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DOE OSTI2011

Waste Acceptance Testing of Secondary Waste Forms: Cast Stone, Ceramicrete and DuraLith

To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions has initiated secondary-waste-form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is conducting tests on four candidate waste forms to evaluate their ability to meet potential waste acceptance criteria for immobilized secondary wastes that would be placed in the IDF. All three waste forms demonstrated compressive strengths above the minimum 3.45 MPa (500 psi) set as a target for cement-based waste forms. Further, none of the waste forms showed any significant degradation in compressive strength after undergoing thermal cycling (30 cycles in a 10 day period) between -40 C and 60 C or water immersion for 90 days. The three leach test methods are intended to measure the diffusion rates of contaminants from the waste forms. Results are reported in terms of diffusion coefficients and a leachability index (LI) calculated based on the diffusion coefficients. A smaller diffusion coefficient and a larger LI are desired. The NRC, in its Waste Form Technical Position (NRC 1991), provides recommendations and guidance regarding methods to demonstrate waste stability for land disposal of radioactive waste. Included is a recommendation to conduct leach tests using the ANS 16.1 method. The resulting leachability index (LI) should be greater than 6.0. For Hanford secondary wastes, the LI > 6.0 criterion applies to sodium leached from the waste form. For technetium and iodine, higher targets of LI > 9 for Tc and LI > 11 for iodine have been set based on early waste-disposal risk and performance assessment analyses. The results of these three leach tests conducted for a total time between 11days (ASTM C1308) to 90 days (ANS 16.1) showed: (1) Technetium diffusivity: ANSI/ANS 16.1, ASTM C1308, and EPA 1315 tests indicated that all the waste forms had leachability indices better than the target LI > 9 for technetium; (2) Rhenium diffusivity: Cast Stone 2M specimens, when tested using EPA 1315 protocol, had leachability indices better than the target LI > 9 for technetium based on rhenium as a surrogate for technetium. All other waste forms tested by ANSI/ANS 16.1, ASTM C1308, and EPA 1315 test methods had leachability indices that were below the target LI > 9 for Tc based on rhenium release. These studies indicated that use of Re(VII) as a surrogate for 99Tc(VII) in low temperature secondary waste forms containing reductants will provide overestimated diffusivity values for 99Tc. Therefore, it is not appropriate to use Re as a surrogate 99Tc in future low temperature waste form studies. (3) Iodine diffusivity: ANSI/ANS 16.1, ASTM C1308, and EPA 1315 tests indicated that the three waste forms had leachability indices that were below the target LI > 11 for iodine. Therefore, it may be necessary to use a more effective sequestering material than silver zeolite used in two of the waste forms (Ceramicrete and DuraLith); (4) Sodium diffusivity: All the waste form specimens tested by the three leach methods (ANSI/ANS 16.1, ASTM C1308, and EPA 1315) exceeded the target LI value of 6; (5) All three leach methods (ANS 16.1, ASTM C1308 and EPA 1315) provided similar 99Tc diffusivity values for both short-time transient diffusivity effects as well as long-term ({approx}90 days) steady diffusivity from each of the three tested waste forms (Cast Stone 2M, Ceramicrete and DuraLith). Therefore, any one of the three methods can be used to determine the contaminant diffusivities from a selected waste form.

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DOE OSTI2015

The evolution of the Nevada National Security site waste acceptance criteria - 15312

