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Sample records for dwpf melter feed

  1. Maximum total organic carbon limit for DWPF melter feed

    International Nuclear Information System (INIS)

    Choi, A.S.

    1995-01-01

    DWPF recently decided to control the potential flammability of melter off-gas by limiting the total carbon content in the melter feed and maintaining adequate conditions for combustion in the melter plenum. With this new strategy, all the LFL analyzers and associated interlocks and alarms were removed from both the primary and backup melter off-gas systems. Subsequently, D. Iverson of DWPF- T ampersand E requested that SRTC determine the maximum allowable total organic carbon (TOC) content in the melter feed which can be implemented as part of the Process Requirements for melter feed preparation (PR-S04). The maximum TOC limit thus determined in this study was about 24,000 ppm on an aqueous slurry basis. At the TOC levels below this, the peak concentration of combustible components in the quenched off-gas will not exceed 60 percent of the LFL during off-gas surges of magnitudes up to three times nominal, provided that the melter plenum temperature and the air purge rate to the BUFC are monitored and controlled above 650 degrees C and 220 lb/hr, respectively. Appropriate interlocks should discontinue the feeding when one or both of these conditions are not met. Both the magnitude and duration of an off-gas surge have a major impact on the maximum TOC limit, since they directly affect the melter plenum temperature and combustion. Although the data obtained during recent DWPF melter startup tests showed that the peak magnitude of a surge can be greater than three times nominal, the observed duration was considerably shorter, on the order of several seconds. The long surge duration assumed in this study has a greater impact on the plenum temperature than the peak magnitude, thus making the maximum TOC estimate conservative. Two models were used to make the necessary calculations to determine the TOC limit

  2. Control of DWPF melter feed composition

    International Nuclear Information System (INIS)

    Brown, K.G.; Edwards, R.E.; Postles, R.L.; Randall, C.T.

    1989-01-01

    The Defense Waste Processing Facility will be used to immobilize Savannah River Site high-level waste into a stable borosilicate glass for disposal in a geologic repository. Proper control of the melter feed composition in this facility is essential to the production of glass which meets product durability constraints dictated by repository regulations and facility processing constraints dictated by melter design. A technique has been developed which utilizes glass property models to determine acceptable processing regions based on the multiple constraints imposed on the glass product and to display these regions graphically. This system along with the batch simulation of the process is being used to form the basis for the statistical process control system for the facility

  3. Defining And Characterizing Sample Representativeness For DWPF Melter Feed Samples

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  4. Defining And Characterizing Sample Representativeness For DWPF Melter Feed Samples

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-29

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

  5. Yield Stress Reduction of DWPF Melter Feed Slurries

    International Nuclear Information System (INIS)

    Stone, M.E.; Smith, M.E.

    2007-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site vitrifies High Level Waste for repository internment. The process consists of three major steps: waste pretreatment, vitrification, and canister decontamination/sealing. The HLW consists of insoluble metal hydroxides and soluble sodium salts. The pretreatment process acidifies the sludge with nitric and formic acids, adds the glass formers as glass frit, then concentrates the resulting slurry to approximately 50 weight percent (wt%) total solids. This slurry is fed to the joule-heated melter where the remaining water is evaporated followed by calcination of the solids and conversion to glass. The Savannah River National Laboratory (SRNL) is currently assisting DWPF efforts to increase throughput of the melter. As part of this effort, SRNL has investigated methods to increase the solids content of the melter feed to reduce the heat load required to complete the evaporation of water and allow more of the energy available to calcine and vitrify the waste. The process equipment in the facility is fixed and cannot process materials with high yield stresses, therefore increasing the solids content will require that the yield stress of the melter feed slurries be reduced. Changing the glass former added during pretreatment from an irregularly shaped glass frit to nearly spherical beads was evaluated. The evaluation required a systems approach which included evaluations of the effectiveness of beads in reducing the melter feed yield stress as well as evaluations of the processing impacts of changing the frit morphology. Processing impacts of beads include changing the settling rate of the glass former (which effects mixing and sampling of the melter feed slurry and the frit addition equipment) as well as impacts on the melt behavior due to decreased surface area of the beads versus frit. Beads were produced from the DWPF process frit by fire polishing. The frit was allowed to free fall through a flame

  6. Maximum total organic carbon limits at different DWPF melter feed maters (U)

    International Nuclear Information System (INIS)

    Choi, A.S.

    1996-01-01

    The document presents information on the maximum total organic carbon (TOC) limits that are allowable in the DWPF melter feed without forming a potentially flammable vapor in the off-gas system were determined at feed rates varying from 0.7 to 1.5 GPM. At the maximum TOC levels predicted, the peak concentration of combustible gases in the quenched off-gas will not exceed 60 percent of the lower flammable limit during a 3X off-gas surge, provided that the indicated melter vapor space temperature and the total air supply to the melter are maintained. All the necessary calculations for this study were made using the 4-stage cold cap model and the melter off-gas dynamics model. A high-degree of conservatism was included in the calculational bases and assumptions. As a result, the proposed correlations are believed to by conservative enough to be used for the melter off-gas flammability control purposes

  7. Control of DWPF [Defense Waste Processing Facility] melter feed composition

    International Nuclear Information System (INIS)

    Edwards, R.E. Jr.; Brown, K.G.; Postles, R.L.

    1990-01-01

    The Defense Waste Processing Facility will be used to immobilize Savannah River Site high-level waste into a stable borosilicate glass for disposal in a geologic repository. Proper control of the melter feed composition in this facility is essential to the production of glass which meets product durability constraints dictated by repository regulations and facility processing constraints dictated by melter design. A technique has been developed which utilizes glass property models to determine acceptable processing regions based on the multiple constraints imposed on the glass product and to display these regions graphically. This system along with the batch simulation of the process is being used to form the basis for the statistical process control system for the facility. 13 refs., 3 figs., 1 tab

  8. Recommendations for rheological testing and modelling of DWPF melter feed slurries

    International Nuclear Information System (INIS)

    Shadday, M.A. Jr.

    1994-08-01

    The melter feed in the DWPF process is a non-Newtonian slurry. In the melter feed system and the sampling system, this slurry is pumped at a wide range of flow rates through pipes of various diameters. Both laminar and turbulent flows are encountered. Good rheology models of the melter feed slurries are necessary for useful hydraulic models of the melter feed and sampling systems. A concentric cylinder viscometer is presently used to characterize the stress/strain rate behavior of the melter feed slurries, and provide the data for developing rheology models of the fluids. The slurries exhibit yield stresses, and they are therefore modelled as Bingham plastics. The ranges of strain rates covered by the viscometer tests fall far short of the entire laminar flow range, and therefore hydraulic modelling applications of the present rheology models frequently require considerable extrapolation beyond the range of the data base. Since the rheology models are empirical, this cannot be done with confidence in the validity of the results. Axial pressure drop versus flow rate measurements in a straight pipe can easily fill in the rest of the laminar flow range with stress/strain rate data. The two types of viscometer tests would be complementary, with the concentric cylinder viscometer providing accurate data at low strain rates, near the yield point if one exists, and pipe flow tests providing data at high strain rates up to and including the transition to turbulence. With data that covers the laminar flow range, useful rheological models can be developed. In the Bingham plastic model, linear behavior of the shear stress as a function of the strain rate is assumed once the yield stress is exceeded. Both shear thinning and shear thickening behavior have been observed in viscometer tests. Bingham plastic models cannot handle this non-linear behavior, but a slightly more complicated yield/power law model can

  9. NOBLE METAL CHEMISTRY AND HYDROGEN GENERATION DURING SIMULATED DWPF MELTER FEED PREPARATION

    Energy Technology Data Exchange (ETDEWEB)

    Koopman, D

    2008-06-25

    Simulations of the Defense Waste Processing Facility (DWPF) Chemical Processing Cell vessels were performed with the primary purpose of producing melter feeds for the beaded frit program plus obtaining samples of simulated slurries containing high concentrations of noble metals for off-site analytical studies for the hydrogen program. Eight pairs of 22-L simulations were performed of the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles. These sixteen simulations did not contain mercury. Six pairs were trimmed with a single noble metal (Ag, Pd, Rh, or Ru). One pair had all four noble metals, and one pair had no noble metals. One supporting 4-L simulation was completed with Ru and Hg. Several other 4-L supporting tests with mercury have not yet been performed. This report covers the calculations performed on SRNL analytical and process data related to the noble metals and hydrogen generation. It was originally envisioned as a supporting document for the off-site analytical studies. Significant new findings were made, and many previous hypotheses and findings were given additional support as summarized below. The timing of hydrogen generation events was reproduced very well within each of the eight pairs of runs, e.g. the onset of hydrogen, peak in hydrogen, etc. occurred at nearly identical times. Peak generation rates and total SRAT masses of CO{sub 2} and oxides of nitrogen were reproduced well. Comparable measures for hydrogen were reproduced with more variability, but still reasonably well. The extent of the reproducibility of the results validates the conclusions that were drawn from the data.

  10. Freeze and restart of the DWPF Scale Glass Melter

    International Nuclear Information System (INIS)

    Choi, A.S.

    1989-01-01

    After over two years of successful demonstration of many design and operating concepts of the DWPF Melter system, the last Scale Glass Melter campaign was initiated on 6/9/88 and consisted of two parts; (1) simulation of noble metal buildup and (2) freeze and subsequent restart of the melter under various scenarios. The objectives were to simulate a prolonged power loss to major heating elements and to examine the characteristics of transient melter operations during a startup with a limited supply of lid heat. Experimental results indicate that in case of a total power loss to the lower electrodes such as due to noble metal deposition, spinel crystals will begin to form in the SRL 165 composite waste glass pool in 24 hours. The total lid heater power required to initiate joule heating was the same as that during slurry-feeding. Results of a radiative heat transfer analysis in the plenum indicate that under the identical operating conditions, the startup capabilities of the SGM and the DWPF Melter are quite similar, despite a greater lid heater to melt surface area ratio in the DWPF Melter

  11. DWPF Glass Melter Technology Manual: Volume 1

    International Nuclear Information System (INIS)

    Iverson, D.C.

    1993-01-01

    This document details information about the design of a glass melter to be used at the Defense Waste Processing Facility located at the Savannah River Site. Topics include: melter overview, design basis, materials, vessel configuration, insulation, refractory configuration, electrical isolation, electrodes, riser and pour spout heater design, dome heaters, feed tubes, drain valves, differential pressure pouring, and melter test results. Information is conveyed using many diagrams and photographs

  12. Literature Review: Assessment of DWPF Melter and Melter Off-gas System Lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Reigel, M. [Savannah River Site (SRS), Aiken, SC (United States)

    2015-07-30

    Testing to date for the MOC for the Hanford Waste Treatment and Immobilization Plant (WTP) melters is being reviewed with the lessons learned from DWPF in mind and with consideration to the changes in the flowsheet/feed compositions that have occurred since the original testing was performed. This information will be presented in a separate technical report that identifies any potential gaps for WTP processing.

  13. Preliminary Analysis of Species Partitioning in the DWPF Melter

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kesterson, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Johnson, F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-15

    The work described in this report is preliminary in nature since its goal was to demonstrate the feasibility of estimating the off-gas entrainment rates from the Defense Waste Processing Facility (DWPF) melter based on a simple mass balance using measured feed and glass pour stream compositions and timeaveraged melter operating data over the duration of one canister-filling cycle. The only case considered in this study involved the SB6 pour stream sample taken while Canister #3472 was being filled over a 20-hour period on 12/20/2010, approximately three months after the bubblers were installed. The analytical results for that pour stream sample provided the necessary glass composition data for the mass balance calculations. To estimate the “matching” feed composition, which is not necessarily the same as that of the Melter Feed Tank (MFT) batch being fed at the time of pour stream sampling, a mixing model was developed involving three preceding MFT batches as well as the one being fed at that time based on the assumption of perfect mixing in the glass pool but with an induction period to account for the process delays involved in the calcination/fusion step in the cold cap and the melter turnover.

  14. Preliminary analysis of species partitioning in the DWPF melter. Sludge batch 7A

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Smith III, F. G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-01-01

    The work described in this report is preliminary in nature since its goal was to demonstrate the feasibility of estimating the off-gas carryover from the Defense Waste Processing Facility (DWPF) melter based on a simple mass balance using measured feed and glass pour stream (PS) compositions and time-averaged melter operating data over the duration of one canister-filling cycle. The DWPF has been in radioactive operation for over 20 years processing a wide range of high-level waste (HLW) feed compositions under varying conditions such as bubbled vs. non-bubbled and feeding vs. idling. So it is desirable to find out how the varying feed compositions and operating parameters would have impacted the off-gas entrainment. However, the DWPF melter is not equipped with off-gas sampling or monitoring capabilities, so it is not feasible to measure off-gas entrainment rates directly. The proposed method provides an indirect way of doing so.

  15. DWPF Glass Melter Technology Manual: Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    Iverson, D.C.

    1993-12-31

    This document details information about the design of a glass melter to be used at the Defense Waste Processing Facility located at the Savannah River Site. Topics discussed include: Information collected during testing, equipment, materials, design basis, feed tubes, and an evaluation of the performance of various components. Information is conveyed using many diagrams and photographs.

  16. DWPF Glass Melter Technology Manual: Volume 3

    International Nuclear Information System (INIS)

    Iverson, D.C.

    1993-01-01

    This document details information about the design of a glass melter to be used at the Defense Waste Processing Facility located at the Savannah River Site. Topics discussed include: Information collected during testing, equipment, materials, design basis, feed tubes, and an evaluation of the performance of various components. Information is conveyed using many diagrams and photographs

  17. The behavior and effects of the noble metals in the DWPF melter system

    International Nuclear Information System (INIS)

    Hutson, N.D.; Smith, M.E.

    1992-01-01

    Fission-product noble metals have caused severe operating problems in numerous worldwide waste vitrification facilities. These dense, highly conductive noble metals have tended to accumulate on the floor of joule-heated glass melters causing electrical distortions which have, in some occurrences, rendered the melter inoperable. A pilot scale vitrification research facility at the U.S. Department of Energy's Savannah River Laboratory has been operated for more than a year with simulated feed streams containing noble metals. In this paper the behavior of these noble metals in the melter system and final glass product and their effects on the scaled DWPF-type melter are discussed

  18. Maximum organic carbon limits at different melter feed rates (U)

    International Nuclear Information System (INIS)

    Choi, A.S.

    1995-01-01

    This report documents the results of a study to assess the impact of varying melter feed rates on the maximum total organic carbon (TOC) limits allowable in the DWPF melter feed. Topics discussed include: carbon content; feed rate; feed composition; melter vapor space temperature; combustion and dilution air; off-gas surges; earlier work on maximum TOC; overview of models; and the results of the work completed

  19. Integrated DWPF Melter System (IDMS) campaign report: The first two noble metals operations

    International Nuclear Information System (INIS)

    Hutson, N.D.; Zamecnik, J.R.; Smith, M.E.; Miller, D.H.; Ritter, J.A.

    1991-01-01

    The Integrated DWPF Melter System (IDMS) is designed and constructed to provide an engineering-scale representation of the DWPF melter and its associated feed preparation and off-gas systems. The facility is the first pilot-scale melter system capable of processing mercury, and flowsheet levels of halides and noble metals. In order to characterize the processing of noble metals (Pd, Rh, Ru, and Ag) on a large scale, the IDMS will be operated batchstyle for at least nine feed preparation cycles. The first two of these operations are complete. The major observation to date occurred during the second run when significant amounts of hydrogen were evolved during the feed preparation cycle. The runs were conducted between June 7, 1990 and March 8, 1991. This time period included nearly six months of ''fix-up'' time when forced air purges were installed on the SRAT MFT and other feed preparation vessels to allow continued noble metals experimentation

  20. Program plan: DWPF/HLWDP stirred Melter Program Plan

    International Nuclear Information System (INIS)

    Smith, M.E.

    1994-01-01

    Slurry Fed Melters (SFM) have been developed in the United States, Europe, and Japan for the conversion of high-level radioactive waste (HLW) to borosilicate glass for permanent disposal. The newest design, the stirred melter, combines the high production rates and high glass quality features of the Joule-heated melters with the low-cost, compact, simple maintenance features of the pot melters. However, further engineering design and demonstrations are needed to operate the stirred melter on a large scale. This document outlines the program which develops a full scale stirred melter for the DWPF (240 pph), and provides a basis which will allow further scale-up of the technology for use in the Hanford High Level Waste Disposal Program (HLWDP) for up to four times the reference capacity

  1. Durability of glasses from the Hg-doped Integrated DWPF Melter System (IDMS) campaign

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1992-01-01

    The Integrated DWPF Melter System (IDMS) for the vitrification of high-level radioactive wastes is designed and constructed to be a 1/9th scale prototype of the full scale Defense Waste Processing Facility (DWPF) melter. The IDMS facility is the first engineering scale melter system capable of processing mercury, and flowsheet levels of halides and noble metals. In order to determine the effects of mercury on the feed preparation process, the off-gas chemistry, glass melting behavior, and glass durability, a three-run mercury (Hg) campaign was conducted. The glasses produced during the Hg campaign were composed of Batch 1 sludge, simulated precipitate hydrolysis aqueous product (PHA) from the Precipitate Hydrolysis Experimental Facility (PHEF), and Frit 202. The glasses were produced using the DWPF process/product models for glass durability, viscosity, and liquidus. The durability model indicated that the glasses would all be more durable than the glass qualified in the DWPF Environmental Assessment (EA). The glass quality was verified by performing the Product Consistency Test (PCT) which was designed for glass durability testing in the DWPF

  2. DWPF Glass Melter Technology Manual: Volume 4

    International Nuclear Information System (INIS)

    Iverson, D.C.

