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At least 847 records · Page 47
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Tank 40 Sludge Batch 9 Waste Acceptance Product Specifications (WAPS) Solids and Chemical Analyses (2018)

A 3 L sample of sludge was collected from High Level Waste (HLW) Tank 40 (sample ID: HTF-40-18-9) at the Savannah River Site (SRS) and subsequently analyzed by the Savannah River National Laboratory (SRNL) in accordance with requirements for reporting the Waste Acceptance Product specification (WAPS) [1]. The collected sample was representative of Sludge Batch (SB) 9 that will be processed by the Defense Waste Processing Facility (DWPF) into a vitrified waste form.

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

Vitrification Tests with Potential Technetium Capture Streams

About 50 million gallons of high-level waste is currently stored in underground tanks at The United States Department of Energy’s (DOE’s) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE’s Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal. The tank waste will be separated into low- and high-activity waste fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High Level Waste (IHLW) products. The ILAW product will be disposed in an engineered facility on the Hanford site while the IHLW product is designed for acceptance at a national deep geological disposal facility for high-level nuclear waste. The ILAW and IHLW products must meet a variety of requirements with respect to protection of the environment before they can be accepted for disposal. The Low Activity Waste (LAW) and High-Level Waste (HLW) at Hanford will be vitrified in Joule Heated Ceramic Melters (JHCMs) to produce stable products for disposal. Technetium is an important radioactive component in the Hanford tank waste and is typically one of the principal contributors to risk in performance assessments for the ILAW disposal facility. Technetium is also one of the more volatile radionuclides and the retention of technetium in the glass under the high temperature conditions prevailing during vitrification can vary over a relatively wide range depending on the feed composition, feed chemistry, and melter operating parameters. In support of the WTP Project, the Vitreous State Laboratory (VSL) at The Catholic University of America (CUA) and EnergySolutions have developed and tested glass formulations for WTP waste envelopes to provide data to meet the WTP contract requirements and to support system design activities. That work is based upon small-scale batch melts (“crucible melts”) using waste simulants. Glass compositions suitable for processing the tank waste streams for the WTP have been identified and tested for processing and product quality requirements at various scales, ranging from crucible melts of a few grams, up to the LAW Pilot Melter at processing rates in excess of 6600 kg/day (2000 kg/m 2 /day). Recent testing performed at VSL for Washington River Protection Solutions, LLC (WRPS) has investigated the baseline “single-pass” retention of technetium over the range of WTP LAW feed compositions and glass formulations as well as the development of potential modifications to the baseline that could increase the single-pass retention. For the WTP melter systems, the liquid waste streams from the primary off-gas system components are recycled back to the pretreatment facility and, ultimately, to subsequent melter feed batches. Such a recycle loop can greatly increase the overall retention of volatile constituents in the glass product over the single-pass value. Accordingly, more recent testing at VSL has been conducted on a melter system that incorporates a recycle flow sheet. The results from that work have provided important performance data for the unit operations in the primary off-gas system with respect to technetium capture and recycle, which is crucial in assessing the amount of technetium that ultimately reports to non-glass waste forms.

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DOE OSTI2023

Conceptual Design of a Salt Dechlorination and Vitrification Apparatus

The dechlorination and vitrification apparatus (DeVA) discussed in this report is a prototype conceptualization of a device that could be used to remove halogens from salt-based waste streams so that the remaining salt cations can be converted to oxides and immobilized in a glass-based waste form.

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DOE OSTI2013

Projected Salt Waste Production from a Commercial Pyroprocessing Facility

Pyroprocessing of used nuclear fuel inevitably produces salt waste from electrorefining and/or oxide reduction unit operations. Various process design characteristics can affect the actual mass of such waste produced. This paper examines both oxide and metal fuel treatment, estimates the amount of salt waste generated, and assesses potential benefit of process options to mitigate the generation of salt waste. For reference purposes, a facility is considered in which 100 MT/year of fuel is processed. Salt waste estimates range from 8 to 20 MT/year from considering numerous scenarios. It appears that some benefit may be derived from advanced processes for separating fission products from molten salt waste, but the degree of improvement is limited. Waste form production is also considered but appears to be economically unfavorable. Direct disposal of salt into a salt basin type repository is found to be the most promising with respect to minimizing the impact of waste generation on the economic feasibility and sustainability of pyroprocessing.