The Nevada National Security Site Waste Acceptance Criteria (DOE/NV--325) is an evolution of requirements spanning almost three decades of waste acceptance at the Nevada National Security Site (NNSS). The original waste acceptance criteria (WAC) document was written in 1978 (document NVO-185) and underwent multiple revisions as time passed and disposal experience was gained. This document came into being when the NNSS first started receiving significant quantities of low-level radioactive waste (LLW) from offsite U.S. Department of Energy (DOE) generators, such as the Mound and Rocky Flats Plants. The original WAC was far less prescriptive and rigorous than the current criteria since many of the current DOE order directives, Federal/State laws, and regulations did not exist. The current WAC reflects the results of numerous revisions in order to address the nuances of the varied waste streams and package configurations accepted for disposal at the NNSS. This includes waste streams that exhibit challenges not previously faced or covered by the WAC and therefore, from time to time, require new information and or changes to be made to the WAC. These wastes now represent a growing percentage of the remaining waste at cleanup sites across the DOE Complex. These waste streams present challenges, such as the technical basis for ensuring safe transportation to and disposal at the NNSS. The current WAC (Rev. 10 dated June 2013), is Available online at: http://www.nv.energy.gov/library/publications/Environmental/DOENV{sub 3}25.pdf). Other challenges for the NNSS include incorporating new and emerging technologies and addressing stakeholder concerns. Addressing these challenges sometimes results in changes to the NNSS WAC. Drivers for revising the NNSS WAC over the years have included operating experience improvements, the need to specify waste origin/ownership, and identifying preferred transportation routing. This paper will discuss various aspects of the NNSS WAC, and how it has evolved based upon DOE orders and requirements, acceptance of offsite mixed wastes, acceptance of non-radioactive classified wastes, stakeholder concerns and agreements, and the changing face of the remaining legacy waste at DOE sites across the United States. (authors)

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DOE OSTI2018

Enhanced Evaluation Criteria to Meet the WIPP Waste Acceptance Criteria - 18655

The Waste Isolation Pilot Plant (WIPP) is a deep geologic repository for the disposal of Transuranic (TRU) waste and the only such disposal facility of its kind. TRU waste is radioactive mixed waste generated by the U. S. Department of Energy in support of National Defense. In 2014, the WIPP facility was shut down when two separate accidents occurred in the repository. The two accidents were a fire on a large underground haul truck that prompted evacuation and filled the mine with smoke and soot, and a radiation release that contaminated the exhaust ventilation and disposal areas of the mine. DOE performed two accident investigations in 2014 that resulted in many corrective actions. Some of the corrective actions resulted in significant changes in the WIPP Waste Acceptance Criteria (WAC) and are required to be applied to all TRU waste upon the resumption of shipments to WIPP. WIPP re-opened and began taking shipments again in 2017 and a large backlog of over 25,000 TRU waste containers that were certified prior to the new WAC have been undergoing re-evaluation since the re-opening. This paper will focus on the new WAC criteria that TRU waste must meet above-and-beyond previous WAC criteria and the implementation of new processes for approval. The primary area for these changes is in the area of waste characterization using Acceptable Knowledge (AK) and new detailed analyses required for chemical compatibility. The accident investigation revealed that the radiation release was due to a thermal runaway and bursting of the waste container due to inadvertent mixtures of nitrates and combustibles resulting in ignitable waste forms. In other words, the TRU waste container had incompatible contents that went undiscovered and created a hazard characteristic that was prohibited at WIPP. Through detailed chemical compatibility evaluations and new oversight, the DOE should be able to prevent a similar accident (see Figure 1). The new chemical compatibility evaluations and evaluations of oxidizing chemicals combined with combustibles are complex and detailed. Chemical compatibility is not a new requirement, but a much more detailed evaluation is a new requirement. And, obtaining historical information about the generation of the final waste form for legacy packaged waste can be very challenging. Specifically, determining possible amounts of chemicals in the waste can be extremely challenging. This paper does not explore in depth, the details of oxidizing chemicals. Other papers presented will discuss oxidizing chemicals. (author)

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DOE OSTI2012

Nevada National Security Site Waste Acceptance Criteria

This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO), Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept DOE non-radioactive classified waste, DOE non-radioactive hazardous classified waste, DOE low-level radioactive waste (LLW), DOE mixed low-level waste (MLLW), and U.S. Department of Defense (DOD) classified waste for permanent disposal. Classified waste is the only waste accepted for disposal that may be non-radioactive and will be required to meet the waste acceptance criteria for radioactive waste as specified in this document. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project (WMP) at (702) 295-7063, and your call will be directed to the appropriate contact.

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DOE OSTI2013

Nevada National Security Site Waste Acceptance Criteria

This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO), Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept the following: • DOE hazardous and non-hazardous non-radioactive classified waste • DOE low-level radioactive waste (LLW) • DOE mixed low-level waste (MLLW) • U.S. Department of Defense (DOD) classified waste The LLW and MLLW listed above may also be classified waste. Classified waste is the only waste accepted for disposal that may be non-radioactive and shall be required to meet the waste acceptance criteria for radioactive waste as specified in this document. Classified waste may be sent to the NNSS as classified matter. Section 3.1.18 provides the requirements that must be met for permanent burial of classified matter. The NNSA/NFO and support contractors are available to assist the generator in understanding or interpreting this document. For assistance, please call the NNSA/NFO Environmental Management Operations (EMO) at (702) 295-7063, and the call will be directed to the appropriate contact.

