Engineering Papers⌕ Search

SEARCH · Engineering Papers

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 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Structural Changes in Molten Salt Fuel and/or Waste Stream Compounds Cs 2 UCl 6 and Cs 2 UO 2 Cl 4 from Room Temperature to Melting and Related Materials [Slides]

Perovskite phases have been explored in recent literature as waste forms for long term storage of radioactive waste from molten salt reactors specifically. Synthesis of new halide perovskite phases can inform these efforts as well as molten salt reactor chemistry in general. Inorganic perovskites have also shown promise as new scintillating materials for detection of X-rays and gamma rays. Modifying the structure and chemistry of these compounds, perhaps with organic components, could be used to synthesize new scintillating compounds or modify the properties of existing ones. Both U 6+ and U 4+ readily form compounds with a striking similarity to vacancy ordered perovskites (Cs 2 UO 2 Cl 4 and Cs 2 UCl 6 ). Each of these compounds is relevant to molten salt reactor chemistry and/or waste streams. These phases could from in uranium chloride salt systems or waste streams: Cs being a fission product.

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

Multicomponent pyrochlore solid solutions with uranium incorporation – A new perspective of materials design for nuclear applications

Multicomponent pyrochlore solid solutions with and without uranium incorporation were fabricated and their thermal-mechanical properties were characterized. Multicomponent pyrochlore solid solutions without uranium exhibit comparable thermal conductivity and higher mechanical strength compared to baseline single component rare-earth titanate pyrochlore (A 2 Ti 2 O 7 ). Uranium incorporation reduces hardness as compared with single component compositions. High entropy pyrochlore with uranium displays the highest thermal conductivity within multicomponent pyrochlore solid solutions with significantly better mechanical properties than UO 2 . The measured thermal conductivity correlates well with A-site cation mixing entropy and a modified size disorder parameter, and thus the size disorder and mixing entropy could be good indicators for predicting thermal conductivity of multicomponent pyrochlore solid solutions. This work opens up the possibility of designing multicomponent oxide solid solutions by controlling their chemical disorder/mixing entropy to achieve acceptable thermal-mechanical properties, desired radiation and corrosion performance for potential nuclear waste form and inert matrix fuel applications.

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

Zero Valent Iron for Reductive Removal of Technetium-99 from Aqueous Sulfate Solutions - 20347

This research investigates the reductive removal of technetium-99 ({sup 99}Tc) by zero valent iron (ZVI) from low activity waste (LAW) off-gas condensate simulant as a secondary treatment after recovery from vitrification planned at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Due to its high volatility, only a fraction of Tc{sup (VII)} will be incorporated into glass waste forms. Volatilized Tc will be captured by an off-gas treatment system with current plans to recycle off-gas condensate back to the vitrification facility. The scheme with off-gas recycling will increase Tc loading in the glass waste, but will also increase the concentrations of sulfate, halides, and other problematic constituents impeding overall LAW processing and increasing volume of the glass product. This study focuses on Tc{sup (VII)} removal by ZVI as a feasible pathway to minimize off-gas condensate recycling, which may result in a reduction in the volumes of LAW waste to be immobilized and subsequent cost savings. ZVI is an established treatment agent for redox-active contaminants such as trichloroethylene, nitrate, arsenic, chromium, phenol, and others. It is a commercially available and cost-effective material. Our previous experiments showed that ZVI is very efficient for the reductive removal of {sup 99}Tc. In this work we studied ZVI oxidation under aerobic conditions at pH 7 with 0.1 M Na{sub 2}SO{sub 4} solution (ionic strength, IS, 0.3 M) in the presence and absence of {sup 99}Tc. The concentration of {sup 99}Tc and the changes in pH, dissolved oxygen (DO), and oxidation-reduction potential (ORP) of the simulated solutions were monitored over 8 days. The formation of iron oxide phases was probed by X-ray diffraction (XRD). Obtained results suggest that ZVI contact time with {sup 99}Tc containing solutions for 6 hours resulted in nearly complete removal of {sup 99}Tc from the aqueous phase. XRD analysis showed that the oxidation of ZVI is rapid with formation of magnetite (Fe{sub 3}O{sub 4}) with minor percentage of goethite and maghemite. This work is a part of a larger set of studies investigating the feasibility of {sup 99}Tc reductive removal from the LAW off-gas condensate. (authors)

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

Criticality Consideration for Geologic Disposal of DOE-Owned Spent Nuclear Fuels - 20255

Decisions must be made relative to packaging spent nuclear fuel (SNF) for storage, transportation, and eventual disposal. This paper evaluates potential criticality-related loading and packaging considerations for Department of Energy (DOE)-owned SNF that could be impacted favorably or unfavorably by disposal uncertainties and proposes approaches to minimize such impacts. Using Al-Based SNF as an example and drawing on analyses performed in support of the Yucca Mountain Repository License Application (LA), this paper explores the role of natural barriers, engineered barriers, and waste form characteristics relative to reducing the potential for postclosure criticality. Several parameters important to postclosure criticality are considered in order to determine which are useful discriminators to support packaging decisions. Using these parameters, a qualitative assessment is performed relative to three options for managing DOE-owned SNFs. (authors)