    1993-01-01

    This document details information about the design of a glass melter to be used at the Defense Waste Processing Facility located at the Savannah River Plant. Information contained in this document consists solely of a machine drawing and parts list and purchase orders with specifications of equipment used in the development of the melter

  3. DWPF Glass Melter Technology Manual: Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    Iverson, D.C.

    1993-12-31

    This document details information about the design of a glass melter to be used at the Defense Waste Processing Facility located at the Savannah River Plant. Information contained in this document consists solely of a machine drawing and parts list and purchase orders with specifications of equipment used in the development of the melter.

  4. The Behavior and Effects of the Noble Metals in the DWPF Melter System

    International Nuclear Information System (INIS)

    Smith, M.E.; Bickford, D.F.

    1997-01-01

    Governments worldwide have committed to stabilization of high-level nuclear waste (HLW) by vitrification to a durable glass form for permanent disposal. All of these nuclear wastes contain the fission-product noble metals: ruthenium, rhodium, and palladium. SRS wastes also contain natural silver from iodine scrubbers. Closely associated with the noble metals are the fission products selenium and tellurium which are chemical analogs of sulfur and which combine with noble metals to influence their behavior and properties. Experience has shown that these melt insoluble metals and their compounds tend to settle to the floor of Joule-heated ceramic melters. In fact, almost all of the major research and production facilities have experienced some operational problem which can be associated with the presence of dense accumulations of these relatively conductive metals and/or their compounds. In most cases, these deposits have led to a loss of production capability, in some cases, to the point that melter operation could not continue. HLW nuclear waste vitrification facilities in the United States are the Department of Energy's Defense Waste Processing Facility (DWPF) at the Savannah River Site, the planned Hanford Waste Vitrification Plant (HWVP) at the Hanford Site and the operating West Valley Demonstration Project (WVDP) at West Valley, NY. The Integrated DWPF Melter System (IDMS) is a vitrification test facility at the Savannah River Technology Center (SRTC). It was designed and constructed to provide an engineering-scale representation of the DWPF melter and its associated feed preparation and off-gas treatment systems. An extensive noble metals testing program was begun in 1990. The objectives of this task were to explore the effects of the noble metals on the DWPF melter feed preparation and waste vitrification processes. This report focuses on the vitrification portion of the test program

  5. DWPF Melter No.2 Prototype Bus Bar Test Report

    International Nuclear Information System (INIS)

    Gordon, J.

    2003-01-01

    Characterization and performance testing of a prototype DWPF Melter No.2 Dome Heater Bus Bar are described. The prototype bus bar was designed to address the design features of the existing system which may have contributed to water leaks on Melter No.1. Performance testing of the prototype revealed significant improvement over the existing design in reduction of both bus bar and heater connection maximum temperature, while characterization revealed a few minor design and manufacturing flaws in the bar. The prototype is recommended as an improvement over the existing design. Recommendations are also made in the area of quality control to ensure that critical design requirements are met

  6. DWPF Melter Off-Gas Flammability Assessment for Sludge Batch 9

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A. S. [Savannah River Site (SRS), Aiken, SC (United States)

    2016-07-11

    The slurry feed to the Defense Waste Processing Facility (DWPF) melter contains several organic carbon species that decompose in the cold cap and produce flammable gases that could accumulate in the off-gas system and create potential flammability hazard. To mitigate such a hazard, DWPF has implemented a strategy to impose the Technical Safety Requirement (TSR) limits on all key operating variables affecting off-gas flammability and operate the melter within those limits using both hardwired/software interlocks and administrative controls. The operating variables that are currently being controlled include; (1) total organic carbon (TOC), (2) air purges for combustion and dilution, (3) melter vapor space temperature, and (4) feed rate. The safety basis limits for these operating variables are determined using two computer models, 4-stage cold cap and Melter Off-Gas (MOG) dynamics models, under the baseline upset scenario - a surge in off-gas flow due to the inherent cold cap instabilities in the slurry-fed melter.

  7. Literature review: Assessment of DWPF melter and melter off-gas system lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Reigel, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-30

    A glass melter for use in processing radioactive waste is a challenging environment for the materials of construction (MOC) resulting from a combination of high temperatures, chemical attack, and erosion/corrosion; therefore, highly engineered materials must be selected for this application. The focus of this report is to review the testing and evaluations used in the selection of the Defense Waste Processing Facility (DWPF), glass contact MOC specifically the Monofrax® K-3 refractory and Inconel® 690 alloy. The degradation or corrosion mechanisms of these materials during pilot scale testing and in-service operation were analyzed over a range of oxidizing and reducing flowsheets; however, DWPF has primarily processed a reducing flowsheet (i.e., Fe2+/ΣFe of 0.09 to 0.33) since the start of radioactive operations. This report also discusses the materials selection for the DWPF off-gas system and the corrosion evaluation of these materials during pilot scale testing and non-radioactive operations of DWPF Melter #1. Inspection of the off-gas components has not been performed during radioactive operations with the exception of maintenance because of plugging.

  8. Literature review: Assessment of DWPF melter and melter off-gas system lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Reigel, M. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-30

    A glass melter for use in processing radioactive waste is a challenging environment for the materials of construction (MOC) resulting from a combination of high temperatures, chemical attack, and erosion/corrosion; therefore, highly engineered materials must be selected for this application. The focus of this report is to review the testing and evaluations used in the selection of the Defense Waste Processing Facility (DWPF), glass contact MOC specifically the Monofrax® K-3 refractory and Inconel® 690 alloy. The degradation or corrosion mechanisms of these materials during pilot scale testing and in-service operation were analyzed over a range of oxidizing and reducing flowsheets; however, DWPF has primarily processed a reducing flowsheet (i.e., Fe2+/ΣFe of 0.09 to 0.33) since the start of radioactive operations. This report also discusses the materials selection for the DWPF off-gas system and the corrosion evaluation of these materials during pilot scale testing and non-radioactive operations of DWPF Melter #1. Inspection of the off-gas components has not been performed during radioactive operations with the exception of maintenance because of plugging.

  9. Proposed Strategies for DWPF Melter Off-Gas Surge Control

    International Nuclear Information System (INIS)

    CHOI, ALEXANDERS.

    2004-01-01

    Off-gas surging is inherent to the operation of slurry-fed melters. Although the melter design and the feed chemistry are both known to significantly affect off-gas surging, the frequency and intensity of surges are in essence unpredictable. In typical off-gas surges, both condensable and non condensable flows spike simultaneously. Condensable or steam surges have been observed to occur as the boiling water layer occasionally falls into the crevices of the cold cap or flows over the edges of the cold cap, thereby coming in contact with the melt surface. The resulting steam surges can pressurize the melter considerably and, therefore, are responsible for the bulk of pressure transients that propagate throughout the off-gas system. The non condensable surges occur as the calcine gases that have been accumulating within the cold cap finally build up enough pressure to be released through the temporary openings of the cold cap. The analysis of off-gas data has shown that over 90 of the gas released during a surge is due to steam.1 Therefore, it is essential to have a large inventory of water in the cold cap for any significant pressure spikes to occur. With the Melter 2 vapor space temperature typically running at 720C, the water layer in the cold cap will quickly evaporate once the feeding stops, and the potential for any large pressure spikes should practically cease to exist. The analysis also showed that large pressure spikes well above 2 inches H2O cannot occur under the steam surge scenarios described above. More severe conditions should prevail and one such condition would be that the feed materials form a mound with a growing lake on top, while the melt below remains very fluidic due to its low viscosity, thus resulting in greater movements both in the lateral as well as vertical directions. Once the mound begins to grow, its rate should accelerate, since the heat transfer rate to the upper regions of the cold cap is inversely proportional to the cold cap

  10. Off-gas chemistry study of melter feed by Springborn Laboratories

    International Nuclear Information System (INIS)

    Crow, K.R.

    1985-01-01

    The purpose of the off-gas chemistry study of melter feed samples was to support and help substantiate glass melter thermochemistry models developed for the DWPF. Both sludge-only and sludge-precipitate feed samples were analyzed. Each slurry sample was pyrolyzed at temperatures from 150 to 1000 0 C in air and inert atmospheres, and the head space products were analyzed by chromatographic and mass spectrometric methods. Thermogravimetric, differential scanning calorimetric and Fourier transform infrared analyses were also performed on each sample. There were no unusually high exothermic reactions that would be cause for concern in the DWPF melter. Results for two types of sludge-precipitate feed were compared. One type contained simulated precipitate hydrolysis aqueous (PHA) product as fed to the SCM-2 melter. The second type contained PHA from the lab-scale acid hydrolysis reactor in 677-T. A major difference between the two types was a small, but distinct, presence of higher aromatics in gas from feed with reactor-produced PHA. This feed also evolved more CO and CO 2 than feed with simulated PHA at high pyrolytic temperatures (>750 0 C). Recent analyses have identified the higher boiling aromatics in reactor-produced PHA as primarily diphenylamine and p-terphenyl. These compounds will be included in future PHA simulations that are fed to research melters. Under an inert atmosphere, benzene and phenol were the two most abundant organics evolved during pyrolysis of sludge-precipitate feed

  11. Improved mixing and sampling systems for vitrification melter feeds

    International Nuclear Information System (INIS)

    Ebadian, M.A.

    1998-01-01

    This report summarizes the methods used and results obtained during the progress of the study of waste slurry mixing and sampling systems during fiscal year 1977 (FY97) at the Hemispheric Center for Environmental Technology (HCET) at Florida International University (FIU). The objective of this work is to determine optimal mixing configurations and operating conditions as well as improved sampling technology for defense waste processing facility (DWPF) waste melter feeds at US Department of Energy (DOE) sites. Most of the research on this project was performed experimentally by using a tank mixing configuration with different rotating impellers. The slurry simulants for the experiments were prepared in-house based on the properties of the DOE sites' typical waste slurries. A sampling system was designed to withdraw slurry from the mixing tank. To obtain insight into the waste mixing process, the slurry flow in the mixing tank was also simulated numerically by applying computational fluid dynamics (CFD) methods. The major parameters investigated in both the experimental and numerical studies included power consumption of mixer, mixing time to reach slurry uniformity, slurry type, solids concentration, impeller type, impeller size, impeller rotating speed, sampling tube size, and sampling velocities. Application of the results to the DWPF melter feed preparation process will enhance and modify the technical base for designing slurry transportation equipment and pipeline systems. These results will also serve as an important reference for improving waste slurry mixing performance and melter operating conditions. These factors will contribute to an increase in the capability of the vitrification process and the quality of the waste glass

  12. Rheology enhancement for remediated PX6 melter feed

    International Nuclear Information System (INIS)

    Marek, J.C.; Eibling, R.E.

    1996-01-01

    This document is referenced in WSRC-TR-94-0556. This memorandum summarizes results of experimental work performed on the original IDMS PX6 melter feed, the remediated IDMS PX6 melter feed, and melter feeds produced in a laboratory simulation to duplicate the IDMS remediation as well as the experimental results on the caustic treatment to enhance the rheology. Characterization of the products of excess caustic addition and what steps to take if excess caustic is inadvertently added to the IDMS PX6 melter feed are also discussed

  13. Feed process studies: Research-Scale Melter

    Energy Technology Data Exchange (ETDEWEB)

    Whittington, K.F.; Seiler, D.K.; Luey, J.; Vienna, J.D.; Sliger, W.A.

    1996-09-01

    In support of a two-phase approach to privatizing the processing of hazardous and radioactive waste at Hanford, research-scale melter (RSM) experiments were conducted to determine feed processing characteristics of two potential privatization Phase 1 high-level waste glass formulations and to determine if increased Ag, Te, and noble metal amounts would have bad effects. Effects of feed compositions and process conditions were examined for processing rate, cold cap behavior, off-gas, and glass properties. The 2 glass formulations used were: NOM-2 with adjusted waste loading (all components except silica and soda) of 25 wt%, and NOM-3 (max waste loaded glass) with adjusted waste loading of 30 wt%. The 25 wt% figure is the minimum required in the privatization Request for Proposal. RSM operated for 19 days (5 runs). 1010 kg feed was processed, producing 362 kg glass. Parts of runs 2 and 3 were run at 10 to 30 degrees above the nominal temperature 1150 C, with the most significant processing rate increase in run 3. Processing observations led to the choice of NOM-3 for noble metal testing in runs 4 and 5. During noble metal testing, processing rates fell 50% from baseline. Destructive analysis showed that a layer of noble metals and noble metal oxides settled on the floor of the melter, leading to current ``channeling`` which allowed the top section to cool, reducing production rates.

  14. Feed process studies: Research-Scale Melter

    International Nuclear Information System (INIS)

    Whittington, K.F.; Seiler, D.K.; Luey, J.; Vienna, J.D.; Sliger, W.A.

    1996-09-01

    In support of a two-phase approach to privatizing the processing of hazardous and radioactive waste at Hanford, research-scale melter (RSM) experiments were conducted to determine feed processing characteristics of two potential privatization Phase 1 high-level waste glass formulations and to determine if increased Ag, Te, and noble metal amounts would have bad effects. Effects of feed compositions and process conditions were examined for processing rate, cold cap behavior, off-gas, and glass properties. The 2 glass formulations used were: NOM-2 with adjusted waste loading (all components except silica and soda) of 25 wt%, and NOM-3 (max waste loaded glass) with adjusted waste loading of 30 wt%. The 25 wt% figure is the minimum required in the privatization Request for Proposal. RSM operated for 19 days (5 runs). 1010 kg feed was processed, producing 362 kg glass. Parts of runs 2 and 3 were run at 10 to 30 degrees above the nominal temperature 1150 C, with the most significant processing rate increase in run 3. Processing observations led to the choice of NOM-3 for noble metal testing in runs 4 and 5. During noble metal testing, processing rates fell 50% from baseline. Destructive analysis showed that a layer of noble metals and noble metal oxides settled on the floor of the melter, leading to current ''channeling'' which allowed the top section to cool, reducing production rates

  15. EVALUATION OF MIXING IN THE SLURRY MIX EVAPORATOR AND MELTER FEED TANK

    International Nuclear Information System (INIS)

    MARINIK, ANDREW

    2004-01-01

    The Defense Waste Processing Facility (DWPF) vitrifies High Level radioactive Waste (HLW) currently stored in underground tanks at the Savannah River Site (SRS). The HLW currently being processed is a waste sludge composed primarily of metal hydroxides and oxides in caustic slurry. These slurries are typically characterized as Bingham Plastic fluids. The HLW undergoes a pretreatment process in the Chemical Process Cell (CPC) at DWPF. The processed HLW sludge is then transferred to the Sludge Receipt and Adjustment Tank (SRAT) where it is acidified with nitric and formic acid then evaporated to concentrate the solids. Reflux boiling is used to strip mercury from the waste and then the waste is transferred to the Slurry Mix Evaporator tank (SME). Glass formers are added as a frit slurry to the SME to prepare the waste for vitrification. This mixture is evaporated in the SME to the final concentration target. The frit slurry mixture is then transferred to the Melter Feed Tank (MFT) to be fed to the melter

  16. MODELING THE IMPACT OF ELEVATED MERCURY IN DEFENSE WASTE PROCESSING FACILITY MELTER FEED ON THE MELTER OFF-GAS SYSTEM-PRELIMINARY REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Zamecnik, J.; Choi, A.

    2010-08-18

    The Defense Waste Processing Facility (DWPF) is currently evaluating an alternative Chemical Process Cell (CPC) flowsheet to increase throughput. It includes removal of the steam-stripping step, which would significantly reduce the CPC processing time and lessen the sampling needs. However, its downside would be to send 100% of the mercury that comes in with the sludge straight to the melter. For example, the new mercury content in the Sludge Batch 5 (SB5) melter feed is projected to be 25 times higher than that in the SB4 with nominal steam stripping of mercury. This task was initiated to study the impact of the worst-case scenario of zero-mercury-removal in the CPC on the DWPF melter offgas system. It is stressed that this study is intended to be scoping in nature, so the results presented in this report are preliminary. In order to study the impact of elevated mercury levels in the feed, it is necessary to be able to predict how mercury would speciate in the melter exhaust under varying melter operating conditions. A homogeneous gas-phase oxidation model of mercury by chloride was developed to do just that. The model contains two critical parameters pertaining to the partitioning of chloride among HCl, Cl, Cl{sub 2}, and chloride salts in the melter vapor space. The values for these parameters were determined at two different melter vapor space temperatures by matching the calculated molar ratio of HgCl (or Hg{sub 2}Cl{sub 2}) to HgCl{sub 2} with those measured during the Experimental-Scale Ceramic Melter (ESCM) tests run at the Pacific Northwest National Laboratory (PNNL). The calibrated model was then applied to the SB5 simulant used in the earlier flowsheet study with an assumed mercury stripping efficiency of zero; the molar ratio of Cl-to-Hg in the resulting melter feed was only 0.4, compared to 12 for the ESCM feeds. The results of the model run at the indicated melter vapor space temperature of 650 C (TI4085D) showed that due to excessive shortage of

  17. Modeling The Impact Of Elevated Mercury In Defense Waste Processing Facility Melter Feed On The Melter Off-Gas System - Preliminary Report

    International Nuclear Information System (INIS)

    Zamecnik, J.; Choi, A.