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DOE OSTI2019

Development of the Waste Acceptance Criteria for the Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Facility - 19143

The Idaho National Laboratory (INL) Remote-Handled Low-Level Waste (RH LLW) Disposal Facility has been constructed, has received operational approval from the Department of Energy (DOE), and is expected to receive its first waste shipment in FY 2019. The RH LLW Disposal Facility will support the needs of INL and the Naval Reactors Facility (NRF). RH LLW destined for the facility includes waste currently in storage from previous programs and projects conducted at INL; newly generated waste from programs and projects conducted at INL; and waste from NRF spent fuel processing activities. This new facility supports RH LLW disposal needs for the ongoing mission of INL and NRF. Waste to be disposed of at the RH LLW Disposal Facility may exhibit dose rates that range from 0.002 Sv/hr to 600 Sv/hr. Waste types include resins and activated metals as well as surface-contaminated debris generated from existing and new missions. The total number of canisters currently planned for disposal at the RH LLW Disposal Facility is 939 with the possibility of future expansion. The projected waste volume to be disposed of is 3,100 m{sup 3}. The RH LLW Disposal Facility will receive a wide variety of RH LLW and is in preparations to receive its first canisters in FY 2019. Waste acceptance criteria (WAC) and the radioactive waste management basis have been developed and implemented in accordance with the applicable DOE Order. In order to ensure that DOE requirements and guidance were appropriately implemented in the RH LLW Disposal Facility WAC, the project worked closely with the DOE Low Level Waste Disposal Facility Federal Review Group (LFRG) representatives and followed the applicable DOE standard. This paper provides an overview of the development of the WAC, including waste and waste generator certification requirements, for the RH LLW Disposal Facility. The paper illustrates the importance of maintaining strong ties to the development and revision of the RH LLW Disposal Facility Performance Assessment (PA), Documented Safety Analysis, Facility Procedures, Readiness Reviews, DOE Order, and applicable Technical Standards. The paper also provides an overview of the RH LLW Disposal Facility, WAC generator certification requirements, distinctive challenges to acceptance of legacy waste, and the unique approach the RH LLW Disposal Facility has taken to ensure compliance with PA driven limits. (authors)

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DOE OSTI2016

Nevada National Security Site Waste Acceptance Criteria (NNSS/WAC)

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DOE OSTI2013

Multi-discipline Waste Acceptance Process at the Nevada National Security Site - 13573

The Nevada National Security Site low-level radioactive waste disposal facility acceptance process requires multiple disciplines to ensure the protection of workers, the public, and the environment. These disciplines, which include waste acceptance, nuclear criticality, safety, permitting, operations, and performance assessment, combine into the overall waste acceptance process to assess low-level radioactive waste streams for disposal at the Area 5 Radioactive Waste Management Site. Four waste streams recently highlighted the integration of these disciplines: the Oak Ridge Radioisotope Thermoelectric Generators and Consolidated Edison Uranium Solidification Project material, West Valley Melter, and classified waste. (authors)

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DOE OSTI2011

Nevada National Security Site Waste Acceptance Criteria

This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NNSSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NNSS Area 3 and Area 5 Radioactive Waste Management Complex for disposal. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project at (702) 295-7063 or fax to (702) 295-1153.

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DOE OSTI2010

Nevada National Security Site Waste Acceptance Criteria

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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.

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DOE OSTI2021

Methylmercury Retention Evaluation to Support Tank 50 and Saltstone Waste Acceptance Criteria: Parts 2 and 3

The objective of this work was to provide a technical basis for the organic mercury waste acceptance criterion (WAC) for Tank 50 salt solution. The current mercury WAC for Tank 50 solution is 325 mg/L as Hg regardless of the form of mercury, which is about three times higher than any total Hg concentrations measured to date in Tank 50 solution. To date, the highest total Hg and MeHg + concentrations in the Tank 50 solution were 105 and 62 mg/L, respectively.