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DOE OSTI2015

Characterization of Ceramic Material Produced From a Cold Crucible Induction Melter Test

This report summarizes the results from characterization of samples from a melt processed surrogate ceramic waste form. Completed in October of 2014, the first scaled proof of principle cold crucible induction melter (CCIM) test was conducted to process a Fe-hollandite-rich titanate ceramic for treatment of high level nuclear waste. X-ray diffraction, electron microscopy, inductively coupled plasma-atomic emission spectroscopy (and inductively coupled plasma-mass spectroscopy for Cs), and product consistency tests were used to characterize the CCIM material produced. Core samples at various radial locations from the center of the CCIM were taken. These samples were also sectioned and analyzed vertically. Together, the various samples were intended to provide an indication of the homogeneity throughout the CCIM with respect to phase assemblage, chemical composition, and chemical durability. Characterization analyses confirmed that a crystalline ceramic with desirable phase assemblage was produced from a melt using a CCIM. Hollandite and zirconolite were identified in addition to possible highly-substituted pyrochlore and perovskite. Minor phases rich in Fe, Al, or Cs were also identified. Remarkably only minor differences were observed vertically or radially in the CCIM material with respect to chemical composition, phase assemblage, and durability. This recent CCIM test and the resulting characterization in conjunction with demonstrated compositional improvements support continuation of CCIM testing with an improved feed composition and improved melter system.

36 MATERIALS SCIENCE↗
DOE OSTI2021

An Evaluation of the Impact of Glycolate and Glycolate Mitigation on the Defense Waste Processing Facility Recycle Diversion Project Flowsheet

The Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) processes radioactive High Level Waste (HLW) sludge solids from the Concentration, Storage, and Transfer Facilities (CSTF); which includes the SRS Tank Farm and Evaporator facilities, and a concentrated Cs-137 laden stream and a Monosodium Titanate (MST) and sludge solids stream from the SRS Salt Waste Processing Facility (SWPF). The waste is chemically adjusted with acids and reductant (currently with 50 wt.% nitric acid and ~90 wt.% formic acid, but eventually formic acid will be substituted with ~70 wt.% glycolic acid), and frit is added so that a durable, borosilicate glass waste form can be produced when the material is vitrified in the melter. As a result of the evaporation of water during both the melter feed preparation and the melter feed vitrification steps in DWPF, a recycle waste stream is generated, neutralized, and sent back to the CSTF. The recycle waste is a dilute aqueous stream originating from the collection of condensate liquids containing some minor sludge, MST, and frit solids and other waste components resulting from melter feed entrainment during foamover events and transfer of volatile species into the condensate. The recycle stream volume is significant and is expected to approach 3 million gallons per year once SWPF reaches full operation, requiring the use of multiple large CSTF tanks for storage. The recycle waste is currently collected in the SRS Tank Farm and periodically evaporated in the 242-16H (2H) Evaporator to conserve storage space.

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DOE OSTI2021

Annual Status Report (FY 2020): Performance Assessment for the Disposal of Low Level Waste in the 200 East Area Burial Grounds

This annual review provides the projected dose estimates of radionuclide inventories disposed in the 200 East Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis (WHC-SD-WM-TI-730). The estimates are compared with requirements of DOE O 435.1 Chg 1 and performance objectives defined in companion documents DOE M 435.1-1 Chg 1and DOE-STD-5002-2017). All performance objectives are currently satisfied, and operational waste acceptance criteria (HNF-EP-0063) and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years’ evaluations that indicate potential impacts well below performance objectives; therefore, future compliance with DOE O 435.1 Chg 1 is expected. A new PA study was initiated in fiscal year (FY) 2019 for evaluation of active disposal sites within the 200 East and 200 West Areas (Trench 94 in 200 East; Trenches 31 and 34 in 200 West) due to extended time elapsing between the current annual status report and the original PA for the active disposal sites. The new PA for the active disposal sites is expected to be completed in FY 2021. Within the active burial grounds in the 200 East Area, low-level waste and mixed low-level waste will continue to be disposed of in the dedicated U.S. Navy reactor compartment trench at the 218-E-12B Burial Ground (Trench 94). During this reporting period (FY 2020, from October 1, 2019, through September 30, 2020), two reactor compartments were disposed in Trench 94. Results from sorption experiments are summarized for this reporting period to quantify the efficacy of concrete waste forms in retaining key radionuclides (e.g., technetium-99 and iodine-129). The test durations ranged from 1 to 3 months. Continued groundwater monitoring of the 200 East Area LLBGs indicates no groundwater contamination due to LLBG waste. Current assumptions about future land use at the Hanford Site are consistent with PA analysis1 assumptions of a post-closure facility that will not be degraded by human activity. The LLBGs are located in an area identified for waste management and containment of residual contamination. This area will remain after final environmental remediation and the proposed shrinkage of Hanford Site boundaries to small sections within the 200 East and 200 West Areas in the Central Plateau (DOE/EIS-0391). The current closure plan for the LLBGs (DOE/RL-2000-707) estimates that the 200 East LLBGs will be closed in the 2050 timeframe. The Disposal Authorization Statement, other technical basis documents, and the radioactive waste management basis are of continued adequacy to meet the performance objectives of DOE O 435.1 Chg 1. Overall, there are no substantive changes to primary PA assumptions nor the PA analysis conclusion; therefore, compliance with DOE O 435.1 Chg 1 and the Disposal Authorization Statement is maintained.