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

Update on Development of a U.S. Rail Transport Capability for Spent Nuclear Fuel and High-Level Waste - 20466

This paper provides an overview of the progress to date, and discussion of the path forward, related to designing, fabricating and testing prototype railcars that will comply with the safety standard S-2043. This standard was developed by the Association of American Railroads (AAR) specifically for railcars used to transport High-Level Radioactive Material (HLRM). AAR defines the term HLRM to include both spent nuclear fuel (SNF) and high-level radioactive waste (HLW). DOE is in the process of developing and testing prototype railcars that will satisfy Standard S-2043. This is a technical paper that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not 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 governs the obligations of the parties, and this paper in no manner supersedes, overrides, or amends the Standard Contract. This paper reflects technical work which could support future decision making by the Department of Energy (DOE or Department). No inferences should be drawn from this paper regarding future actions by DOE, which are limited both by the terms of the Standard Contract and a lack of Congressional appropriations for the Department to fulfill its obligations under the Nuclear Waste Policy Act including licensing and construction of a spent nuclear fuel repository. (authors)

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

The TRANSCEND University Consortium: Theme 4: Nuclear Materials - 20431

The safe and secure management of Pu is a matter of international concern, with ∼250 t of separated Pu currently stockpiled worldwide. The UK's civil inventory of nuclear materials contains significant stocks of separated Pu from the reprocessing of Magnox and AGR spent fuels. The preferred option for the 138.5 tonnes of Pu is re-use as mixed oxide (MOx) fuel, although 5% is not suitable for re-use and is recommended for direct disposal. However, it will take more than 15 years to implement re-use, requiring that the Pu be kept in interim storage in its current state for that period, i.e. as PuO{sub 2} powder within inert steel storage cans at Sellafield. The focus of the work presented here is thus plutonium storage and the direct disposal of plutonium. The Research and Development needs of both are now pressing: in the case of storage due to it being the current default; in the case of immobilization and disposal because of a comparative lack of Research and Development on Pu conditioning and packaging due to policy uncertainty as to whether it would be disposed of in a Geologic Disposal Facility (GDF). Addressing these needs is complicated by Pu's high radioactivity, decay heat and radiotoxicity, criticality, nuclear safeguard requirements and, for some UK Pu contaminated materials targeted for disposal, poor inventory. Thus, there is also a critical requirement for underpinning research on Pu bearing materials in these two contexts. In response to these needs, the TRANSCEND Consortium (Transformative Science and Engineering for Nuclear Decommissioning, a multi-disciplinary collaboration of 11 universities and 8 key industry partners from across the UK's civil nuclear sector) is seeking to provide technical underpinning to ongoing option development for the UK's civil Pu stockpile. Whilst understanding the behaviour of plutonium during its re-use as MOx is beyond the scope of the TRANSCEND Consortium work plan, the main objectives of the work are: (1) For interim storage: to understand how the surface structure and properties of pristine and radiation damaged PuO{sub 2} change with time in the absence and presence of water; and (2) For immobilization and disposal: to understand the mechanisms of incorporation of Pu into ceramic and glass-ceramic waste-forms, as well as the effect on these of self-induced radiation damage. Each objective is being addressed through separate work packages, the details of which are discussed in this paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Compositional effects on the chemical durabilities of aluminophosphate glasses: A review

Phosphate glasses have a range of applications including as hosts for immobilizing radioactive wastes. Studies have shown that addition of iron and/or aluminum oxides can drastically improve the chemical durability of phosphate glasses where the accurate measurement of chemical durability is one of the most important factors for determining the long-term viability of a given waste form. However, due to inconsistencies with the experimental methods used to generate chemical durability data, comparing and interpreting such data is a tedious task. These variables include the temperature of the test, the specimen form (e.g., coupon, particles), the pressure of the test (e.g., atmospheric pressure, elevated pressure in an autoclave), the exposure time, the exposure medium, and how the loss is documented (e.g., total mass lost, normalized elemental release). This review paper summarizes a large collection of chemical durability tests on aluminophosphate glasses in various studies. In addition, the effects of different oxides on the properties of phosphate glasses are summarized.