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DOE OSTI2014

Waste Acceptance and Disposal at the Nevada National Security Site

The U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO) relies on the Performance Assessment (PA) and the PA model of the Area 5 Radioactive Waste Management Site (RWMS) at the Nevada National Security Site (NNSS) to make disposal decisions on waste stream profiles submitted by waste generators. Disposal decisions are made following a formal process called the PA review process. This paper presents the PA review process, which is a part of the NNSA/NFO’s Waste Generator Documentation Approval Process. The review process includes three major steps and employs quantitative tools such as the PA-derived factors and probabilistic inventory and PA models developed in GoldSim. The PA-derived radionuclide action levels for waste characterization and reporting are included in the NNSS Waste Acceptance Criteria. Each year, NNSA/NFO reviews approximately 100 to 150 new or revised waste stream profiles. The PA review group ensures that, with the disposal of approved waste streams, the Area 5 RWMS will remain in compliance with the performance objectives of DOE O 435.1, “Radioactive Waste Management.” Examples of the PA review process are given that demonstrate how waste streams with special characteristics are evaluated and disposal decisions are made. The most rigorous profile evaluation is done with the PA model, run with the proposed profile inventory added and results compared to the PA results. Additional analysis and process-modeling are performed for those waste streams with unique features not addressed with the PA model. The results of these analyses and modeling are documented in special analysis reports, reviewed and approved by NNSA/NFO. The examples include (1) profiles with radionuclides not in the PA, (2) profiles including radionuclides with high radon-generation potential, and (3) profiles with high-heat generating radionuclides.

Yucel, Vefa↗

DOE OSTI2021

Tank 50 Simulant Grout and Toxicity Characteristic Leaching Procedure (TCLP) Results for Methyl Mercury Waste Acceptance Criteria (WAC) Limit

Testing of simulated Tank 50 salt solution grout, spiked with methyl mercury (MeHg) and ionic mercury, used the Toxicity Characteristic Leaching Procedure (TCLP) to evaluate leaching performance with respect to Hg and the Waste Acceptance Criteria (WAC) limits for the Saltstone Production Facility (SPF). This report describes the test parameters for simulant preparation and confirmatory analysis, grout preparation, TCLP leach testing and subsequent leachate analyses. These scoping tests were designed to provide an initial evaluation of the effect of increased MeHg and ionic Hg on the TCLP and were requested by Savannah River Remediation LLC (SRR) through a Technical Assistance Request (TAR). It is expected that this preliminary testing will likely lead to future, more detailed, testing involving the mercury (Hg) WAC and Hg limits with respect to the Tank 50 TCLP for mercury. The TCLP concentrations for total Hg have approached 0.06 mg/L in recent years for Tank 50 radioactive solutions containing up to 105 mg/L of Hg in the form of ionic (Hg(I/II), elemental (Hg(0)) and organo-mercury such as MeHg. In discussions with SRR personnel, it was decided for this non-baseline work to target Hg spike concentrations, using predominately MeHg, in the range of 150 to 350 mg/L total Hg to investigate corresponding Hg TCLP leaching behaviors.

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DOE OSTI2019

Understanding Nitrate Chemistry - The Key to Safer and More Efficient TRU Waste Acceptance - 19618