    2009-01-01

    The Defense Waste Processing Facility (DWPF) is currently evaluating an alternative Chemical Process Cell (CPC) flowsheet to increase throughput. It includes removal of the steam-stripping step, which would significantly reduce the CPC processing time and lessen the sampling needs. However, its downside would be to send 100% of the mercury that come in with the sludge straight to the melter. For example, the new mercury content in the Sludge Batch 5 (SB5) melter feed is projected to be 25 times higher than that in the SB4 with nominal steam stripping of mercury. This task was initiated to study the impact of the worst-case scenario of zero-mercury-removal in the CPC on the DWPF melter off-gas system. It is stressed that this study is intended to be scoping in nature, so the results presented in this report are preliminary. In order to study the impact of elevated mercury levels in the feed, it is necessary to be able to predict how mercury would speciate in the melter exhaust under varying melter operating conditions. A homogeneous gas-phase oxidation model of mercury by chloride was developed to do just that. The model contains two critical parameters pertaining to the partitioning of chloride among HCl, Cl, Cl 2 , and chloride salts in the melter vapor space. The values for these parameters were determined at two different melter vapor space temperatures by matching the calculated molar ratio of HgCl (or Hg 2 Cl 2 ) to HgCl 2 with those measured during the Experimental-Scale Ceramic Melter (ESCM) tests run at the Pacific Northwest National Laboratory (PNNL). The calibrated model was then applied to the SB5 simulant used in the earlier flowsheet study with an assumed mercury stripping efficiency of zero; the molar ratio of Cl-to-Hg in the resulting melter feed was only 0.4, compared to 12 for the ESCM feeds. The results of the model run at the indicated melter vapor space temperature of 650 C (TI4085D) showed that due to excessive shortage of chloride, only 6% of

  18. DWPF waste glass Product Composition Control System

    International Nuclear Information System (INIS)

    Brown, K.G.; Postles, R.L.

    1992-01-01

    The Defense Waste Processing Facility (DWPF) will be used to blend aqueous radwaste (PHA) with solid radwaste (Sludge) in a waste receipt vessel (the SRAT). The resulting SRAT material is transferred to the SME an there blended with ground glass (Frit) to produce a batch of melter feed slurry. The SME material is passed to a hold tank (the MFT) which is used to continuously feed the DWPF melter. The melter. The melter produces a molten glass wasteform which is poured into stainless steel canisters for cooling and, ultimately, shipment to and storage in a geologic repository. The Product Composition Control System (PCCS) is the system intended to ensure that the melt will be processible and that the glass wasteform will be acceptable. This document provides a description of this system

  19. Melter feed viscosity during conversion to glass: Comparison between low-activity waste and high-level waste feeds

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Tongan [Pacific Northwest National Laboratory, Richland Washington; Chun, Jaehun [Pacific Northwest National Laboratory, Richland Washington; Dixon, Derek R. [Pacific Northwest National Laboratory, Richland Washington; Kim, Dongsang [Pacific Northwest National Laboratory, Richland Washington; Crum, Jarrod V. [Pacific Northwest National Laboratory, Richland Washington; Bonham, Charles C. [Pacific Northwest National Laboratory, Richland Washington; VanderVeer, Bradley J. [Pacific Northwest National Laboratory, Richland Washington; Rodriguez, Carmen P. [Pacific Northwest National Laboratory, Richland Washington; Weese, Brigitte L. [Pacific Northwest National Laboratory, Richland Washington; Schweiger, Michael J. [Pacific Northwest National Laboratory, Richland Washington; Kruger, Albert A. [U.S. Department of Energy, Office of River Protection, Richland Washington; Hrma, Pavel [Pacific Northwest National Laboratory, Richland Washington

    2017-12-07

    During nuclear waste vitrification, a melter feed (generally a slurry-like mixture of a nuclear waste and various glass forming and modifying additives) is charged into the melter where undissolved refractory constituents are suspended together with evolved gas bubbles from complex reactions. Knowledge of flow properties of various reacting melter feeds is necessary to understand their unique feed-to-glass conversion processes occurring within a floating layer of melter feed called a cold cap. The viscosity of two low-activity waste (LAW) melter feeds were studied during heating and correlated with volume fractions of undissolved solid phase and gas phase. In contrast to the high-level waste (HLW) melter feed, the effects of undissolved solid and gas phases play comparable roles and are required to represent the viscosity of LAW melter feeds. This study can help bring physical insights to feed viscosity of reacting melter feeds with different compositions and foaming behavior in nuclear waste vitrification.

  20. Preliminary melter performance assessment report

    International Nuclear Information System (INIS)

    Elliott, M.L.; Eyler, L.L.; Mahoney, L.A.; Cooper, M.F.; Whitney, L.D.; Shafer, P.J.

    1994-08-01

    The Melter Performance Assessment activity, a component of the Pacific Northwest Laboratory's (PNL) Vitrification Technology Development (PVTD) effort, was designed to determine the impact of noble metals on the operational life of the reference Hanford Waste Vitrification Plant (HWVP) melter. The melter performance assessment consisted of several activities, including a literature review of all work done with noble metals in glass, gradient furnace testing to study the behavior of noble metals during the melting process, research-scale and engineering-scale melter testing to evaluate effects of noble metals on melter operation, and computer modeling that used the experimental data to predict effects of noble metals on the full-scale melter. Feed used in these tests simulated neutralized current acid waste (NCAW) feed. This report summarizes the results of the melter performance assessment and predicts the lifetime of the HWVP melter. It should be noted that this work was conducted before the recent Tri-Party Agreement changes, so the reference melter referred to here is the Defense Waste Processing Facility (DWPF) melter design

  1. RHEOLOGICAL AND ELEMENTAL ANALYSES OF SIMULANT SB5 SLURRY MIX EVAPORATOR-MELTER FEED TANK SLURRIES

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, A.

    2010-02-08

    The Defense Waste Processing Facility (DWPF) will complete Sludge Batch 5 (SB5) processing in fiscal year 2010. DWPF has experienced multiple feed stoppages for the SB5 Melter Feed Tank (MFT) due to clogs. Melter throughput is decreased not only due to the feed stoppage, but also because dilution of the feed by addition of prime water (about 60 gallons), which is required to restart the MFT pump. SB5 conditions are different from previous batches in one respect: pH of the Slurry Mix Evaporator (SME) product (9 for SB5 vs. 7 for SB4). Since a higher pH could cause gel formation, due in part to greater leaching from the glass frit into the supernate, SRNL studies were undertaken to check this hypothesis. The clogging issue is addressed by this simulant work, requested via a technical task request from DWPF. The experiments were conducted at Aiken County Technology Laboratory (ACTL) wherein a non-radioactive simulant consisting of SB5 Sludge Receipt and Adjustment Tank (SRAT) product simulant and frit was subjected to a 30 hour SME cycle at two different pH levels, 7.5 and 10; the boiling was completed over a period of six days. Rheology and supernate elemental composition measurements were conducted. The caustic run exhibited foaming once, after 30 minutes of boiling. It was expected that caustic boiling would exhibit a greater leaching rate, which could cause formation of sodium aluminosilicate and would allow gel formation to increase the thickness of the simulant. Xray Diffraction (XRD) measurements of the simulant did not detect crystalline sodium aluminosilicate, a possible gel formation species. Instead, it was observed that caustic conditions, but not necessarily boiling time, induced greater thickness, but lowered the leach rate. Leaching consists of the formation of metal hydroxides from the oxides, formation of boric acid from the boron oxide, and dissolution of SiO{sub 2}, the major frit component. It is likely that the observed precipitation of Mg

  2. Methods of Off-Gas Flammability Control for DWPF Melter Off-Gas System at Savannah River Site

    International Nuclear Information System (INIS)

    Choi, A.S.; Iverson, D.C.

    1996-01-01

    Several key operating variables affecting off-gas flammability in a slurry-fed radioactive waste glass melter are discussed, and the methods used to prevent potential off-gas flammability are presented. Two models have played a central role in developing such methods. The first model attempts to describe the chemical events occurring during the calcining and melting steps using a multistage thermodynamic equilibrium approach, and it calculates the compositions of glass and calcine gases. Volatile feed components and calcine gases are fed to the second model which then predicts the process dynamics of the entire melter off-gas system including off-gas flammability under both steady state and various transient operating conditions. Results of recent simulation runs are also compared with available data

  3. Copper solubility in DWPF, Batch 1 waste glass: Update report

    International Nuclear Information System (INIS)

    Schumacker, R.F.

    1992-01-01

    The ''Late Washing'' Step in the processing of precipitate will require the use of additional copper formate in the Precipitate Reactor to catalyze the hydrolysis reaction. The increased copper concentration in the melter feed increases the potential for metal precipitation during the vitrification of the melter feed. This report describes recent results with a conservative glass selected from the DWPF acceptable region in the Batch 1 Variability Study

  4. Slurry feed variability in West Valley's melter feed tank and sampling system

    International Nuclear Information System (INIS)

    Fow, C.L.; Kurath, D.E.; Pulsipher, B.A.; Bauer, B.P.

    1989-04-01

    The present plan for disposal of high-level wastes at West Valley is to vitrify the wastes for disposal in deep geologic repository. The vitrification process involves mixing the high-level wastes with glass-forming chemicals and feeding the resulting slurry to a liquid-fed ceramic melter. Maintaining the quality of the glass product and proficient melter operation depends on the ability of the melter feed system to produce and maintain a homogeneous mixture of waste and glass-former materials. To investigate the mixing properties of the melter feed preparation system at West Valley, a statistically designed experiment was conducted using synthetic melter feed slurry over a range of concentrations. On the basis of the statistical data analysis, it was found that (1) a homogeneous slurry is produced in the melter feed tank, (2) the liquid-sampling system provides slurry samples that are statistically different from the slurry in the tank, and (3) analytical measurements are the major source of variability. A statistical quality control program for the analytical laboratory and a characterization test of the actual sampling system is recommended. 1 ref., 5 figs., 1 tab

  5. Rheological Studies on Pretreated Feed and Melter Feed from AW-101 and AN-107

    International Nuclear Information System (INIS)

    Bredt, Paul R; Swoboda, Robert G

    2001-01-01

    Rheological and physical properties testing were conducted on actual AN-107 and AW-101 pretreated feed samples prior to the addition of glass formers. Analyses were repeated following the addition of glass formers. The AN-107 and AW-101 pretreated feeds were tested at the target sodium values of nominally 6, 8, and 10 M. The AW-101 melter feeds were tested at these same concentrations, while the AN-107 melter feeds were tested at 5, 6, and 8 M with respect to sodium. These data on actual waste are required to validate and qualify results obtained with simulants

  6. EFFECT OF MELTER-FEED-MAKEUP ON VITRIFICATION PROCESS

    International Nuclear Information System (INIS)

    Kruger, A.A.; Hrma, P.R.; Schweiger, M.J.; Humrickhouse, C.J.; Moody, J.A.; Tate, R.M.; Tegrotenhuis, N.E.; Arrigoni, B.M.; Rodriguez, C.P.

    2009-01-01

    Increasing the rate of glass processing in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) will allow shortening the life cycle of waste cleanup at the Hanford Site. While the WTP melters have approached the limit of increasing the rate of melting by enhancing the heat transfer rate from molten glass to the cold cap, a substantial improvement can still be achieved by accelerating the feed-to-glass conversion kinetics. This study investigates how the feed-to-glass conversion process responds to the feed makeup. By identifying the means of control of primary foam formation and silica grain dissolution, it provides data needed for a meaningful and economical design of large-scale experiments aimed at achieving faster melting

  7. Redox control of electric melters with complex feed compositions. Part I: analytical methods and models

    International Nuclear Information System (INIS)

    Bickford, D.F.; Diemer, R.B. Jr.

    1985-01-01

    The redox state of glass from electric melters with complex feed compositions is determined by balance between gases above the melt, and transition metals and organic compounds in the feed. Part I discusses experimental and computational methods of relating flowrates and other melter operating conditions to the redox state of glass, and composition of the melter offgas. Computerized thermodynamic computational methods are useful in predicting the sequence and products of redox reactions and in assessing individual process variations. Melter redox state can be predicted by combining monitoring of melter operating conditions, redox measurement of fused melter feed samples, and periodic redox measurement of product. Mossbauer spectroscopy, and other methods which measure Fe(II)/Fe(III) in glass, can be used to measure melter redox state. Part II develops preliminary operating limits for the vitrification of High-Level Radioactive Waste. Limits on reducing potential to preclude the accumulation of combustible gases, accumulation of sulfides and selenides, and degradation of melter components are the most critical. Problems associated with excessively oxidizing conditions, such as glass foaming and potential ruthenium volatility, are controlled when sufficient formic acid is added to adjust melter feed rheology

  8. Corrosion Testing of Monofrax K-3 Refractory in Defense Waste Processing Facility (DWPF) Alternate Reductant Feeds

    Energy Technology Data Exchange (ETDEWEB)

    Williams, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jantzen, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Burket, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-04-06

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) uses a combination of reductants and oxidants while converting high level waste (HLW) to a borosilicate waste form. A reducing flowsheet is maintained to retain radionuclides in their reduced oxidation states which promotes their incorporation into borosilicate glass. For the last 20 years of processing, the DWPF has used formic acid as the main reductant and nitric acid as the main oxidant. During reaction in the Chemical Process Cell (CPC), formate and formic acid release measurably significant H2 gas which requires monitoring of certain vessel’s vapor spaces. A switch to a nitric acid-glycolic acid (NG) flowsheet from the nitric-formic (NF) flowsheet is desired as the NG flowsheet releases considerably less H2 gas upon decomposition. This would greatly simplify DWPF processing from a safety standpoint as close monitoring of the H2 gas concentration could become less critical. In terms of the waste glass melter vapor space flammability, the switch from the NF flowsheet to the NG flowsheet showed a reduction of H2 gas production from the vitrification process as well. Due to the positive impact of the switch to glycolic acid determined on the flammability issues, evaluation of the other impacts of glycolic acid on the facility must be examined.

  9. HWVP NCAW melter feed rheology FY 1993 testing and analyses: Letter report

    International Nuclear Information System (INIS)

    Smith, P.A.

    1996-03-01

    The Hanford Waste Vitrification Plant (HWVP) program has been established to immobilize selected Hanford nuclear wastes before shipment to a geologic repository. The HWVP program is directed by the U.S. Department of Energy (DOE). The Pacific Northwest Laboratory (PNL) provides waste processing and vitrification technology to assist the design effort. The focus of this letter report is melter feed rheology, Process/Product Development, which is part of the Task in the PNL HWVP Technology Development (PHTD) Project. Specifically, the melter feed must be transported to the liquid fed ceramic melter (LFCM) to ensure HWVP operability and the manufacture of an immobilized waste form. The objective of the PHTD Project slurry flow technology development is to understand and correlate dilute and concentrated waste, formatted waste, waste with recycle addition, and melter feed transport properties. The objectives of the work described in this document were to examine frit effects and several processing conditions on melter feed rheology. The investigated conditions included boiling time, pH, noble metal containing melter feed, solids loading, and aging time. The results of these experiments contribute to the understanding of melter feed rheology. This document is organized in eight sections. This section provides the introductory remarks, followed by Section 2.0 that contains conclusions and recommendations. Section 3.0 reviews the scientific principles, and Section 4.0 details the experimental methods. The results and discussion and the review of related rheology data are in Sections 5.0 and 6.0, respectively. Section 7.0, an analysis of NCAW melter feed rheology data, provides an overall review of melter feed with FY 91 frit. References are included in Section 8.0. This letter report satisfies contractor milestone PHTD C93-03.02E, as described in the FY 1993 Pacific Northwest Hanford Laboratory Waste Plant Technology Development (PHTD) Project Work Plan

  10. Two new research melters at the Savannah River Technology Center

    International Nuclear Information System (INIS)

    Gordon, J.R.; Coughlin, J.T.; Minichan, R.L.; Zamecnik, J.R.

