critical importance for the key research activities at the Pacific Northwest National Laboratory (PNNL) Waste Form Development Laboratory (WFDL) related to Re behavior during waste glass melting process and solubility of S in glass. Southwest Research Institute (SwRI) has provided reliable services for chemical analysis needs at WFDL; however, its long turnaround time, occasionally up to 5 months, has caused some difficulties. An in-house method was developed for dissolving and fusing glass and feed samples in preparation for analyses by inductively coupled plasma-optical emission spectrometry and -mass spectrometry at a PNNL facility. Using a previously developed sodium hydroxide fusion method as a basis, four groups of samples from Hanford waste glass research were digested and fused at WFDL. The samples included 1) Enhanced low-activity waste glass correlation phase 2/3 melter glasses from the Vitreous State Laboratory at the Catholic University of America, 2) AN-102 dried feeds and glasses, 3) SS-AN-102 dried feeds and glasses, and 4) Integrated Research Project S solubility study glasses. We evaluated three different sets of solution preparation and analysis combinations, which included 1) a WFDL-prepared solution analyzed at SwRI, 2) a PNNLprepared solution analyzed at a laboratory in the Building-331at PNNL, and 3) an SwRIprepared solution analyzed at SwRI. We concluded that the PNNL in-house Re results can be used in parallel with the SwRI data based on the relative standard deviation values less than 6% from three different sets of solution preparation and analysis for all three groups of rhenium-containing samples tested. However, it was proposed that the in-house S data be used for the barium oxide-free glasses only because of the issue of barium sulfate precipitation in digestion/fusion solutions prepared from glasses containing barium oxide. In addition, higher analytical biases between the PNNL and SwRI S results are expected compared to the Re results, based on the relative standard deviation values of less than 10% from three different sets of solution preparation and analysis for two groups of S-containing samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Melter Tests to Define LAW Halide, Chromium and Phosphate Concentrations, VSL-14R3070-1 (February 2014)

The Office of River Protection is examining options to optimize the Low Activity Waste (LAW) Facility and LAW glass waste form. One option under evaluation is to enhance the waste processing rate of the vitrification plant currently under construction. It is likely that the capacity of the LAW vitrification plant can be increased incrementally by implementation of a variety of low-risk, high-probability changes, either separately or in combination. The present work addressed the potential formation of secondary sulfate phases while processing high waste loading LAW formulations with elevated halide, chromium, and phosphate feed concentrations over a wide range of LAW waste types, including those with high potassium concentrations. The present work will be followed by a second phase of testing, described in a separate Test Plan, the results of which will be reported elsewhere. The combined results from these tests will be used as inputs to develop concentration limits for these components in the LAW glass property-composition models and to support the modification and updating of the current WTP LAW glass formulation algorithm.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Improved High-Alkali Low-Activity Waste Formulations. VSL-15R3290-1, Rev. 0 (Final Report)

The Office of River Protection is examining options to optimize the Low Activity Waste (LAW) Facility and LAW glass waste form. One option under evaluation is to enhance the waste processing rate of the vitrification plant currently under construction. It is likely that the capacity of the LAW vitrification plant can be increased incrementally by implementation of a variety of low-risk, high-probability changes, either separately or in combination. The Vitreous State Laboratory (VSL) at The Catholic University of America (CUA) and EnergySolutions, Inc. have evaluated several of these potential incremental improvements for ORP in support of its evaluation of WTP LAW facility optimization. The present work was conducted using a Hanford LAW composition with low sulfur-to-sodium ratio with the objective of determining the maximum waste loading that can be achieved without Cr 2 O 3 addition by employing alternatives to control K-3 corrosion. LAW from Hanford Tank AN-105 and the glass formulation ORPLA20, which was previously developed to treat this waste, were selected as the base compositions for this study. As shown in Table 1.1, the ORPLA20 glass composition has a waste loading of 31.5 wt% and a sodium content of 24 wt% Na 2 O. Other major components contributed by the waste are Al 2 O 3 , SO 3 , Cl and K 2 O at concentrations of 5.55, 0.70, 0.67 and 0.53 wt%, respectively.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