Investigations into the Waste Isolation Pilot Plant (WIPP) radiological release from a transuranic (TRU) waste drum concluded that the event was initiated by a thermal runaway, principally involving interactions between the polysaccharide-based kitty litter and nitric acid, which released radiological material along with hot gases and combustible materials from the drum. In this paper, the hazards of polysaccharides present as either starch based kitty litter or cellulose rags in a TRU waste drum exposed to nitric acid (and by extension metal nitrate salts) have been evaluated, based on considerations of how complex nitrogen redox cycles carry out the aqueous phase oxidation of alcohols and related organic functional groups. These exothermic reactions, coupled with a positive feedback loop inherent to nitrogen redox chemistry, can initiate an autocatalytic thermal runaway and explosive gas generation, which poses the most significant risk under ambient conditions where acidic solution may be present (in small quantities as free liquid or sorbed within a solid matrix). Nitric acid is an extremely potent oxidant but is itself remarkably inert, rarely interacting directly with oxidizable substrates and instead exerting its vast potential energy through lower oxidation state reactants such as nitrogen dioxide and nitrous acid that are generated from its slow self-decomposition. Nitrous acid is the reactant that carries out substrate oxidation, and systems that are not pre-loaded with nitrous acid will experience an induction period before significant reactivity commences. Under oxidizing fuel-lean conditions, nitrogen dioxide may induce autocatalytic nitrous acid formation, which may supplement additional side-reactions, potentially leading to thermal runaway. Therefore, the system is most vulnerable at the end of an induction period. Metal nitrate salts promote this ambient temperature solution chemistry (cf. traditional oxidizer properties of metal nitrate salts, which operate in the solid state at elevated temperatures) by releasing nitric acid into the system via hydrolysis. Unlike simple alcohols with a single hydroxyl functional group, the chemistry of polymeric carbohydrates (polysaccharides) such as starches and cellulose is more diverse. Therefore, the oxidation of the simple monomeric carbohydrate glucose is first examined to provide insight into the analogous reactivity with polysaccharides. Although less reactive than alcohols by virtue of extensive hydrogen bonding, carbohydrates are oxidized by nitric acid through the same fundamental mechanisms, complicated by a variety of competing side reactions including acid-catalyzed hydrolytic depolymerization to more reactive (ultimately monomeric) units, as well as other degradation reactions promoted by radical mechanisms. For this reason, polysaccharides such as starches that are more susceptible to hydrolysis are more reactive than cellulose, with the attendant microbial population in starch based kitty litter adding a special aspect of reactivity. From this safety evaluation, a bounding latency period has been developed beyond which a stored TRU waste drum that contains polysaccharides (starch based kitty litter excluded) commingled with nitric acid and/or metal nitrate salts may no longer present a hazard under ambient conditions, based on a compilation of induction periods and reaction times involving nitric acid with various oxidizable substrates under a variety of conditions from synthetic procedures, tests, and adverse incident scenarios. By obviating the need for costly treatment options or characterization, a better understanding of nitrate redox chemistry and the hazards it presents with polysaccharide oxidation thus provides a more efficient means to safely dispose of TRU waste generated within the Department of Energy (DOE) complex. (author)

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DOE OSTI2023

Chemical Characterization Results for Tank 40 Sludge Batch 10 Waste Acceptance Product Specifications (WAPS) Sample

A 3-L sample (HTF-40-23-24) obtained from Tank 40 following transfer of Tank 51 to achieve the final Sludge Batch 10 processing composition was received by Savannah River National Laboratory (SRNL). SRNL was tasked with characterization of the sample in accordance with requirements for reporting the Waste Acceptance Product Specifications. A representative subsample, approximately 500 mL, was prepared by continuously agitating the 3-L sample via a mixing blade inserted into the bottle. This subsample was then used for the chemical and physical characterization reported here. This characterization includes the following: (1) Supernate and slurry density; (2) Weight percent solids; (3) Mercury Analysis (supernate and slurry); (4) Anions; (5) Total Organic/Inorganic Carbon; (6) Free Hydroxide; (7) Other Base; (8) Elementals; (9) Volatile Organic Analysis and Semi-volatile Organic Analysis; and (10) Fissile Radionuclides and Uranium Enrichment. The results of this physical and chemical characterization of the sample are documented in this report.

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DOE OSTI2019

Methylmercury Speciation and Retention Evaluation to Support Saltstone Waste Acceptance Criteria

Mercury is a contaminant of concern in Savannah River Site (SRS) high level waste (HLW). For five decades elemental mercury was used in the catalytic dissolution of aluminum cladding from the enrichedalloy uranium recovery process. Over 60,000 kg of mercury are now distributed in the SRS HLW tanks. The mercury speciation in the liquid waste streams resulting from conditioning HLW sludge for vitrification and in the decontaminated low activity sodium salt solution removed from the tanks was initially thought to be primarily ionic with a minor amount of elemental mercury. Recent sampling detected organic mercury, primarily methylmercury, HgCH 3 + (MeHg + ), and a trace amount of ethylmercury (EtHg) in these streams in addition to the ionic and elemental forms.