    2000-01-01

    The Savannah River Technology Center (SRTC) is a US Department of Energy (DOE) complex leader in the development of vitrification technology. To maintain and expand this SRTC core technology, two new melter systems are currently under construction in SRTC. This paper discusses the development of these two new systems, which will be used to support current as well as future vitrification programs in the DOE complex. The first of these is the new minimelter, which is a joule-heated glass melter intended for experimental melting studies with nonradioactive glass waste forms. Testing will include surrogates of Defense Waste processing Facility (DWPF) high-level wastes. To support the DWPF testing, the new minimelter was scaled to the DWPF melter based on melt surface area. This new minimelter will replace an existing system and provide a platform for the research and development necessary to support the SRTC vitrification core technology mission. The second new melter is the British Nuclear Fuels, Inc., research melter system (BNFL melter), which is a scaled version of the BNFL low-activity-waste (LAW) melter proposed for vitrification of LAW at Hanford. It is designed to process a relatively large amount of actual radiative Hanford tank waste and to gather data on the composition of off-gases that will be generated by the LAW melter. Both the minimelter and BNFL melter systems consist of five primary subsystems: melter vessel, off-gas treatment, feed, power supply, and instrumentation and controls. The configuration and design of these subsystems are tailored to match the current system requirements and provide the flexibility to support future DOE vitrification programs. This paper presents a detailed discussion of the unique design challenges represented by these two new melter systems

  11. PHYSICAL CHARACTERIZATION OF VITREOUS STATE LABORATORY AY102/C106 AND AZ102 HIGH LEVEL WASTE MELTER FEED SIMULANTS (U)

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, E

    2005-03-31

    The objective of this task is to characterize and report specified physical properties and pH of simulant high level waste (HLW) melter feeds (MF) processed through the scaled melters at Vitreous State Laboratories (VSL). The HLW MF simulants characterized are VSL AZ102 straight hydroxide melter feed, VSL AZ102 straight hydroxide rheology adjusted melter feed, VSL AY102/C106 straight hydroxide melter feed, VSL AY102/C106 straight hydroxide rheology adjusted melter feed, and Savannah River National Laboratory (SRNL) AY102/C106 precipitated hydroxide processed sludge blended with glass former chemicals at VSL to make melter feed. The physical properties and pH were characterized using the methods stated in the Waste Treatment Plant (WTP) characterization procedure (Ref. 7).

  12. Radioactive demonstration of DWPF product control strategy

    International Nuclear Information System (INIS)

    Andrews, M.K.; Bibler, N.E.

    1992-01-01

    The effectiveness of the product and process control strategies that will be utilized by the Defense Waste Processing Facility (DWPF) was demonstrated during a campaign in the Shielded Cells Facility (SCF) of the Savannah River Technology Center (SRTC). The remotely operated process included the preparation of the melter feed, vitrification in a slurry-fed 1/100th scale melter and analysis of the glass product both for its composition and durability. The campaign processed approximately 10 kg (on a dry basis) of radioactive sludge from Tank 51. This sludge is representative of the first batch of sludge that will be sent to the DWPF for immobilization into borosilicate glass. Additions to the sludge were made based on calculations using the Product Composition Control System (PCCS). Analysis of the glass produced during the campaign showed that a durable glass was produced with a composition similar to that predicted using the PCCS

  13. Density of simulated americium/curium melter feed solution

    International Nuclear Information System (INIS)

    Rudisill, T.S.

    1997-01-01

    Vitrification will be used to stabilize an americium/curium (Am/Cm) solution presently stored in F-Canyon for eventual transport to Oak Ridge National Laboratory and use in heavy isotope production programs. Prior to vitrification, a series of in-tank oxalate precipitation and nitric/oxalic acid washes will be used to separate these elements and lanthanide fission products from the bulk of the uranium and metal impurities present in the solution. Following nitric acid dissolution and oxalate destruction, the solution will be denitrated and evaporated to a dissolved solids concentration of approximately 100 g/l (on an oxide basis). During the Am/Cm vitrification, an airlift will be used to supply the concentrated feed solution to a constant head tank which drains through a filter and an in-line orifice to the melter. Since the delivery system is sensitive to the physical properties of the feed, a simulated solution was prepared and used to measure the density as a function of temperature between 20 to 70 degrees C. The measured density decreased linearly at a rate of 0.0007 g/cm3/degree C from an average value of 1.2326 g/cm 3 at 20 degrees C to an average value of 1.1973g/cm 3 at 70 degrees C

  14. Density of simulated americium/curium melter feed solution

    Energy Technology Data Exchange (ETDEWEB)

    Rudisill, T.S.

    1997-09-22

    Vitrification will be used to stabilize an americium/curium (Am/Cm) solution presently stored in F-Canyon for eventual transport to Oak Ridge National Laboratory and use in heavy isotope production programs. Prior to vitrification, a series of in-tank oxalate precipitation and nitric/oxalic acid washes will be used to separate these elements and lanthanide fission products from the bulk of the uranium and metal impurities present in the solution. Following nitric acid dissolution and oxalate destruction, the solution will be denitrated and evaporated to a dissolved solids concentration of approximately 100 g/l (on an oxide basis). During the Am/Cm vitrification, an airlift will be used to supply the concentrated feed solution to a constant head tank which drains through a filter and an in-line orifice to the melter. Since the delivery system is sensitive to the physical properties of the feed, a simulated solution was prepared and used to measure the density as a function of temperature between 20 to 70{degrees} C. The measured density decreased linearly at a rate of 0.0007 g/cm3/{degree} C from an average value of 1.2326 g/cm{sup 3} at 20{degrees} C to an average value of 1.1973g/cm{sup 3} at 70{degrees} C.

  15. Crystallization in high level waste (HLW) glass melters: Savannah River Site operational experience

    Energy Technology Data Exchange (ETDEWEB)

    Fox, Kevin M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Peeler, David K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kruger, Albert A. [USDOE Office of River Protection, Richland, WA (United States)

    2015-06-12

    This paper provides a review of the scaled melter testing that was completed for design input to the Defense Waste Processing Facility (DWPF) melter. Testing with prototype melters provided the data to define the DWPF operating limits to avoid bulk (volume) crystallization in the un-agitated DWPF melter and provided the data to distinguish between spinels generated by refractory corrosion versus spinels that precipitated from the HLW glass melt pool. A review of the crystallization observed with the prototype melters and the full-scale DWPF melters (DWPF Melter 1 and DWPF Melter 2) is included. Examples of actual DWPF melter attainment with Melter 2 are given. The intent is to provide an overview of lessons learned, including some example data, that can be used to advance the development and implementation of an empirical model and operating limit for crystal accumulation for a waste treatment and immobilization plant.

  16. Radioactive demonstration of DWPF product control strategy

    International Nuclear Information System (INIS)

    Andrews, M.K.; Bibler, N.E.

    1994-01-01

    The Defense Waste Processing Facility at the Savannah River Site (SRS) will vitrify high-level nuclear waste into borosilicate glass. The waste will be mixed with properly formulated glass-making frit and fed to a melter at 1150 degrees C. Process reliability and product quality are ensured by proper control of the melter feed composition. The effectiveness of the product and process control strategies that will be utilized by the Defense Waste Processing Facility (DWPF) was demonstrated during a campaign in the Shielded Cells Facility of the Savannah River Technology Center (SRTC). The remotely operated process included the preparation of the melter feed, vitrification in a slurry-fed 1/100th scale melter an analysis of the glass product both for its composition an durability. The campaign processed approximately 10 kg (on a dry basis) of radioactive sludge from Tank 51. This sludge is representative of the first batch of sludge that will be sent to the DWPF for immobilization into borosilicate glass. Additions to the sludge were made based on calculations using the Product Composition Control System (PCCS). Analysis of the glass produced during the campaign showed that a durable glass was produced with a composition very close to that predicted using the PCCS. 10 refs., 4 tabs

  17. Effect of melter feed foaming on heat flux to the cold cap

    Energy Technology Data Exchange (ETDEWEB)

    Lee, SeungMin; Hrma, Pavel; Pokorny, Richard; Klouzek, Jaroslav; VanderVeer, Bradley J.; Dixon, Derek R.; Luksic, Steven A.; Rodriguez, Carmen P.; Chun, Jaehun; Schweiger, Michael J.; Kruger, Albert A.

    2017-12-01

    The glass production rate, which is crucial for the nuclear waste cleanup lifecycle, is influenced by the chemical and mineralogical nature of melter feed constituents. The choice of feed materials affects both the conversion heat and the thickness of the foam layer that forms at the bottom of the cold cap and controls the heat flow from molten glass. We demonstrate this by varying the alumina source, namely, substituting boehmite or corundum for gibbsite, in a high-alumina high-level-waste melter feed. The extent of foaming was determined using the volume expansion test and the conversion heat with differential scanning calorimetry. Evolved gas analysis was used to identify gases responsible for the formation of primary and secondary foam. The foam thickness, a critical factor in the rate of melting, was estimated using known values of heat conductivities and melting rates. The result was in reasonable agreement with the foam thickness experimentally observed in the laboratory-scale melter.

  18. History of the small cylindrical melter

    International Nuclear Information System (INIS)

    Allen, T.L.; Iverson, D.C.; Plodinec, M.J.

    1985-08-01

    The small cylindrical melter (SCM) was designed to provide engineering data useful for operation and design of full-scale glass melters for vitrification of high-level radioactive waste. This melter was part of the research and development program for the Defense Waste Processing Facility (DWPF) at the Savannah River Plant (SRP). Extensive corrosion testing of melter materials of construction (Monofrax K3, Inconel 690), simulated radioactive waste glass characterization, and melter component development were conducted in support of the DWPF full-scale melter design. 66 figs., 14 tabs

  19. Control of high-level radioactive waste-glass melters

    International Nuclear Information System (INIS)

    Bickford, D.F.; Coleman, C.J.

    1990-01-01

    The Defense Waste Processing Facility (DWPF) will immobilize Savannah River Site High Level Waste as a durable borosilicate glass for permanent disposal in a repository. The DWPF will be controlled based on glass composition. The following discussion is a preliminary analysis of the capability of the laboratory methods that can be used to control the glass composition, and the relationships between glass durability and glass properties important to glass melting. The glass durability and processing properties will be controlled by controlling the chemical composition of the glass. The glass composition will be controlled by control of the melter feed transferred from the Slurry Mix Evaporator (SME) to the Melter Feed Tank (MFT). During cold runs, tests will be conducted to demonstrate the chemical equivalence of glass sampled from the pour stream and glass removed from cooled canisters. In similar tests, the compositions of glass produced from slurries sampled from the SME and MFT will be compared to final product glass to determine the statistical relationships between melter feed and glass product. The total error is the combination of those associated with homogeneity in the SME or MFT, sampling, preparation of samples for analysis, instrument calibration, analysis, and the composition/property model. This study investigated the sensitivity of estimation of property data to the combination of variations from sampling through analysis. In this or a similar manner, the need for routine glass product sampling will be minimized, and glass product characteristics will be assured before the melter feed is committed to the melter

  20. Analysis of the DWPF glass pouring system using neural networks

    International Nuclear Information System (INIS)

    Calloway, T.B. Jr.; Jantzen, C.M.

    1997-01-01

    Neural networks were used to determine the sensitivity of 39 selected Melter/Melter Off Gas and Melter Feed System process parameters as related to the Defense Waste Processing Facility (DWPF) Melter Pour Spout Pressure during the overall analysis and resolution of the DWPF glass production and pouring issues. Two different commercial neural network software packages were used for this analysis. Models were developed and used to determine the critical parameters which accurately describe the DWPF Pour Spout Pressure. The model created using a low-end software package has a root mean square error of ± 0.35 inwc ( 2 = 0.77) with respect to the plant data used to validate and test the model. The model created using a high-end software package has a R 2 = 0.97 with respect to the plant data used to validate and test the model. The models developed for this application identified the key process parameters which contribute to the control of the DWPF Melter Pour Spout pressure during glass pouring operations. The relative contribution and ranking of the selected parameters was determined using the modeling software. Neural network computing software was determined to be a cost-effective software tool for process engineers performing troubleshooting and system performance monitoring activities. In remote high-level waste processing environments, neural network software is especially useful as a replacement for sensors which have failed and are costly to replace. The software can be used to accurately model critical remotely installed plant instrumentation. When the instrumentation fails, the software can be used to provide a soft sensor to replace the actual sensor, thereby decreasing the overall operating cost. Additionally, neural network software tools require very little training and are especially useful in mining or selecting critical variables from the vast amounts of data collected from process computers

  1. Modifying the rheological properties of melter feed for the Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Blair, H.T.; McMakin, A.H.

    1986-03-01

    Selected high-level nuclear wastes from the Hanford Site may be vitrified in the future Hanford Waste Vitrification Plant (HWVP) by Rockwell Hanford Company, the contractor responsible for reprocessing and waste management at the Hanford Site. The Pacific Northwest Laboratory (PNL), is responsible for providing technical support for the HWVP. In this capacity, PNL performed rheological evaluations of simulated HWVP feed in order to determine which processing factors could be modified to best optimize the vitrification process. To accomplish this goal, a simulated HWVP feed was first created and characterized. Researchers then evaluated how the chemical and physical form of the glass-forming additives affected the rheological properties and melting behavior of melter feed prepared with the simulated HWVP feed. The effects of adding formic acid to the waste were also evaluated. Finally, the maximum melter feed concentration with acceptable rheological properties was determined

  2. Hydrogen generation and foaming during tests in the GFPS simulating DWPF operations with Tank 42 sludge and CST

    Energy Technology Data Exchange (ETDEWEB)

    Koopman, D.C.

    1999-12-08

    This report summarizes the pilot-scale research requested by the salt disposition team to examine the effect of crystalline silicotitanate (CST) resin with adsorbed noble metals on the maximum hydrogen generation rate produced during the DWPF melter feed preparation processes.

  3. Hydrogen generation and foaming during tests in the GFPS simulating DWPF operations with Tank 42 sludge and CST

    International Nuclear Information System (INIS)

    Koopman, D.C.

    1999-01-01

    This report summarizes the pilot-scale research requested by the salt disposition team to examine the effect of crystalline silicotitanate (CST) resin with adsorbed noble metals on the maximum hydrogen generation rate produced during the DWPF melter feed preparation processes

  4. SCIX IMPACT ON DWPF CPC

    Energy Technology Data Exchange (ETDEWEB)

    Koopman, D.

    2011-07-14

    A program was conducted to systematically evaluate potential impacts of the proposed Small Column Ion Exchange (SCIX) process on the Defense Waste Processing Facility (DWPF) Chemical Processing Cell (CPC). The program involved a series of interrelated tasks. Past studies of the impact of crystalline silicotitanate (CST) and monosodium titanate (MST) on DWPF were reviewed. Paper studies and material balance calculations were used to establish reasonable bounding levels of CST and MST in sludge. Following the paper studies, Sludge Batch 10 (SB10) simulant was modified to have both bounding and intermediate levels of MST and ground CST. The SCIX flow sheet includes grinding of the CST which is larger than DWPF frit when not ground. Nominal ground CST was not yet available, therefore a similar CST ground previously in Savannah River National Laboratory (SRNL) was used. It was believed that this CST was over ground and that it would bound the impact of nominal CST on sludge slurry properties. Lab-scale simulations of the DWPF CPC were conducted using SB10 simulants with no, intermediate, and bounding levels of CST and MST. Tests included both the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles. Simulations were performed at high and low acid stoichiometry. A demonstration of the extended CPC flowsheet was made that included streams from the site interim salt processing operations. A simulation using irradiated CST and MST was also completed. An extensive set of rheological measurements was made to search for potential adverse consequences of CST and MST and slurry rheology in the CPC. The SCIX CPC impact program was conducted in parallel with a program to evaluate the impact of SCIX on the final DWPF glass waste form and on the DWPF melter throughput. The studies must be considered together when evaluating the full impact of SCIX on DWPF. Due to the fact that the alternant flowsheet for DWPF has not been selected, this study did not

  5. Control of radioactive waste-glass melters

    International Nuclear Information System (INIS)

    Bickford, D.F.; Smith, P.K.; Hrma, P.; Bowan, B.W.

    1987-01-01

    Radioactive waste-glass melters require physical control limits and redox control of glass to assure continuous operation, and maximize production rates. Typical waste-glass melter operating conditions, and waste-glass chemical reaction paths are discussed. Glass composition, batching and melter temperature control are used to avoid the information of phases which are disruptive to melting or reduce melter life. The necessity and probable limitations of control for electric melters with complex waste feed compositions are discussed. Preliminary control limits, their bases, and alternative control methods are described for use in the Defense Waste Processing Facility (DWPF) at the US Department of Energy's Savannah River Plant (SRP), and at the West Valley Demonstration Project (WVDP). Slurries of simulated high level radioactive waste and ground glass frit or glass formers have been isothermally reacted and analyzed to identify the sequence of the major chemical reactions in waste vitrification, and their effect on waste-glass production rates. Relatively high melting rates of waste batches containing mixtures of reducing agents (formic acid, sucrose) and nitrates are attributable to exothermic reactions which occur at critical stages in the vitrification process. The effect of foaming on waste glass production rates is analyzed, and limits defined for existing waste-glass melters, based upon measurable thermophysical properties. Through balancing the high nitrate wastes of the WVDP with reducing agents, the high glass melting rates and sustained melting without foaming required for successful WVDP operations have been demonstrated. 65 refs., 4 figs., 15 tabs

  6. Evaluation of vitrification factors from DWPF's macro-batch 1

    International Nuclear Information System (INIS)

    Edwards, T.B.