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

Enhanced crevice corrosion of stainless steel 316 by degradation of Cr-containing hollandite crevice former

This study explores the corrosion interactions between stainless steel 316 (SS316), a candidate nuclear waste canister material, and Cr-containing hollandite (Ba 1.15 Cr 2.3 Ti 5.7 O 16 ), a model ceramic nuclear waste form. The two materials were corroded in close proximity in 0.6 M NaCl solution at 90 °C for 28 days. Severe crevice corrosion was observed on the stainless steel surface after the corrosion experiment, which is likely associated with the rapid accumulation of Cr 3+ cations originating from both SS316 and the Cr-containing hollandite. Furthermore, the enhanced corrosion of SS316 by the corrosion products of hollandite is quantitatively explained using a classical crevice corrosion model.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Review of recent developments in iodine wasteform production

Radioiodine capture and immobilization is not only important to consider during the operation of reactors (i.e., I-131), during nuclear accidents (i.e., I-131 and I-129) or nuclear fuel reprocessing (i.e., I-131 and I-129), but also during disposal of nuclear wastes (i.e., I-129). Most disposal plans for I-129-containing waste forms (including spent nuclear fuel) propose to store them in underground repositories. Here, iodine can be highly mobile and, given its radiotoxicity, needs to be carefully managed to minimize long-term environmental impacts arising from disposal. Typically, any process that has been used to capture iodine from reprocessing or in a reactor is not suitable for direct disposal, rather conversion into a wasteform for disposal is required. The objectives of these materials are to use either chemical immobilization or physical encapsulation to reduce the leaching of iodine by groundwaters. Some of the more recent ideas have been to design capture materials that better align with disposal concepts, making the industrial processing requirements easier. Research on iodine capture materials and wasteforms has been extensive. This review will act as both an update on the state of the research since the last time it was comprehensively summarized, and an evaluation of the industrial techniques required to create the proposed iodine wasteforms in terms of resulting material chemistry and applicability.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
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 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 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 OSTI2020

Composite Analysis Solid Waste Release Data Reduction of Activity Flux from Waste Sites to the Vadose Zone for Baseline Assessment

This environmental calculation file (ECF) describes the data reduction methodology for the solid waste release datasets generated by the Composite Analysis Solid Waste Release (CASWR) model (CP-62766, Model Package Report: Composite Analysis Solid Waste Release Model [CASWR Model]) and documents the resulting reduced datasets for the updated Composite Analysis (CA) baseline assessment. This ECF is in support of the updated Hanford Site CA. The purpose of the updated Hanford Site CA is to provide an estimate of the cumulative radiological impacts from active and planned low-level radioactive waste disposals and other potentially interacting radioactive waste disposal sources that will remain following Hanford Site closure. The vadose zone modeling activities associated with the CA includes solid waste forms as contaminant sources. To support the modeling of solid waste transport from the vadose zone to the groundwater at the Hanford Site, the CASWR model was designed to generate deterministic radionuclide release rates for the Hanford Site Central Plateau solid waste disposal sites. The complete CASWR model-generated dataset for the CA Baseline Assessment consists of 2,378,831 sets of time and release rates (1,073,869 sets for the 200 East Area and 1,304,962 sets for the 200 West Area), representing a course of the simulation from year 1953 to year 12,069, 16 radiological contaminants of potential concern (COPCs), and 237 solid waste sources as documented in ECFHANFORD- 19-0112, Solid Waste Release Calculations for the Composite Analysis Baseline Assessment. For several of the STOMP vadose zone models, the number of release rate data pairs required for release at all waste sites would exceed the limitations imposed by the STOMP modelling software package used to simulate contaminant transport through the vadose zone. As a result, a data reduction of the CASWR dataset is required prior to its use as a source input for the CA vadose zone modeling activities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Affinity of LDR Organics to Cementitious Materials and Activated Carbon

The Hanford site stores approximately 56 million gallons of radioactive and hazardous waste in underground storage tanks. The current low-activity waste (LAW) vitrification facility does not have the capacity to immobilize all of the LAW. Therefore, cementitious waste forms are being evaluated as a cost-effective supplemental waste treatment, focusing on the solidification/immobilization of Land Disposal Restricted (LDR) organics, which is currently not recognized by the Environmental Protection Agency (EPA) as a standard treatment method (40 CFR 268.42). Previous efforts by the EPA and waste management agencies have highlighted the potential for cementitious materials to retain certain organic species through physical (encapsulation) and chemical (sorption) interactions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Structural Origins of High MoO 3 Solubility in Peraluminous Borosilicate Glasses