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DOE OSTI2010

DWPF COAL-CARBON WASTE ACCEPTANCE CRITERIA LIMIT EVALUATION BASED ON EXPERIMENTAL WORK (TANK 48 IMPACT STUDY)

This report summarizes the results of both experimental and modeling studies performed using Sludge Batch 10 (SB10) simulants and FBSR product from Tank 48 simulant testing in order to develop higher levels of coal-carbon that can be managed by DWPF. Once the Fluidized Bed Steam Reforming (FBSR) process starts up for treatment of Tank 48 legacy waste, the FBSR product stream will contribute higher levels of coal-carbon in the sludge batch for processing at DWPF. Coal-carbon is added into the FBSR process as a reductant and some of it will be present in the FBSR product as unreacted coal. The FBSR product will be slurried in water, transferred to Tank Farm and will be combined with sludge and washed to produce the sludge batch that DWPF will process. The FBSR product is high in both water soluble sodium carbonate and unreacted coal-carbon. Most of the sodium carbonate is removed during washing but all of the coal-carbon will remain and become part of the DWPF sludge batch. A paper study was performed earlier to assess the impact of FBSR coal-carbon on the DWPF Chemical Processing Cell (CPC) operation and melter off-gas flammability by combining it with SB10-SB13. The results of the paper study are documented in Ref. 7 and the key findings included that SB10 would be the most difficult batch to process with the FBSR coal present and up to 5,000 mg/kg of coal-carbon could be fed to the melter without exceeding the off-gas flammability safety basis limits. In the present study, a bench-scale demonstration of the DWPF CPC processing was performed using SB10 simulants spiked with varying amounts of coal, and the resulting seven CPC products were fed to the DWPF melter cold cap and off-gas dynamics models to determine the maximum coal that can be processed through the melter without exceeding the off-gas flammability safety basis limits. Based on the results of these experimental and modeling studies, the presence of coal-carbon in the sludge feed to DWPF is found to have both positive (+) and negative (-) impact as summarized below: (-) Coal-carbon is a melter reductant. If excess coal-carbon is present, the resulting melter feed may be too reducing, potentially shortening the melter life. During this study, the Reduction/Oxidation Potential (REDOX) of the melter could be controlled by varying the ratio of nitric and formic acid. (-) The addition of coal-carbon increases the amount of nitric acid added and decreases the amount of formic acid added to control melter REDOX. This means that the CPC with the FBSR product is much more oxidizing than current CPC processing. In this study, adequate formic acid was present in all experiments to reduce mercury and manganese, two of the main goals of CPC processing. (-) Coal-carbon will be oxidized to carbon dioxide or carbon monoxide in the melter. The addition of coal-carbon to the FBSR product will lead to approximately 55% higher offgas production from formate, nitrate and carbon due to the decomposition of the carbon at the maximum levels in this testing. Higher offgas production could lead to higher cold cap coverage or melter foaming which could decrease melt rate. No testing was performed to evaluate the impact of the higher melter offgas flow. (+) The hydrogen production is greatly reduced in testing with coal as less formic acid is added in CPC processing. In the high acid run without coal, the peak hydrogen generation was 15 times higher than in the high acid run with added coal-carbon. (+) Coal-carbon is a less problematic reducing agent than formic acid, since the content of both carbon and hydrogen are important in evaluating the flammability of the melter offgas. Processing with coal-carbon decreases the amount of formic acid added in the CPC, leading to a lower flammability risk in processing with coal-carbon compared to the current DWPF flowsheet. (+) The seven SB10 formulations which were tested during the bench-scale CPC demonstration were all determined to be within the off-gas flammability safety basis limits during the 9X/5X off-gas surge for normal bubbled melter operation. The concentration of coal-carbon in these baseline melter feeds varied widely from 0 to 17,863 ppm, depending on the acid addition strategy used and the extent to which the required reductant (formic acid) was replaced with coal-carbon. All baseline feeds were redox-adjusted and three of them contained TOC higher than the current theoretical TSR limit of 18,900 ppm. (-) Additional coal-carbon was then added to each baseline feed until the calculated off-gas flammability equaled the safety basis limit of 60% of the LFL at the peak of off-gas surge ('max-coal'). In doing so, however, no counterbalancing nitrate was added, thus simulating the scenario where slugs of coal enter the melter as a result of uneven distribution of coal in the slurry.

01 COAL, LIGNITE, AND PEAT↗