    2000-01-01

    The Defense Waste Processing Facility (DWPF) is evaluating new sampling and analytical methods that may be used to support future Slurry Mix Evaporator (SME) batch acceptability decisions. This report uses data acquired during DWPF's processing of macro-batch 1 to determine a set of vitrification factors covering several SME and Melter Feed Tank (MFT) batches. Such values are needed for converting the cation measurements derived from the new methods to a ''glass'' basis. The available data from macro-batch 1 were used to examine the stability of these vitrification factors, to estimate their uncertainty over the course of a macro-batch, and to provide a recommendation on the use of a single factor for an entire macro-batch. The report is in response to Technical Task Request HLW/DWPF/TTR-980015

  7. Determination of heat conductivity and thermal diffusivity of waste glass melter feed: Extension to high temperatures

    International Nuclear Information System (INIS)

    Rice, Jarrett A.; Pokorny, Richard; Schweiger, Michael J.; Hrma, Pavel R.

    2014-01-01

    The heat conductivity (λ) and the thermal diffusivity (a) of reacting glass batch, or melter feed, control the heat flux into and within the cold cap, a layer of reacting material floating on the pool of molten glass in an all-electric continuous waste glass melter. After previously estimating λ of melter feed at temperatures up to 680 deg C, we focus in this work on the λ(T) function at T > 680 deg C, at which the feed material becomes foamy. We used a customized experimental setup consisting of a large cylindrical crucible with an assembly of thermocouples, which monitored the evolution of the temperature field while the crucible with feed was heated at a constant rate from room temperature up to 1100°C. Approximating measured temperature profiles by polynomial functions, we used the heat transfer equation to estimate the λ(T) approximation function, which we subsequently optimized using the finite-volume method combined with least-squares analysis. The heat conductivity increased as the temperature increased until the feed began to expand into foam, at which point the conductivity dropped. It began to increase again as the foam turned into a bubble-free glass melt. We discuss the implications of this behavior for the mathematical modeling of the cold cap

  8. Impact of Spherical Frit Beads on Simulated DWPF Slurries

    International Nuclear Information System (INIS)

    SMITH, MICHAEL

    2005-01-01

    It has been shown that the rheological properties of simulated Defense Waste Processing Facility (DWPF) melter feed with the glass former frit as mostly (90 weight percent) solid spherical particles (referred to as beads) were improved as the feed was less viscous as compared to DWPF melter feed that contained the normal irregular shaped frit particles. Because the physical design of the DWPF Slurry Mix Evaporator (SME), Melter Feed Tank (MFT), and melter feed loop are fixed, the impact of changing the rheology might be very beneficial. Most importantly, higher weight percent total solids feed might be processed by reducing the rheological properties (specifically yield stress) of the feed. Additionally, if there are processing problems, such as air entrainment or pumping, these problems might be alleviated by reducing the rheological properties, while maintaining targeted throughputs. Rheology modifiers are chemical, physical, or a combination of the two and can either thin or thicken the rheology of the targeted slurry. The beads are classified as a physical rheological modifier in this case. Even though the improved rheological properties of the feed in the above mentioned DWPF tanks could be quite beneficial, it is the possibility of increased melt rate that is the main driver for the use of beaded glass formers. By improving the rheological properties of the feed, the weight percent solids of the feed could be increased. This higher weight percent solids (less water) feed could be processed faster by the melter as less energy would be required to evaporate the water, and more would be available for the actual melting of the waste and the frit. In addition, the use of beads to thin the feed could possibly allow for the use of a lower targeted acid stoichiometry in the feed preparation process (if in fact acid stoichiometry is being driven by feed rheology as opposed to feed chemistry). Previous work by the Savannah River National Laboratory (SRNL) with the lab

  9. Off-gas characteristics of defense waste vitrification using liquid-fed Joule-heated ceramic melters

    International Nuclear Information System (INIS)

    Goles, R.W.; Sevigny, G.J.

    1983-09-01

    Off-gas and effluent characterization studies have been established as part of a PNL Liquid-Fed Ceramic Melter development program supporting the Savannah River Laboratory Defense Waste Processing Facility (SRL-DWPF). The objectives of these studies were to characterize the gaseous and airborne emission properties of liquid-fed joule-heated melters as a function of melter operational parameters and feed composition. All areas of off-gas interest and concern including effluent characterization, emission control, flow rate behavior and corrosion effects have been studied using alkaline and formic-acid based feed compositions. In addition, the behavioral patterns of gaseous emissions, the characteristics of melter-generated aerosols and the nature and magnitude of melter effluent losses have been established under a variety of feeding conditions with and without the use of auxiliary plenum heaters. The results of these studies have shown that particulate emissions are responsible for most radiologically important melter effluent losses. Melter-generated gases have been found to be potentially flammable as well as corrosive. Hydrogen and carbon monoxide present the greatest flammability hazard of the combustibles produced. Melter emissions of acidic volatile compounds of sulfur and the halogens have been responsible for extensive corrosion observed in melter plenums and in associated off-gas lines and processing equipment. The use of auxiliary plenum heating has had little effect upon melter off-gas characteristics other than reducing the concentrations of combustibles

  10. Effect of melter feed foaming on heat flux to the cold cap

    Science.gov (United States)

    Lee, SeungMin; Hrma, Pavel; Pokorny, Richard; Klouzek, Jaroslav; VanderVeer, Bradley J.; Dixon, Derek R.; Luksic, Steven A.; Rodriguez, Carmen P.; Chun, Jaehun; Schweiger, Michael J.; Kruger, Albert A.

    2017-12-01

    The glass production rate, which is crucial for the nuclear waste cleanup lifecycle, is influenced by the chemical and mineralogical nature of melter feed constituents. The choice of feed materials affects both the conversion heat and the thickness of the foam layer that forms at the bottom of the cold cap and controls the heat flow from molten glass. We demonstrate this by varying the alumina source, namely, substituting boehmite or corundum for gibbsite, in a high-alumina high-level-waste melter feed. The extent of foaming was determined using the volume expansion test and the conversion heat with differential scanning calorimetry. Evolved gas analysis was used to identify gases responsible for the formation of primary and secondary foam. The foam thickness, a critical factor in the rate of melting, was estimated using known values of heat conductivities and melting rates. The result was in reasonable agreement with the foam thickness experimentally observed in quenched cold caps from the laboratory-scale melter.

  11. CHARACTERIZATION OF A PRECIPITATE REACTOR FEED TANK (PRFT) SAMPLE FROM THE DEFENSE WASTE PROCESSING FACILITY (DWPF)

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, C.; Bannochie, C.

    2014-05-12

    A sample of from the Defense Waste Processing Facility (DWPF) Precipitate Reactor Feed Tank (PRFT) was pulled and sent to the Savannah River National Laboratory (SRNL) in June of 2013. The PRFT in DWPF receives Actinide Removal Process (ARP)/ Monosodium Titanate (MST) material from the 512-S Facility via the 511-S Facility. This 2.2 L sample was to be used in small-scale DWPF chemical process cell testing in the Shielded Cells Facility of SRNL. A 1L sub-sample portion was characterized to determine the physical properties such as weight percent solids, density, particle size distribution and crystalline phase identification. Further chemical analysis of the PRFT filtrate and dissolved slurry included metals and anions as well as carbon and base analysis. This technical report describes the characterization and analysis of the PRFT sample from DWPF. At SRNL, the 2.2 L PRFT sample was composited from eleven separate samples received from DWPF. The visible solids were observed to be relatively quick settling which allowed for the rinsing of the original shipping vials with PRFT supernate on the same day as compositing. Most analyses were performed in triplicate except for particle size distribution (PSD), X-ray diffraction (XRD), Scanning Electron Microscopy (SEM) and thermogravimetric analysis (TGA). PRFT slurry samples were dissolved using a mixed HNO3/HF acid for subsequent Inductively Coupled Plasma Atomic Emission Spectroscopy (ICPAES) and Inductively Coupled Plasma Mass Spectroscopy (ICP-MS) analyses performed by SRNL Analytical Development (AD). Per the task request for this work, analysis of the PRFT slurry and filtrate for metals, anions, carbon and base were primarily performed to support the planned chemical process cell testing and to provide additional component concentrations in addition to the limited data available from DWPF. Analysis of the insoluble solids portion of the PRFT slurry was aimed at detailed characterization of these solids (TGA, PSD

  12. A pilot scale demonstration of the DWPF process control and product verification strategy

    International Nuclear Information System (INIS)

    Hutson, N.D.; Jantzen, C.M.; Beam, D.C.

    1992-01-01

    The Defense Waste Processing Facility (DWPF) has been designed and constructed to immobilize Savannah River Site high level liquid waste within a durable borosilicate glass matrix for permanent storage. The DWPF will be operated to produce a glass product which must meet a number of product property constraints which are dependent upon the final product composition. During actual operations, the DWPF will control the properties of the glass product by the controlled blending of the waste streams with a glass-forming frit to produce the final melter feed slurry. The DWPF will verify control of the glass product through analysis of vitrified samples of slurry material. In order to demonstrate the DWPF process control and product verification strategy, a pilot-scale vitrification research facility was operated in three discrete batches using simulated DWPF waste streams. All of the DWPF process control methodologies were followed and the glass produce from each experiment was leached according to the Product Consistency Test. Results of the campaign are summarized

  13. A pilot scale demonstration of the DWPF process control and product verification strategy

    International Nuclear Information System (INIS)

    Hutson, N.D.; Jantzen, C.M.; Beam, D.C.

    1992-01-01

    The Defense Waste Processing Facility (DWPF) has been designed and constructed to immobilize Savannah River Site high level liquid waste within a durable borosilicate glass matrix for permanent storage. The DWPF will be operated to produce a glass product which must meet a number of product property constraints which are dependent upon the final product composition. During actual operations, the DWPF will control the properties of the glass product by the controlled blending of the waste streams with a glass-forming frit to produce the final melter feed slurry. The DWPF will verify control of the glass product through analysis of vitrified samples of slurry material. In order to demonstrate the DWPF process control and product verification strategy, a pilot-scale vitrification research facility was operated in three discrete batches using simulated DWPF waste streams. All of the DWPF process control methodologies were followed and the glass product from each experiment was leached according to the Product Consistency Test. In this paper results of the campaign are summarized

  14. Melting characteristics of a plasma torch melter according to the waste feeding method

    International Nuclear Information System (INIS)

    Kim, T. W.; Choi, J. R.; Park, S. C.; Lu, C. S.; Park, J. K.; Hwang, T. W.; Shin, S. W.

    2001-01-01

    By using a batch type plasma torch melting system, continuous feeding and melting tests of non-combustible waste were executed. Using the results, the establishment of a heat transfer model and its verification were executed; the characteristics of the molten slag, exhaust gas, fly dust, volatilization of Cs, and leaching of slag were analyzed. In order to establish the heat transfer mode, the followings were considered; the electrical energy supplied to the plasma torch, the absorbed energy to the plasma torch for generating the plasma gas, the absorbed energy to the cooling water of the plasma torch, the energy supplied to the melter from the plasma gas by radiant heat, the energy loss through the exhaust gas, the waste melting energy, and the heating energy of an inner crucible and the melter. The concrete and soil were melted for the verification of the model. The waste was fed through waste feeder by the amount of 0.5kg or 1kg that was calculated by using the model. The experiment for the verification resulted in that the model was fitted well until the melter was heated sufficiently. If the electrical energy of 128kW were supplied to the plasma torch, energy balance of the plasma melting system was calculated with the model: the absorbed energy to the plasma torch for generating the plasma gas (27kW), the absorbed energy to the cooling water of the plasma torch (0∼ 36kW), the energy loss through the exhaust gas (5 ∼ 8kW), the waste melting energy (14kW), and the heating energy of an inner crucible and the melter (82 ∼ 43kW)

  15. LFCM [liquid-fed eramic melter] emission and off-gas system performance for feed component cesium

    International Nuclear Information System (INIS)

    Goles, R.W.; Andersen, C.M.

    1986-09-01

    Except for volatile off-gas effluents, overall adequacy of the liquid-fed ceramic melter (LFCM) system depends most upon its effectiveness in dealing with cesium. However, the mechanism responsible for melter cesium losses has proved insensitive to many LFCM operating and processing conditions. As a result, variations in inleakage, plenum temperature, feeding rate and waste loading do not significantly influence melter cesium performance. Feed composition, specifically halogen content, is the only processing variable that has had a significant effect. Due to the submicron nature of LFCM-generated aerosols, melter disengagement design features are not expected to be particularly effective in reducing cesium emission rates. For the same reason, the cesium performance of conventional quench scrubbers is quite low, being dependent only upon the magnitude of melter entrainment losses. Although a deep bed washable filter has been effective in removing submicron aerosols from the process exhaust, high performance has only been achieved under dry operating conditions. The melter's idling state does not appear to place additional demands upon the off-gas treatment system

  16. Numerical modeling of liquid feeding in the liquid-fed ceramic melter

    International Nuclear Information System (INIS)

    Hjelm, R.L.; Donovan, T.E.

    1979-10-01

    A modeling scheme developed by the Pacific Northwest Laboratory numerically simulates the behavior of the Liquid-Fed Ceramic Melter (LFCM) during liquid feeding. The computer code VECTRA (Vorticity Energy Code for TRansport Analysis) was used to simulate the LFCM in the idling and liquid feeding modes. Results for each simulation include molten glass temperature profiles and isotherm contour plots, stream function contour plots, heat generation rate contour plots, refractory isotherms, and heat balances. The results indicated that the model showed no major deviations from real LFCM behavior and that high throughput should be attainable. They also indicated that reboil was a possibility as a steady liquid feeding state was approached, very steep temperature gradients exist in the Monofrax K-3, and that phase separation could occur in the bottom corners during liquid feeding and over the entire floor while idling

  17. Final Report - Engineering Study for DWPF Bubblers, VSL-10R1770-1, Rev. 0, dated 12/22/10

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Joseph, I.; Matlack, K. S.; Kot, W. K.; Diener, G. A.; Pegg, I. L.; Callow, R. A.

    2013-11-13

    The objective of this work was to perform an engineering assessment of the impact of implementation of bubblers to improve mixing of the glass pool, and thereby increase throughput, in the Defense Waste Processing Facility (DWPF) on the melter and off-gas system. Most of the data used for this evaluation were from extensive melter tests performed on non-SRS feeds. This information was supplemented by more recent results on SRS HLW simulants that were tested on a melter system at VSL under contracts from ORP and SRR. Per the work scope, the evaluation focused on the following areas: Glass production rate; Corrosion of melter components; Power requirements; Pouring stability; Off-gas characteristics; Safety and flammability.

  18. Relaxation of the lower frit loading constraint for DWPF process control

    International Nuclear Information System (INIS)

    Brown, K.G.

    2000-01-01

    The lower limit on the frit loading parameter when measurement uncertainty is introduced has impacted DWPF performance during immobilization of Tank 42 Sludge; therefore, any defensible relaxation or omission of this constraint should correspondingly increase DWPF waste loading and efficiency. Waste loading should be increased because the addition of frit is the current remedy for exceeding the lower frit loading constraint. For example, frit was added to DWPF SME Batches 94, 97 and 98 to remedy these batches for low frit loading. Attempts were also made to add frit in addition to the optimum computed to assure the lower frit loading constraint would be satisfied; however, approximately half of the SME Batches produced after Batch 98 have violated the lower frit loading constraint. If the DWPF batches did not have to be remediated and additional frit added because of the lower frit loading limit, then both, the performance of the DWPF process and the waste loading in the glass produced would be increased. Before determining whether or not the lower frit loading limit can be relaxed or omitted, the origin of this and the other constraints related to durability prediction must be examined. The lower limit loading constraint results from the need to make highly durable glass in DWPF. It is required that DWPF demonstrate that the glass produced would have durability that is at least two standard deviations greater than that of the Environmental Assessment (EA) glass. Glass durability cannot be measured in situ, it must be predicted from composition which can be measured. Fortunately, the leaching characteristics of homogeneous waste glasses is strongly related to the total molar free energy of the constituent species. Thus the waste acceptance specification has been translated into a requirement that the total molar free energy associated with the glass composition that would be produced from a DWPF melter feed batch be less than that of the EA glass accounting for

  19. Analysis of high-level radioactive slurries as a method to reduce DWPF turnaround times

    International Nuclear Information System (INIS)

    Coleman, C.J.; Bibler, N.E.; Ferrara, D.M.; Hay, M.S.

    1996-01-01

    Analysis of Defense Waste Processing Facility (DWPF) samples as slurries rather than as dried or vitrified samples is an effective way to reduce sample turnaround times. Slurries can be dissolved with a mixture of concentrated acids to yield solutions for elemental analysis by inductively coupled plasma-atomic emission spectroscopy (ICP-AES). Slurry analyses can be performed in eight hours, whereas analyses of vitrified samples require up to 40 hours to complete. Analyses of melter feed samples consisting of the DWPF borosilicate frit and either simulated or actual DWPF radioactive sludge were typically within a range of 3--5% of the predicted value based on the relative amounts of sludge and frit added to the slurry. The results indicate that the slurry analysis approach yields analytical accuracy and precision competitive with those obtained from analyses of vitrified samples. Slurry analyses offer a viable alternative to analyses of solid samples as a simple way to reduce analytical turnaround times

  20. Fabrication of remote steam atomized scrubbers for DWPF off-gas system

    International Nuclear Information System (INIS)

    Nielsen, M.G.; Lafferty, J.D.