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DOE OSTI2022

Results of Re-evaluation of FEPs Related to Implementing the ABD Glass Program

The Savannah River Site plans to reprocess defense spent nuclear fuel currently stored in their L-Basin via the Accelerated Basin Deinventory (ABD) Program. The previous plan for the L-Basin spent nuclear fuel was to dispose of it directly in the federal repository without reprocessing. Implementing the ABD Program will result in final disposal of approximately 900 fewer canisters of defense spent nuclear fuel and the production of approximately 521 more canisters of vitrified high-level waste glass with some specific differences from the planned high-level waste glass. Because the 235U in the L-Basin spent nuclear fuel is not intended to be recovered, the fissile mass loading of the vitrified high-level glass waste form to be produced must be increased above the current value of 897 g/m 3 to a maximum of 2,500 g/m 3 . Therefore, implementing the ABD Program would produce a variant of high-level waste glass—the ABD glass—that needs to be evaluated for future repository licensing, which includes both preclosure safety and postclosure performance. This report describes the approach to and summarizes the results of an evaluation of the potential effects of implementing the ABD Program at the Savannah River Site on the technical basis for future repository licensing for a generic repository that is similar to Yucca Mountain and for one that is fully generic. This evaluation includes the effects on preclosure safety analyses and postclosure performance assessment for both repository settings. The license application for the proposed Yucca Mountain repository (DOE 2008), which is serving as a framework for this evaluation, concluded that the proposed Yucca Mountain repository would meet all applicable regulatory requirements. The evaluation documented in this report found that implementing the ABD Program is not expected to change that conclusion for a generic repository similar to Yucca Mountain or for a generic repository with respect to the preclosure safety analyses. With respect to the postclosure performance of a generic repository, no concerns were identified.

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DOE OSTI2017

Tracking the Key Constituents of Concern of the WTP LAW Stream

The testing results presented in the present report were also obtained on a DM10 melter system operated with the primary WTP LAW offgas system components with recycle, as specified in the statement of work (SOW) [6] and detailed in the Test Plan for this work [7]. The primary offgas system components include the SBS, the WESP, and a recycle system that allows recycle of liquid effluents back to the melter, as in the present baseline for the WTP LAW vitrification. The partitioning of technetium and other key constituents between the glass waste form, the offgas system liquid effluents, the offgas stream that exits the WESP, and the liquid condensate from the vacuum evaporator were quantified in this work. The tests employed three different LAW streams spanning a range of waste compositions anticipated for WTP. Modifications to the offgas system and operational strategy were made to expedite the approach to steady state concentrations of key constituents in the glass and offgas effluent solutions during each test.

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DOE OSTI2023

Increasing the Fissile Mass Loading of High-Level Waste Glass Canisters to Greater Than 2,500 g/m 3 at the Savannah River Site

To eliminate future fissile mass loading constraints for the Savannah River Site H-Canyon Facility and Liquid Waste system, the Savannah River National Laboratory recommends a repository evaluation of a uranium fissile mass loading in glass at 7,144 g/m 3 in addition to the existing International Atomic Energy Agency safeguards and security limit of 2,500 g/m 3 total plutonium. This recommended increase above the authorized 2,500 g/m 3 fissile mass loading limit is based on concentrations of uranium and plutonium shown to produce an acceptable glass waste form rather than projections of the maximum fissile mass loading in future sludge batches. An authorized fissile mass loading limit greater than 2,500 g/m 3 will increase facility flexibility, reduce the number of high-level waste canisters produced, and potentially avoid an increase in the Department of Energy Environmental Management mission life without compromising safety or glass product quality.

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DOE OSTI2024

Viscosity Measurements of the HLW APPS Glasses

The U.S. Department of Energy is responsible for building the Hanford Waste Treatment and Immobilization Plant (WTP) at the Hanford site in Washington to remediate 56 million gallons of radioactive waste historically stored in 177 underground tanks. The Office of River Protection has requested that the Savannah River National Laboratory (SRNL) contribute in areas of recognized capabilities and expertise for glass waste form development to support successful startup of the WTP. SRNL support of this work is defined in the Task Technical and Quality Assurance Plan.1 This report provides results from viscosity measurements on a series of simulated nuclear waste glasses designed and fabricated at Pacific Northwest National Laboratory (PNNL). The glasses were designated the high-level waste – Aspen Process Performance Simulation (HLW-APPS) study glasses. The data provided in this report is to be used in the development, validation, and implementation of enhanced property/composition models for nuclear waste glasses.