Molybdenum (Mo) imposes strict loading limits in conventional borosilicate nuclear waste glasses due to the tendency of tetrahedral molybdate [MoO 4 ] 2− species to phase-separate and crystallize as alkali molybdates. Here, we demonstrate an unprecedented 13.96 wt % (7.51 mol %) MoO 3 solubility in peraluminous sodium aluminoborosilicate glasses a ∼15× increase over their peralkaline counterparts. Using Raman spectroscopy, multinuclear and dipolarcorrelation magic angle spinning nuclear magnetic resonance (MAS NMR), electron paramagnetic resonance (EPR), and scanning transmission electron microscopy (STEM)-energy dispersive spectroscopy (EDS), we reveal that Nadeficient, low optical basicity conditions stabilize octahedral MoO 6 units, which polymerize into molybdite-like Mo−O clusters dispersed within the glass matrix. These Mo-rich clusters suppress the formation of depolymerized [MoO 4 ] 2− environments typically responsible for Na 2 MoO 4 precipitation and instead promote the formation of Na 2 Mo 2 O 7 as the saturation phase. Concurrently, Mo solubility drives the conversion of AlO 4 − to higher-coordination AlO 5 species, liberating Na + that is subsequently sequestered in molybdate-rich domains. The combined evolution of Mo coordination, modifier redistribution, and network depolymerization provides a mechanistic basis for the markedly enhanced Mo solubility in peraluminous compositions. These findings establish new structural guidelines for designing aluminoborosilicate waste forms with substantially greater capacity to incorporate Mo-rich nuclear waste streams.

Amorphous materials↗
DOE OSTI2020

Reactions during conversion of simplified low-activity waste glass feeds

The mechanisms that affect the incorporation of 99Tc, a volatile radioactive component of concern, into glass melt during vitrification of low-activity waste (LAW) are being investigated to develop the method to increase the retention of 99Tc in glass waste form. Previous studies with simulated LAW glass feeds (slurry mixture of liquid waste and chemical/mineral additives) demonstrated that the early stage feed-to-glass conversion reactions below 800°C are critical for the Re (used as a nonradioactive surrogate of 99Tc) retention in glass. To examine the effect of feed composition on the feed-to-glass conversion reactions, simplified systems containing major LAW components (NaNO3 and NaOH) and representative additive components (SiO2 and H3BO3) were designed and tested. The ratio of H3BO3 to NaNO3 was varied in three-component system without NaOH and that of NaOH to NaNO3 was varied in the four-component system at a fixed H3BO3 to NaNO3 ratio. As a first step of testing with simplified feeds, this study applied thermal analyses and phase characterization of the reacting feeds, which were performed without the addition of Re, to investigate the evolution of salt phases during slurry drying process and upon heating of dried feeds.

Jin, Tongan↗
DOE OSTI2022

Follow-on Report of Analysis of Approaches to Supplemental Treatment of Low-Activity Waste at the Hanford Nuclear Reservation (Vol. I)

The Hanford Site, in southeast Washington State, is preparing to disposition approximately 56,000,000 gallons (56 Mgal) of radioactive and chemically hazardous wastes currently stored in underground tanks at the site. Tank wastes will be divided into a high-activity fraction and a low-activity fraction for subsequent treatment and disposition. A waste processing and treatment facility, the Waste Treatment and Immobilization Plant (WTP), will include the high-level waste (HLW) vitrification facility (WTP HLW Vitrification Facility) for immobilizing the high-activity fraction and a low-activity waste (LAW) vitrification facility (WTP LAW Vitrification Facility) for immobilizing the low-activity fraction. Both facilities will use vitrification technology to immobilize the Hanford tank wastes in a glass waste form. The volume of LAW to be treated and disposed of following waste retrieval and WTP operations will exceed the planned processing capacity of the WTP LAW Vitrification Facility. ORP-11242,-River Protection Project System Plan, estimates a shortfall in LAW treatment capacity of approximately 56 Mgal, approximately 50% of the projected LAW volume. To maintain the planned tank waste processing mission schedule, the U.S. Department of Energy (DOE) will require additional LAW treatment capacity (termed “supplemental LAW”) external to the WTP process. LAW must be solidified by a treatment technology before the waste can be permanently disposed of in an approved DOE on-site disposal facility or a commercial (state or U.S. Nuclear Regulatory Commission [NRC-licensed]) off-site mixed low-level waste disposal facility. A decision on the approach to supplemental LAW treatment, processing, and disposal has not yet been made

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Solidification of REDC Organics For Disposal As Solid Waste

The Radiochemical Engineering Development Center (REDC) generates various liquid organic wastes from processing irradiated targets to recover heavy elements. In the past, these organics were discharged to the Oak Ridge National Laboratory (ORNL) liquid low-level waste (LLLW) system, along with the aqueous waste. Because of a reduction in aqueous LLLW from other ORNL generators and an increase in the radionuclide concentration, particularly 238 Pu, in the REDC organic waste stream, the organics can no longer be discharged to the LLLW system. The plan is to solidify the liquid organic waste for disposal as solid waste. Solidifying the organic liquids using PM-199 Organoclay ® would produce solid waste forms with no free liquid, which should qualify for disposal at the Waste Isolation Pilot Plant. Granular Organoclay can be added to the organic waste solutions until there is a dry layer on top of the liquid and then allowed to cure for a few hours to produce a solid wasteform.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