    1988-01-01

    The defense waste processing facility (DWPF) is being constructed for the purpose of processing high-level waste from sludge to a vitrified borosilicate glass. In the operation of continuous slurry-fed melters, off-gas aerosols are created by entrainment of feed slurries and the vaporization of volatile species from the molten glass mixture. It is necessary to decontaminate these aerosols in order to minimize discharge of airborne radionuclide particulates. A steam atomized scrubber (SAS) has been developed for DWPF which utilizes a patented hydro- sonic system gas scrubbing method. The Hydro-Sonic System utilizes a steam aspirating-type venturi scrubber that requires very precise fabrication tolerances in order to obtain acceptable decontamination factors. In addition to the process-related tolerances, precision mounting and nozzle tolerances are required for remote service at DWPF

  1. DWPF Sample Vial Insert Study-Statistical Analysis of DWPF Mock-Up Test Data

    Energy Technology Data Exchange (ETDEWEB)

    Harris, S.P. [Westinghouse Savannah River Company, AIKEN, SC (United States)

    1997-09-18

    This report is prepared as part of Technical/QA Task Plan WSRC-RP-97-351 which was issued in response to Technical Task Request HLW/DWPF/TTR-970132 submitted by DWPF. Presented in this report is a statistical analysis of DWPF Mock-up test data for evaluation of two new analytical methods which use insert samples from the existing HydragardTM sampler. The first is a new hydrofluoric acid based method called the Cold Chemical Method (Cold Chem) and the second is a modified fusion method.Either new DWPF analytical method could result in a two to three fold improvement in sample analysis time.Both new methods use the existing HydragardTM sampler to collect a smaller insert sample from the process sampling system. The insert testing methodology applies to the DWPF Slurry Mix Evaporator (SME) and the Melter Feed Tank (MFT) samples.The insert sample is named after the initial trials which placed the container inside the sample (peanut) vials. Samples in small 3 ml containers (Inserts) are analyzed by either the cold chemical method or a modified fusion method. The current analytical method uses a HydragardTM sample station to obtain nearly full 15 ml peanut vials. The samples are prepared by a multi-step process for Inductively Coupled Plasma (ICP) analysis by drying, vitrification, grinding and finally dissolution by either mixed acid or fusion. In contrast, the insert sample is placed directly in the dissolution vessel, thus eliminating the drying, vitrification and grinding operations for the Cold chem method. Although the modified fusion still requires drying and calcine conversion, the process is rapid due to the decreased sample size and that no vitrification step is required.A slurry feed simulant material was acquired from the TNX pilot facility from the test run designated as PX-7.The Mock-up test data were gathered on the basis of a statistical design presented in SRT-SCS-97004 (Rev. 0). Simulant PX-7 samples were taken in the DWPF Analytical Cell Mock

  2. Production and remediation of low-sludge, simulated Purex waste glasses, 1: Effects of sludge oxide additions on melter operation

    International Nuclear Information System (INIS)

    Ramsey, W.G.

    1993-01-01

    Glass produced during the Purex 4 campaigns of the Integrated Defense Waste Processing Facility (DWPF) Melter System (IDMS) and the 774 Research Melter contained a lower fraction of sludge components than targeted by the Product Composition Control System (PCCS). Purex 4 glass was more durable than the benchmark (EA) glass, but less durable than most simulated SRS high-level waste glasses. Also, Purex 4 glass was considerably less durable than predicted by the algorithm which will be used to control production of DWPF glass. A melter run was performed using the 774 Research Melter to determine if the initial PCCS target composition determined for Purex 4 would produce acceptable glass whose durability could be accurately modeled by Hydration Thermodynamics. Reagent grade oxides and carbonates were added to Purex 4 melter feed stock to simulate a higher sludge loading. Each canister of glass produced was sampled and the composition, crystallinity, and durability was determined. This document details the melter operation and composition and crystallinity analyses

  3. Sampling data summary for the ninth run of the Large Slurry Fed Melter

    International Nuclear Information System (INIS)

    Sabatino, D.M.

    1983-01-01

    The ninth experimental run of the Large Slurry Fed Melter (LSFM) was completed June 27, 1983, after 63 days of continuous operation. During the run, the various melter and off-gas streams were sampled and analyzed to determine melter material balances and to characterize off-gas emissions. Sampling methods and preliminary results were reported earlier. The emphasis was on the chemical analyses of the off-gas entrainment, deposits, and scrubber liquid. The significant sampling results from the run are summarized below: Flushing the Frit 165 with Frit 131 without bubbler agitation required 3 to 4.5 melter volumes. The off-gas cesium concentration during feeding was on the order of 36 to 56 μgCs/scf. The cesium concentration in the melter plenum (based on air in leakage only) was on the order of 110 to 210 μgCs/scf. Using <1 micron as the cut point for semivolatile material 60% of the chloride, 35% of the sodium and less than 5% of the managanese and iron in the entrainment are present as semivolatiles. A material balance on the scrubber tank solids shows good agreement with entrainment data. An overall cesium balance using LSFM-9 data and the DWPF production rate indicates an emission of 0.11 mCi/yr of cesium from the DWPF off-gas. This is a factor of 27 less than the maximum allowable 3 mCi/yr

  4. Development of the high-level waste high-temperature melter feed preparation flowsheet for vitrification process testing

    International Nuclear Information System (INIS)

    Seymour, R.G.

    1995-01-01

    High-level waste (HLW) feed preparation flowsheet development was initiated in fiscal year (FY) 1994 to evaluate alternative flowsheets for preparing melter feed for high-temperature melter (HTM) vitrification testing. Three flowsheets were proposed that might lead to increased processing capacity relative to the Hanford Waste Vitrification Plant (HWVP) and that were flexible enough to use with other HLW melter technologies. This document describes the decision path that led to the selection of flowsheets to be tested in the FY 1994 small-scale HTM tests. Feed preparation flowsheet development for the HLW HTM was based on the feed preparation flowsheet that was developed for the HWVP. This approach allowed the HLW program to build upon the extensive feed preparation flowsheet database developed under the HWVP Project. Primary adjustments to the HWVP flowsheet were to the acid adjustment and glass component additions. Developmental background regarding the individual features of the HLW feed preparation flowsheets is provided. Applicability of the HWVP flowsheet features to the new HLW vitrification mission is discussed. The proposed flowsheets were tested at the laboratory-scale at Pacific Northwest Laboratory. Based on the results of this testing and previously established criteria, a reductant-based flowsheet using glycolic acid and a nitric acid-based flowsheet were selected for the FY 1994 small-scale HTM testing

  5. Conditions for precipitation of copper phases in DWPF waste glass

    International Nuclear Information System (INIS)

    Schumacher, R.F.; Ramsey, W.G.

    1993-01-01

    The Defense Waste Processing Facility (DWPF) precipitate hydrolysis process requires the use of copper formate catalyst. The expected absorbed radiation doses to the precipitate require levels of copper formate that increase the potential for the precipitation of metallic copper in the DWPF Melter. The conditions required to avoid the precipitation of copper are described

  6. DWPF MATERIALS EVALUATION SUMMARY REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Gee, T.; Chandler, G.; Daugherty, W.; Imrich, K.; Jankins, C.

    1996-09-12

    To better ensure the reliability of the Defense Waste Processing Facility (DWPF) remote canyon process equipment, a materials evaluation program was performed as part of the overall startup test program. Specific test programs included FA-04 ('Process Vessels Erosion/Corrosion Studies') and FA-05 (melter inspection). At the conclusion of field testing, Test Results Reports were issued to cover the various test phases. While these reports completed the startup test requirements, DWPF-Engineering agreed to compile a more detailed report which would include essentially all of the materials testing programs performed at DWPF. The scope of the materials evaouation programs included selected equipment from the Salt Process Cell (SPC), Chemical Process Cell (CPC), Melt Cell, Canister Decon Cell (CDC), and supporting facilities. The program consisted of performing pre-service baseline inspections (work completed in 1992) and follow-up inspections after completion of the DWPF cold chemical runs. Process equipment inspected included: process vessels, pumps, agitators, coils, jumpers, and melter top head components. Various NDE (non-destructive examination) techniques were used during the inspection program, including: ultrasonic testing (UT), visual (direct or video probe), radiography, penetrant testing (PT), and dimensional analyses. Finally, coupon racks were placed in selected tanks in 1992 for subsequent removal and corrosion evaluation after chemical runs.

  7. Glass formulation requirements for DWPF coupled operations using crystalline silicotitanates

    International Nuclear Information System (INIS)

    Harbour, J.R.; Andrews, M.K.

    1997-01-01

    The design basis DWPF flowsheet couples the vitrification of two waste streams: (1) a washed sludge and (2) a hydrolyzed sodium tetraphenylborate precipitate product, PHA. The PHA contains cesium-137 which had been precipitated from the tank supernate with sodium tetraphenylborate. Smaller amounts of strontium and plutonium adsorbed on sodium titanate are also present with the PHA feed. Currently, DWPF is running a sludge-only flowsheet while working towards solutions to the problems encountered with In Tank Precipitation (ITP). The sludge loading for the sludge-only flowsheet and for the anticipated coupled operations is 28 wt% on an oxide basis. For the coupled operation, it is essential to balance the treatment of the two waste streams such that no supernate remains after immobilization of all the sludge. An alternative to ITP and sodium titanate is the removal of Cs-137, Sr-90, and plutonium from the tank supernate by ion exchange using crystalline silicotitanate (CST). This material has been shown to effectively sorb these elements from the supernate. It is also known that CST sorbs plutonium. The loaded CST could then be immobilized with the sludge during vitrification. It has recently been demonstrated that CST loadings approaching 70 wt% for a CST-only glass can be achieved using a borosilicate glass formulation which can be processed by the DWPF melter. Initial efforts on coupled waste streams with simulated DWPF sludge show promise that a borosilicate glass formulation can incorporate both sludge and CST. This paper presents the bases for research efforts to develop a glass formulation which will incorporate sludge and CST at loadings appropriate for DWPF operation

  8. Conversion of nuclear waste to molten glass: Formation of porous amorphous alumina in a high-Al melter feed

    Science.gov (United States)

    Xu, Kai; Hrma, Pavel; Washton, Nancy; Schweiger, Michael J.; Kruger, Albert A.

    2017-01-01

    The transition of Al phases in a simulated high-Al high-level nuclear waste melter feed heated at 5 K min-1 to 700 °C was investigated with transmission electron microscopy, 27Al nuclear magnetic resonance spectroscopy, the Brunauer-Emmett-Teller method, and X-ray diffraction. At temperatures between 300 and 500 °C, porous amorphous alumina formed from the dehydration of gibbsite, resulting in increased specific surface area of the feed (∼8 m2 g-1). The high-surface-area amorphous alumina formed in this manner could potentially stop salt migration in the cold cap during nuclear waste vitrification.

  9. DWPF Sample Vial Insert Study-Statistical Analysis of DWPF Mock-Up Test Data

    International Nuclear Information System (INIS)

    Harris, S.P.

    1997-01-01

    This report is prepared as part of Technical/QA Task Plan WSRC-RP-97-351 which was issued in response to Technical Task Request HLW/DWPF/TTR-970132 submitted by DWPF. Presented in this report is a statistical analysis of DWPF Mock-up test data for evaluation of two new analytical methods which use insert samples from the existing HydragardTM sampler. The first is a new hydrofluoric acid based method called the Cold Chemical Method (Cold Chem) and the second is a modified fusion method.Both new methods use the existing HydragardTM sampler to collect a smaller insert sample from the process sampling system. The insert testing methodology applies to the DWPF Slurry Mix Evaporator (SME) and the Melter Feed Tank (MFT) samples. Samples in small 3 ml containers (Inserts) are analyzed by either the cold chemical method or a modified fusion method. The current analytical method uses a HydragardTM sample station to obtain nearly full 15 ml peanut vials. The samples are prepared by a multi-step process for Inductively Coupled Plasma (ICP) analysis by drying, vitrification, grinding and finally dissolution by either mixed acid or fusion. In contrast, the insert sample is placed directly in the dissolution vessel, thus eliminating the drying, vitrification and grinding operations for the Cold chem method. Although the modified fusion still requires drying and calcine conversion, the process is rapid due to the decreased sample size and that no vitrification step is required.A slurry feed simulant material was acquired from the TNX pilot facility from the test run designated as PX-7.The Mock-up test data were gathered on the basis of a statistical design presented in SRT-SCS-97004 (Rev. 0). Simulant PX-7 samples were taken in the DWPF Analytical Cell Mock-up Facility using 3 ml inserts and 15 ml peanut vials. A number of the insert samples were analyzed by Cold Chem and compared with full peanut vial samples analyzed by the current methods. The remaining inserts were analyzed by

  10. DWPF coupled feed flowsheet material balance with batch one sludge and copper nitrate catalyst

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A.S.

    1993-09-28

    The SRTC has formally transmitted a recommendation to DWPF to replace copper formate with copper nitrate as the catalyst form during precipitate hydrolysis [1]. The SRTC was subsequently requested to formally document the technical bases for the recommendation. A memorandum was issued on August 23, 1993 detailing the activities (and responsible individuals) necessary to address the impact of this change in catalyst form on process compatibility, safety, processibility environmental impact and product glass quality [2]. One of the activities identified was the preparation of a material balance in which copper nitrate is substituted for copper formate and the identification of key comparisons between this material balance and the current Batch 1 sludge -- Late Wash material balance [3].

  11. Conversion of nuclear waste to molten glass: Formation of porous amorphous alumina in a high-Al melter feed

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Kai, E-mail: kaixu@whut.edu.cn [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Hrma, Pavel, E-mail: pavel.hrma@pnnl.gov [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Washton, Nancy; Schweiger, Michael J. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Kruger, Albert A. [U.S. Department of Energy, Office of River Protection, Richland, WA 99352 (United States)

    2017-01-15

    The transition of Al phases in a simulated high-Al high-level nuclear waste melter feed heated at 5 K min{sup −1} to 700 °C was investigated with transmission electron microscopy, {sup 27}Al nuclear magnetic resonance spectroscopy, the Brunauer-Emmett-Teller method, and X-ray diffraction. At temperatures between 300 and 500 °C, porous amorphous alumina formed from the dehydration of gibbsite, resulting in increased specific surface area of the feed (∼8 m{sup 2} g{sup −1}). The high-surface-area amorphous alumina formed in this manner could potentially stop salt migration in the cold cap during nuclear waste vitrification. - Highlights: • Porous amorphous alumina formed in a simulated high-Al HLW melter feed during heating. • The feed had a high specific surface area at 300 °C ≤ T ≤ 500 °C. • Porous amorphous alumina induced increased specific surface area.

  12. Final flush of the shielded cells melter

    International Nuclear Information System (INIS)

    Marshall, K.M.; Fellinger, T.L.; Harbour, J.R.

    1997-01-01

    A flush of the Savannah River Technology Center (SRTC) Shielded Cells melter was performed after the completion of a campaign to vitrify loaded crystalline silicotitanate (CST) ion exchange medium. The purpose of the flush was to lower levels of radioisotopes accumulated during the campaign and to lower the level of titanium dioxide present in the glass. This in turn would ready the melter for future campaigns involving the Defense Waste Processing Facility (DWPF)

  13. Nuclear safety of extended sludge processing on tank 42 and 51 sludge (DWPF sludge feed batch one)

    International Nuclear Information System (INIS)

    Clemons, J.S.

    1993-01-01

    The sludge in tanks 42 and 51 is to be washed with inhibited water to remove soluble salts and combined in tank 51 in preparation for feed to DWPF. Since these tanks contain uranium and plutonium, the process of washing must be evaluated to ensure subcriticality is maintained. When the sludge is washed, inhibited water is added, the tank contents are slurried and allowed to settle. The sludge wash water is then decanted to the evaporator feed tank where it is fed to the evaporator to reduce the volume. The resulting evaporator concentrate is sent to a salt tank where it cools and forms crystallized salt cake. This salt cake will later be dissolved, processed in ITP and sent to Z-Area. This report evaluates the supernate and sludge during washing, the impact on the evaporator during concentration of decanted wash water, and the salt tank where the concentrated supernate is deposited. The conclusions generated in this report are specific to the sludge currently contained in tanks 42 and 51

  14. Thermal effects of electrically conductive deposits in melter

    International Nuclear Information System (INIS)

    Choi, I.G.; Bickford, D.F.; Carter, J.T.