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DOE OSTI2015

Develop Plan for Analysis of the Effluent from GCM Production

This milestone is focused on developing a plan for the analysis of the effluent from the Sandia low temperature sintering Bi-Si-Zn oxide glass composite material (GCM) waste form for the long term storage of iodine and its capture materials.

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DOE OSTI2025

Mercury sequestration in alkaline salt low-level radioactive waste

Liquid low-level radioactive waste at the Savannah River Site contains several species of mercury, including inorganic, elemental and methylmercury. This waste is solidified and stabilized in a cementitious waste form referred to as saltstone. Soluble mercury is stabilized as β-cinnabar, HgS as the result of reaction between the mercury and sulfur present in blast furnace slag, one of the cementitious regents. In this investigation, Mersorb®, a commercial granular activated carbon impregnated with sulfur, was evaluated as a pretreatment to remove mercury from the solution prior to cementation. Mersorb® was found to remove more than 96 mass percent of the methylmercury in simulated tank waste solution when the mass ratio of Mersorb® to mercury was above 2.5. Slag sequestered relatively more inorganic mercury than organic mercury in simulated tank waste after 24 hours of contact. This is likely due to the mercury-carbon bond being more covalent than the mercury-oxygen bond and therefore more difficult to break and slower to form HgS.

Cementitious Material↗
DOE OSTI2015

Immobilized Low-Activity Waste Glass Release Data Package for the Integrated Disposal Facility Performance Assessment

The Hanford Site in southeastern Washington State has been used extensively by the U.S. Department of Energy (DOE) to produce nuclear materials for the U.S. defense arsenal. A large inventory of radioactive and mixed waste has accumulated in 177 buried single- and double-shell tanks. Liquid waste recovered from the tanks will be pretreated to separate the low-activity fraction from the high-level and transuranic wastes. The low-activity waste (LAW) is planned to be immobilized in glass, via vitrification, and placed into a near-surface disposal system on the Hanford Site. Before the immobilized low-activity waste (ILAW) can be placed into the disposal system, DOE must approve a performance assessment (PA), which documents the assumptions and provides the quantitative demonstration of compliance with the performance objectives for the long-term protection of the public and the environment. A critical component of the PA will be to demonstrate that releases from the selected glass waste form do not result in the performance objectives being exceeded. The PA must also, for purposes of establishing limits on radionuclides that may be buried near-surface, assess impacts to a hypothetical intruder and impacts to water resources. The key information presented in this data package consists of the specific glass compositions and associated glass dissolution rate law parameters for five ILAW glasses. In addition, the results of geochemical modeling performed on several glasses that are not specifically described in this document are provided as supplemental data to help identify potential end-point alteration phases that may form during the weathering of ILAW glasses. These data (glass compositions, rate law parameters, and potential alteration phases) will be used to identify contaminant release source terms for the Integrated Disposal Facility PA. In addition, a general discussion on nuclear waste glass corrosion and a brief overview of modeling approaches and corresponding uncertainties are discussed. Table ES.1 through Table ES.4 provide the specific data to be used to support the 2017 PA.

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DOE OSTI2019

Dechlorination of molten chloride waste salt from electrorefining via ion-exchange using pelletized ultra-stable H-Y zeolite in a fluidized particle reactor

Dechlorination of eutectic LiCl–KCl based electrorefner (ER) salt is reported via ion-exchange reaction with protonated ultrastable Y-type (USHY) zeolite bound into mechanically fuidized 45–250 μm diameter particles. Evidence of exchange of cations from the salt (Li + , K + , and fission product cations) into the zeolite lattice replacing H + ions was found based on a change in unit cell size, ICP-MS, XRD and TEM–EDS in addition to detection of HCl of gas. Ion exchange reaction was carried out at 625 and 650 °C, temperatures above the melting point of eutectic LiCl–KCl. Experiments were carried out to optimize zeolite drying temperature, estimate maximum ion-exchange capacity, and determine the thermal stability of USHY zeolite. The results indicate over 90% dechlorination can be achieved without zeolite structure collapse at 625 °C. Finally, this provides a promising route to stabilizing waste from radioactive chloride salts into dechlorinated waste forms for permanent geologic disposal.

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