    1992-01-01

    The radioactive waste processed by the Defense Waste Processing Facility melter at the Savannah river Site contains noble metal fission-products. Operation of waste-glass melters treating commercial power reactor wastes indicates that accumulation of noble metals on melter floors can lead to distortion of electric heating patterns, loss of power, and possible electrode damage. Changes in melter geometry have been developed in Japan and Germany to minimize these effects. The two existing melters for the US Department of Energy's Defense Waste Processing Facility were designed in 1982, before this effect was known or had been characterized. Modeling and pilot scale tests are being conducted in the Integrated DWPF melter system to determine if the effect is significant for melters processing defense wastes, and if the effect can be diagnosed and corrected without significant damage or changes to the melter design. This document provides a discussion of these tests

  15. DWPF remotable television and cell lighting facilities

    International Nuclear Information System (INIS)

    Heckendorn, F.M. II.

    1984-01-01

    The Defense Waste Processing Facility (DWPF) for radioactive waste vitrification at the Savannah River Plant (SRP) is now under construction. Development of specialized low cost television (TV) viewing equipment for in-cell and within-melter applications is now complete. High resolution TV cameras not originally designed for high radiation environments have been demonstrated in crane remotable packages to be well suited to the DWPF. High intensity in-cell lighting has also been demonstrated in crane remotable assemblies. These dual 1000 W units (2000 W total) are used to support the multiplicity of TV and cell window viewing requirements. 8 figures

  16. Task technical plan: DWPF air permit/dispersion modeling

    International Nuclear Information System (INIS)

    Lambert, D.P.

    1993-01-01

    This Task Technical Plan summarizes work required to project the benzene emissions from the Late Wash Facility (LWF) as well as update the benzene, mercury, and NO x emissions from the remainder of the Defense Waste Processing Facility (DWPF). These calculations will reflect (1) the addition of the LWF and (2) the replacement of formic acid with nitric acid in the melter preparation process. The completed calculations will be used to assist DWPF in applying for the LWF Air Quality Permit

  17. Defense Waste Processing Facility (DWPF) Viscosity Model: Revisions for Processing High TiO2 Containing Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-30

    Radioactive high-level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the Defense Waste Processing Facility (DWPF) since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it is poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to guarantee, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository. The DWPF SPC system is known as the Product Composition Control System (PCCS). The DWPF will soon be receiving wastes from the Salt Waste Processing Facility (SWPF) containing increased concentrations of TiO2, Na2O, and Cs2O . The SWPF is being built to pretreat the high-curie fraction of the salt waste to be removed from the HLW tanks in the F- and H-Area Tank Farms at the SRS. In order to process TiO2 concentrations >2.0 wt% in the DWPF, new viscosity data were developed over the range of 1.90 to 6.09 wt% TiO2 and evaluated against the 2005 viscosity model. An alternate viscosity model is also derived for potential future use, should the DWPF ever need to process other titanate-containing ion exchange materials. The ultimate limit on the amount of TiO2 that can be accommodated from SWPF will be determined by the three PCCS models, the waste composition of a given sludge

  18. Melter viewing system for liquid-fed ceramic melters

    International Nuclear Information System (INIS)

    Westsik, J.H. Jr.; Brenden, B.B.

    1988-01-01

    Melter viewing systems are an integral component of the monitoring and control systems for liquid-fed ceramic melters. The Pacific Northwest Laboratory (PNL) has designed cameras for use with glass melters at PNL, the Hanford Waste Vitrification Plant (HWVP), and West Valley Demonstration Project (WVDP). This report is a compilation of these designs. Operating experiences with one camera designed for the PNL melter are discussed. A camera has been fabricated and tested on the High-Bay Ceramic Melter (HBCM) and the Pilot-Scale Ceramic Melter (PSCM) at PNL. The camera proved to be an effective tool for monitoring the cold cap formed as the feed pool developed on the molten glass surface and for observing the physical condition of the melter. Originally, the camera was built to operate using the visible light spectrum in the melter. It was later modified to operate using the infrared (ir) spectrum. In either configuration, the picture quality decreases as the size of the cold cap increases. Large cold caps cover the molten glass, reducing the amount of visible light and reducing the plenum temperatures below 600 0 C. This temperature corresponds to the lowest level of blackbody radiation to which the video tube is sensitive. The camera has been tested in melter environments for about 1900 h. The camera has withstood mechanical shocks and vibrations. The cooling system in the camera has proved effective in maintaining the optical and electronic components within acceptable temperature ranges. 10 refs., 15 figs

  19. Impact Of Melter Internal Design On Off-Gas Flammability

    International Nuclear Information System (INIS)

    Choi, A. S.; Lee, S. Y.

    2012-01-01

    The purpose of this study was to: (1) identify the more dominant design parameters that can serve as the quantitative measure of how prototypic a given melter is, (2) run the existing DWPF models to simulate the data collected using both DWPF and non-DWPF melter configurations, (3) confirm the validity of the selected design parameters by determining if the agreement between the model predictions and data is reasonably good in light of the design and operating conditions employed in each data set, and (4) run Computational Fluid Dynamics (CFD) simulations to gain new insights into how fluid mixing is affected by the configuration of melter internals and to further apply the new insights to explaining, for example, why the agreement is not good

  20. The DWPF waste form qualification program

    International Nuclear Information System (INIS)

    Marra, S.L.; Plodinec, M.J.

    1994-01-01

    Prior to the introduction of radioactive feed into the Defense Waste Processing Facility for immobilization in borosilicate glass an extensive waste qualification program must be completed. The DWPF must demonstrate its ability to comply with the Waste Acceptance Product Specifications. This ability is being demonstrated through laboratory and pilot scale work and will be completed after the full operation of the DWPF using various simulated feeds

  1. Product/Process (P/P) Models For The Defense Waste Processing Facility (DWPF): Model Ranges And Validation Ranges For Future Processing

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-25

    Radioactive high level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the Defense Waste Processing Facility (DWPF) since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it is poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to guarantee, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository.

  2. Design of a mixing system for simulated high-level nuclear waste melter feed slurries

    International Nuclear Information System (INIS)

    Peterson, M.E.; McCarthy, D.; Muhlstein, K.D.

    1986-03-01

    The Nuclear Waste Treatment Program development program consists of coordinated nonradioactive and radioactive testing combined with numerical modeling of the process to provide a complete basis for design and operation of a vitrification facility. The radioactive demonstration tests of equipment and processes are conducted before incorporation in radioactive pilot-scale melter systems for final demonstration. The mixing system evaluation described in this report was conducted as part of the nonradioactive testing. The format of this report follows the sequence in which the design of a large-scale mixing system is determined. The initial program activity was concerned with gaining an understanding of the theoretical foundation of non-Newtonian mixing systems. Section 3 of this report describes the classical rheological models that are used to describe non-Newtonian mixing systems. Since the results obtained here are only valid for the slurries utilized, Section 4, Preparation of Simulated Hanford and West Valley Slurries, describes how the slurries were prepared. The laboratory-scale viscometric and physical property information is summarized in Section 5, Laboratory Rheological Evaluations. The bench-scale mixing evaluations conducted to define the effects of the independent variables described above on the degree of mixing achieved with each slurry are described in Section 6. Bench-scale results are scaled-up to establish engineering design requirements for the full-scale mixing system in Section 7. 24 refs., 37 figs., 44 tabs

  3. Material compatibility evaluataion for DWPF nitric-glycolic acid - literature review

    International Nuclear Information System (INIS)

    Mickalonis, J.I; Skidmore, T.E.

    2013-01-01

    species and to verify the performance of materials in the key process vessels as well as downstream vessels and processes such as the evaporator where heating is occurring. The following testing would provide data for establishing the viability of these components: Electrochemical testing - evaluate the corrosion rate and susceptibility to localized corrosion within the SRAT, SME, OGCT, Quencher and Evaporator. Testing would be conducted at operational temperatures in simulants with ranges of glycolic acid, iron, chloride, sulfate, mercury, and antifoaming agents; Hot-wall testing – evaluate the corrosion under heat transfer conditions to simulate those for heating coils and evaporator coil surfaces. Testing would be at nominal chemistries with concentration of glycolic acid, chloride, sulfate and mercury at high expected concentrations. Some tests would be performed with antifoaming agents; Melter coupon testing – evaluate the performance of alloy 690 in melter feeds containing glycolic acid. This testing would be conducted as part of the melter flammability testing; and, Polymer testing – evaluate changes in polymer properties in immersion testing with DWPF simulants to provide product-specific data for service life evaluation and analyze the Hansen solubility parameters for relevant polymers in glycolic vs. formic acid. During this literature review process, the difficulties associated with measuring the liquid level in formic acid tanks were revealed. A test is recommended to resolve this issue prior to the introduction of glycolic acid into the DWPF. This testing would evaluate the feasibility of using ultrasonic inspection techniques to determine liquid level and other desirable attributes of glycolic acid in DWPF storage tanks and related equipment.

  4. Interim glycol flowsheet reduction/oxidation (redox) model for the Defense Waste Processing Facility (DWPF)

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Zamecnik, J. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Missimer, D. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-08

    Control of the REDuction/OXidation (REDOX) state of glasses containing high concentrations of transition metals, such as High Level Waste (HLW) glasses, is critical in order to eliminate processing difficulties caused by overly reduced or overly oxidized melts. Operation of a HLW melter at Fe+2/ΣFe ratios of between 0.09 and 0.33, a range which is not overly oxidizing or overly reducing, helps retain radionuclides in the melt, i.e. long-lived radioactive 99Tc species in the less volatile reduced Tc4+ state, 104Ru in the melt as reduced Ru+4 state as insoluble RuO2, and hazardous volatile Cr6+ in the less soluble and less volatile Cr+3 state in the glass. The melter REDOX control balances the oxidants and reductants from the feed and from processing additives such as antifoam. Currently, the Defense Waste Processing Facility (DWPF) is running a formic acid-nitric acid (FN) flowsheet where formic acid is the main reductant and nitric acid is the main oxidant. During decomposition formate and formic acid releases H2 gas which requires close control of the melter vapor space flammability. A switch to a nitric acid-glycolic acid (GN) flowsheet is desired as the glycolic acid flowsheet releases considerably less H2 gas upon decomposition. This would greatly simplify DWPF processing. Development of an EE term for glycolic acid in the GN flowsheet is documented in this study.

  5. Characterization of high level nuclear waste glass samples following extended melter idling

    Energy Technology Data Exchange (ETDEWEB)

    Fox, Kevin M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Peeler, David K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kruger, Albert A. [USDOE Office of River Protection, Richland, WA (United States)

    2015-06-16

    The Savannah River Site Defense Waste Processing Facility (DWPF) melter was recently idled with glass remaining in the melt pool and riser for approximately three months. This situation presented a unique opportunity to collect and analyze glass samples since outages of this duration are uncommon. The objective of this study was to obtain insight into the potential for crystal formation in the glass resulting from an extended idling period. The results will be used to support development of a crystal-tolerant approach for operation of the high-level waste melter at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Two glass pour stream samples were collected from DWPF when the melter was restarted after idling for three months. The samples did not contain crystallization that was detectible by X-ray diffraction. Electron microscopy identified occasional spinel and noble metal crystals of no practical significance. Occasional platinum particles were observed by microscopy as an artifact of the sample collection method. Reduction/oxidation measurements showed that the pour stream glasses were fully oxidized, which was expected after the extended idling period. Chemical analysis of the pour stream glasses revealed slight differences in the concentrations of some oxides relative to analyses of the melter feed composition prior to the idling period. While these differences may be within the analytical error of the laboratories, the trends indicate that there may have been some amount of volatility associated with some of the glass components, and that there may have been interaction of the glass with the refractory components of the melter. These changes in composition, although small, can be attributed to the idling of the melter for an extended period. The changes in glass composition resulted in a 70-100 °C increase in the predicted spinel liquidus temperature (TL) for the pour stream glass samples relative to the analysis of the melter feed prior to

  6. Balance of oxygen throughout the conversion of a high-level waste melter feed to glass

    Czech Academy of Sciences Publication Activity Database

    Lee, S.M.; Hrma, P.; Kloužek, Jaroslav; Pokorný, R.; Hujová, Miroslava; Dixon, D.R.; Schweiger, M. J.; Kruger, A.A.

    2017-01-01

    Roč. 43, č. 16 (2017), s. 13113-13118 ISSN 0272-8842 Institutional support: RVO:67985891 Keywords : oxygen mass balance * feed-to-glass conversion * evolved gas * oxygen partial pressure * Fe redox ratio Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass OBOR OECD: Ceramics Impact factor: 2.986, year: 2016

  7. Bounding estimate of DWPF mercury emissions

    International Nuclear Information System (INIS)

    Jacobs, R.A.

    1992-01-01

    Purges required for H2 flammability control and verification of elevated Formic Acid Vent Condenser (FAVC) exit temperatures due to NO x reactions have lead to significant changes in Chemical Process Cell (CPC) operating conditions. Accordingly, mercury emissions estimates have been updated based upon the new operating requirements, IDMS (Integrated DWPF Melter System) experience, and development of an NO x /FAVC model which predicts FAVC exit temperatures. Using very conservative assumptions and maximum purge rates, the maximum calculated Hg emissions is approximately 130 lbs/yr. A range of 100 to 120 lbs/yr is conservatively predicted for other operating conditions. Defense Waste Processing Facility (DWPF) permitted Hg emissions are 175 lbs/yr (0.02 lbs/hr annual average)

  8. Melter Feed Reactions at T ≤ 700°C for Nuclear Waste Vitrification

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Kai [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hrma, Pavel R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rice, Jarrett A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Riley, Brian J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schweiger, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-07-23

    Batch reactions and phase transitions in a nuclear waste feed heated at 5 K min-1 up to 600°C were investigated by optical microscopy, scanning electron microscopy with energy dispersive X-ray spectrometer, and X-ray diffraction. Quenched samples were leached in deionized water at room temperature and 80°C to extract soluble salts and early glass-forming melt, respectively. To determine the content and composition of leachable phases, the leachates were analyzed by the inductively-coupled plasma spectroscopy. By ~400°C, gibbsite and borax lost water and converted to amorphous and intermediate crystalline phases. Between 400°C and 600°C, the sodium borate early glass-forming melt reacted with amorphous aluminum oxide and calcium oxide to form intermediate products containing Al and Ca. At ~600°C, half Na and B converted to the early glass-forming melt, and quartz began to dissolve in the melt.

  9. Evaluation Of The Impact Of The Defense Waste Processing Facility (DWPF) Laboratory Germanium Oxide Use On Recycle Transfers To The H-Tank Farm

    International Nuclear Information System (INIS)

    Jantzen, C.; Laurinat, J.

    2011-01-01

    When processing High Level Waste (HLW) glass, the Defense Waste Processing Facility (DWPF) cannot wait until the melt or waste glass has been made to assess its acceptability, since by then no further changes to the glass composition and acceptability are possible. Therefore, the acceptability decision is made on the upstream feed stream, rather than on the downstream melt or glass product. This strategy is known as 'feed forward statistical process control.' The DWPF depends on chemical analysis of the feed streams from the Sludge Receipt and Adjustment Tank (SRAT) and the Slurry Mix Evaporator (SME) where the frit plus adjusted sludge from the SRAT are mixed. The SME is the last vessel in which any chemical adjustments or frit additions can be made. Once the analyses of the SME product are deemed acceptable, the SME product is transferred to the Melter Feed Tank (MFT) and onto the melter. The SRAT and SME analyses have been analyzed by the DWPF laboratory using a 'Cold Chemical' method but this dissolution did not adequately dissolve all the elemental components. A new dissolution method which fuses the SRAT or SME product with cesium nitrate (CsNO 3 ), germanium (IV) oxide (GeO 2 ) and cesium carbonate (Cs 2 CO 3 ) into a cesium germanate glass at 1050 C in platinum crucibles has been developed. Once the germanium glass is formed in that fusion, it is readily dissolved by concentrated nitric acid (about 1M) to solubilize all the elements in the SRAT and/or SME product for elemental analysis. When the chemical analyses are completed the acidic cesium-germanate solution is transferred from the DWPF analytic laboratory to the Recycle Collection Tank (RCT) where the pH is increased to ∼12 M to be released back to the tank farm and the 2H evaporator. Therefore, about 2.5 kg/yr of GeO 2 /year will be diluted into 1.4 million gallons of recycle. This 2.5 kg/yr of GeO 2 may increase to 4 kg/yr when improvements are implemented to attain an annual canister production

  10. RECENT PROCESS IMPROVEMENTS TO INCREASE HLW THROUGHPUT AT THE DWPF

    International Nuclear Information System (INIS)

    Herman, C

    2007-01-01

    The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF), the world's largest operating high level waste (HLW) vitrification plant, began stabilizing about 35 million gallons of SRS liquid radioactive waste by-product in 1996. The DWPF has since filled over 2000 canisters with about 4000 pounds of radioactive glass in each canister. In the past few years there have been several process and equipment improvements at the DWPF to increase the rate at which the waste can be stabilized. These improvements have either directly increased waste processing rates or have desensitized the process and therefore minimized process upsets and thus downtime. These improvements, which include glass former optimization, increased waste loading of the glass, the melter heated bellows liner, and glass surge protection software, will be discussed in this paper

  11. DWPF GLASS BEADS AND GLASS FRIT TRANSPORT DEMONSTRATION

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, D; Bradley Pickenheim, B

    2008-11-24

    DWPF is considering replacing irregularly shaped glass frit with spherical glass beads in the Slurry Mix Evaporator (SME) process to decrease the yield stress of the melter feed (a non-Newtonian Bingham Plastic). Pilot-scale testing was conducted on spherical glass beads and glass frit to determine how well the glass beads would transfer when compared to the glass frit. Process Engineering Development designed and constructed the test apparatus to aid in the understanding and impacts that spherical glass beads may have on the existing DWPF Frit Transfer System. Testing was conducted to determine if the lines would plug with the glass beads and the glass frit slurry and what is required to unplug the lines. The flow loop consisted of vertical and horizontal runs of clear PVC piping, similar in geometry to the existing system. Two different batches of glass slurry were tested: a batch of 50 wt% spherical glass beads and a batch of 50 wt% glass frit in process water. No chemicals such as formic acid was used in slurry, only water and glass formers. The glass beads used for this testing were commercially available borosilicate glass of mesh size -100+200. The glass frit was Frit 418 obtained from DWPF and is nominally -45+200 mesh. The spherical glass beads did not have a negative impact on the frit transfer system. The transferring of the spherical glass beads was much easier than the glass frit. It was difficult to create a plug with glass bead slurry in the pilot transfer system. When a small plug occurred from setting overnight with the spherical glass beads, the plug was easy to displace using only the pump. In the case of creating a man made plug in a vertical line, by filling the line with spherical glass beads and allowing the slurry to settle for days, the plug was easy to remove by using flush water. The glass frit proved to be much more difficult to transfer when compared to the spherical glass beads. The glass frit impacted the transfer system to the point

  12. DWPF GLASS BEADS AND GLASS FRIT TRANSPORT DEMONSTRATION

    International Nuclear Information System (INIS)

    Adamson, D.; Pickenheim, Bradley

    2008-01-01

    DWPF is considering replacing irregularly shaped glass frit with spherical glass beads in the Slurry Mix Evaporator (SME) process to decrease the yield stress of the melter feed (a non-Newtonian Bingham Plastic). Pilot-scale testing was conducted on spherical glass beads and glass frit to determine how well the glass beads would transfer when compared to the glass frit. Process Engineering Development designed and constructed the test apparatus to aid in the understanding and impacts that spherical glass beads may have on the existing DWPF Frit Transfer System. Testing was conducted to determine if the lines would plug with the glass beads and the glass frit slurry and what is required to unplug the lines. The flow loop consisted of vertical and horizontal runs of clear PVC piping, similar in geometry to the existing system. Two different batches of glass slurry were tested: a batch of 50 wt% spherical glass beads and a batch of 50 wt% glass frit in process water. No chemicals such as formic acid was used in slurry, only water and glass formers. The glass beads used for this testing were commercially available borosilicate glass of mesh size -100+200. The glass frit was Frit 418 obtained from DWPF and is nominally -45+200 mesh. The spherical glass beads did not have a negative impact on the frit transfer system. The transferring of the spherical glass beads was much easier than the glass frit. It was difficult to create a plug with glass bead slurry in the pilot transfer system. When a small plug occurred from setting overnight with the spherical glass beads, the plug was easy to displace using only the pump. In the case of creating a man made plug in a vertical line, by filling the line with spherical glass beads and allowing the slurry to settle for days, the plug was easy to remove by using flush water. The glass frit proved to be much more difficult to transfer when compared to the spherical glass beads. The glass frit impacted the transfer system to the point

  13. Research-scale melter test report

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, M.F.; Elliott, M.L.; Eyler, L.L.; Freeman, C.J.; Higginson, J.J.; Mahoney, L.A.; Powell, M.R.

    1994-05-01

    The Melter Performance Assessment (MPA) activity in the Pacific Northwest Laboratory`s (PNL) Hanford Waste Vitrification Plant (HWVP) Technology Development (PHTD) effort is intended to determine the impact of noble metals on the operational life of the reference HWVP melter. As a part of this activity, a parametric melter test was completed using a Research-Scale Melter (RSM). The RSM is a small, approximately 1/100-scale melter, 6-in.-diameter, that allows rapid changing of process conditions and subsequent re-establishment of a steady-state condition. The test matrix contained nine different segments that varied the melter operating parameters (glass and plenum temperatures) and feed properties (oxide concentration, redox potential, and noble metal concentrations) so that the effects of these parameters on noble metal agglomeration on the melter floor could be evaluated. The RSM operated for 48 days and consumed 1,300 L of feed, equating to 153 tank turnovers. The run produced 531 kg of glass. During the latter portion of the run, the resistance between the electrodes decreased. Upon destructive examination of the melter, a layer of noble metals was found on the bottom. This was surprising because the glass residence time in the RSM is only 10% of the HWVP plant melter. The noble metals layer impacted the melter significantly. Approximately 1/3 of one paddle electrode was melted or corroded off. The cause is assumed to be localized heating from short circuiting of the electrode to the noble metal layer. The metal layer also removed approximately 1/2 in. of the refractory on the bottom of the melter. The mechanism for this damage is not presently known.

  14. Research-scale melter test report

    International Nuclear Information System (INIS)

    Cooper, M.F.; Elliott, M.L.; Eyler, L.L.; Freeman, C.J.; Higginson, J.J.; Mahoney, L.A.; Powell, M.R.

    1994-05-01

    The Melter Performance Assessment (MPA) activity in the Pacific Northwest Laboratory's (PNL) Hanford Waste Vitrification Plant (HWVP) Technology Development (PHTD) effort is intended to determine the impact of noble metals on the operational life of the reference HWVP melter. As a part of this activity, a parametric melter test was completed using a Research-Scale Melter (RSM). The RSM is a small, approximately 1/100-scale melter, 6-in.-diameter, that allows rapid changing of process conditions and subsequent re-establishment of a steady-state condition. The test matrix contained nine different segments that varied the melter operating parameters (glass and plenum temperatures) and feed properties (oxide concentration, redox potential, and noble metal concentrations) so that the effects of these parameters on noble metal agglomeration on the melter floor could be evaluated. The RSM operated for 48 days and consumed 1,300 L of feed, equating to 153 tank turnovers. The run produced 531 kg of glass. During the latter portion of the run, the resistance between the electrodes decreased. Upon destructive examination of the melter, a layer of noble metals was found on the bottom. This was surprising because the glass residence time in the RSM is only 10% of the HWVP plant melter. The noble metals layer impacted the melter significantly. Approximately 1/3 of one paddle electrode was melted or corroded off. The cause is assumed to be localized heating from short circuiting of the electrode to the noble metal layer. The metal layer also removed approximately 1/2 in. of the refractory on the bottom of the melter. The mechanism for this damage is not presently known

  15. Effects of Quartz Particle Size and Sucrose Addition on Melting Behavior of a Melter Feed for High-Level Waste Glass

    International Nuclear Information System (INIS)

    Marcial, Jose; Hrma, Pavel R.; Schweiger, Michael J.; Swearingen, Kevin J.; Tegrotenhuis, Nathan E.; Henager, Samuel H.

    2010-01-01

    The behavior of melter feed (a mixture of nuclear waste and glass-forming additives) during waste-glass processing has a significant impact on the rate of the vitrification process. We studied the effects of silica particle size and sucrose addition on the volumetric expansion (foaming) of a high-alumina feed and the rate of dissolution of silica particles in feed samples heated at 5 C/min up to 1200 C. The initial size of quartz particles in feed ranged from 5 to 195 (micro)m. The fraction of the sucrose added ranged from 0 to 0.20 g per g glass. Extensive foaming occurred only in feeds with 5-(micro)m quartz particles; particles (ge) 150 (micro)m formed clusters. Particles of 5 (micro)m completely dissolved by 900 C whereas particles (ge) 150 (micro)m did not fully dissolve even when the temperature reached 1200 C. Sucrose addition had virtually zero impact on both foaming and the dissolution of silica particles.

  16. DWPF Recycle Evaporator Simulant Tests

    International Nuclear Information System (INIS)

    Stone, M

    2005-01-01

    Testing was performed to determine the feasibility and processing characteristics of an evaporation process to reduce the volume of the recycle stream from the Defense Waste Processing Facility (DWPF). The concentrated recycle would be returned to DWPF while the overhead condensate would be transferred to the Effluent Treatment Plant. Various blends of evaporator feed were tested using simulants developed from characterization of actual recycle streams from DWPF and input from DWPF-Engineering. The simulated feed was evaporated in laboratory scale apparatus to target a 30X volume reduction. Condensate and concentrate samples from each run were analyzed and the process characteristics (foaming, scaling, etc) were visually monitored during each run. The following conclusions were made from the testing: Concentration of the ''typical'' recycle stream in DWPF by 30X was feasible. The addition of DWTT recycle streams to the typical recycle stream raises the solids content of the evaporator feed considerably and lowers the amount of concentration that can be achieved. Foaming was noted during all evaporation tests and must be addressed prior to operation of the full-scale evaporator. Tests were conducted that identified Dow Corning 2210 as an antifoam candidate that warrants further evaluation. The condensate has the potential to exceed the ETP WAC for mercury, silicon, and TOC. Controlling the amount of equipment decontamination recycle in the evaporator blend would help meet the TOC limits. The evaporator condensate will be saturated with mercury and elemental mercury will collect in the evaporator condensate collection vessel. No scaling on heating surfaces was noted during the tests, but splatter onto the walls of the evaporation vessels led to a buildup of solids. These solids were difficult to remove with 2M nitric acid. Precipitation of solids was not noted during the testing. Some of the aluminum present in the recycle streams was converted from gibbsite to

  17. Analysis Of DWPF Sludge Batch 7A (Macrobatch 8) Pour Stream Samples

    International Nuclear Information System (INIS)

    Johnson, F.

    2012-01-01

    The Defense Waste Processing Facility (DWPF) began processing Sludge Batch 7a (SB7a), also referred to as Macrobatch 8 (MB8), in June 2011. SB7a is a blend of the heel of Tank 40 from Sludge Batch 6 (SB6) and the SB7a material that was transferred to Tank 40 from Tank 51. SB7a was processed using Frit 418. During processing of each sludge batch, the DWPF is required to take at least one glass sample to meet the objectives of the Glass Product Control Program (GPCP), which is governed by the DWPF Waste Compliance Plan, and to complete the necessary Production Records so that the final glass product may be disposed of at a Federal Repository. Three pour stream glass samples and two Melter Feed Tank (MFT) slurry samples were collected while processing SB7a. These additional samples were taken during SB7a to understand the impact of antifoam and the melter bubblers on glass redox chemistry. The samples were transferred to the Savannah River National Laboratory (SRNL) where they were analyzed. The following conclusions were drawn from the analytical results provided in this report: (1) The sum of oxides for the official SB7a pour stream glass is within the Product Composition Control System (PCCS) limits (95-105 wt%). (2) The average calculated Waste Dilution Factor (WDF) for SB7a is 2.3. In general, the measured radionuclide content of the official SB7a pour stream glass is in good agreement with the calculated values from the Tank 40 dried sludge results from the SB7a Waste Acceptance Program Specification (WAPS) sample. (3) As in previous pour stream samples, ruthenium and rhodium inclusions were detected by Scanning Electron Microscopy-Electron Dispersive Spectroscopy (SEM-EDS) in the official SB7a pour stream sample. (4) The Product Consistency Test (PCT) results indicate that the official SB7a pour stream glass meets the waste acceptance criteria for durability with a normalized boron release of 0.64 g/L, which is an order of magnitude less than the Environmental

  18. Crystallization In High Level Waste (HLW) Glass Melters: Operational Experience From The Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-02-27

    processing strategy for the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The basis of this alternative approach is an empirical model predicting the crystal accumulation in the WTP glass discharge riser and melter bottom as a function of glass composition, time, and temperature. When coupled with an associated operating limit (e.g., the maximum tolerable thickness of an accumulated layer of crystals), this model could then be integrated into the process control algorithms to formulate crystal tolerant high level waste (HLW) glasses targeting higher waste loadings while still meeting process related limits and melter lifetime expectancies. This report provides a review of the scaled melter testing that was completed in support of the Defense Waste Processing Facility (DWPF) melter. Testing with scaled melters provided the data to define the DWPF operating limits to avoid bulk (volume) crystallization in the un-agitated DWPF melter and provided the data to distinguish between spinels generated by K-3 refractory corrosion versus spinels that precipitated from the HLW glass melt pool. This report includes a review of the crystallization observed with the scaled melters and the full scale DWPF melters (DWPF Melter 1 and DWPF Melter 2). Examples of actual DWPF melter attainment with Melter 2 are given. The intent is to provide an overview of lessons learned, including some example data, that can be used to advance the development and implementation of an empirical model and operating limit for crystal accumulation for WTP. Operation of the first and second (current) DWPF melters has demonstrated that the strategy of using a liquidus temperature predictive model combined with a 100 °C offset from the normal melter operating temperature of 1150 °C (i.e., the predicted liquidus temperature (TL) of the glass must be 1050 °C or less) has been successful in preventing any detrimental accumulation of spinel in the DWPF melt pool, and spinel has not been

  19. Remote viewing of melter interior Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Heckendorn, F.M. II.

    1986-01-01

    A remote system has been developed and demonstrated for continuous reviewing of the interior of a glass melter, which is used to vitrify highly radioactive waste. The system is currently being implemented with the Defense Waste Processing Facility (DWPF) now under construction at the Savannah River Plant (SRP). The environment in which the borescope/TV unit is implemented combines high temperature, high ionizing radiation, low light, spattering, deposition, and remote maintenance

  20. FEASIBILITY EVALUATION AND RETROFIT PLAN FOR COLD CRUCIBLE INDUCTION MELTER DEPLOYMENT IN THE DEFENSE WASTE PROCESSING FACILITY AT SAVANNAH RIVER SITE 8118

    International Nuclear Information System (INIS)

    Barnes, A; Dan Iverson, D; Brannen Adkins, B

    2008-01-01

    increase heat transfer to the slurry fed High Level Waste (HLW) sludge, the CCIM may be equipped with bubblers and/or water cooled mechanical agitators. The DWPF could benefit from use of CCIM technology, especially in light of our latest projections of waste volume to be vitrified. Increased waste loading and increased throughput could result in substantial life cycle cost reduction. In order to significantly surpass the waste throughput capability of the currently installed JHM, it may be necessary to install two 950 mm CCIMs in the DWPF Melt Cell. A cursory evaluation of system design requirements and modifications to the facility that may be required to support installation and operation of two 950 mm CCIMs was performed. Based on this evaluation, it appears technically feasible to position two CCIMs in the Melt Cell of the DWPF within the existing footprint of the current melter. Interfaces with support systems and controls including Melter Feed, Power, Melter Cooling Water, Melter Off-gas, and Canister Operations must be designed to support dual CCIM operations. This paper describes the CCIM technology and identifies technical challenges that must be addressed in order to implement CCIMs in the DWPF

  1. DWPF simulant CPC studies for SB8

    Energy Technology Data Exchange (ETDEWEB)

    Koopman, D. C.; Zamecnik, J. R.

    2013-06-25

    recommendations for DWPF along with some data related to Safety Class documentation at DWPF. Some significant observations regarding SB8 follow: Reduced washing in Tank 51 led to an increase in the wt.% soluble solids of the DWPF feed. If wt.% total solids for the SRAT and SME product weren’t adjusted upward to maintain insoluble solids levels similar to past sludge batches, then the rheological properties of the slurry went below the low end of the DWPF design bases for the SRAT and SME. Much higher levels of dissolved manganese were found in the SRAT and SME products than in recent sludge batches. Closed crucible melts were more reduced than expected. The working hypothesis is that the soluble Mn is less oxidizing than assumed in the REDOX calculations. A change in the coefficient for Mn in the REDOX equation was recommended in a separate report. The DWPF (Hsu) stoichiometric acid equation was examined in detail to better evaluate how to control acid in DWPF. The existing DWPF equation can likely be improved without changing the required sample analyses through a paper study using existing data. The recommended acid stoichiometry for initial SB8 SRAT batches is 115-120% stoichiometry until some processing experience is gained. The conservative range (based on feed properties) of stoichiometric factors derived in this study was from 110-147%, but SRNL recommends using only the lower half of this range, 110-126% even after initial batches provide processing experience. The stoichiometric range for sludge-only processing appears to be suitable for coupled operation based on results from the run in the middle of the range. Catalytic hydrogen was detectable (>0.005 vol%) in all SRAT and SME cycles. Hydrogen reached 30-35% of the SRAT and SME limits at the mid-point of the stoichiometry window (bounding noble metals and acid demand).

  2. DWPF simulant CPC studies for SB8

    International Nuclear Information System (INIS)

    Koopman, D. C.; Zamecnik, J. R.

    2013-01-01

    recommendations for DWPF along with some data related to Safety Class documentation at DWPF. Some significant observations regarding SB8 follow: Reduced washing in Tank 51 led to an increase in the wt.% soluble solids of the DWPF feed. If wt.% total solids for the SRAT and SME product weren't adjusted upward to maintain insoluble solids levels similar to past sludge batches, then the rheological properties of the slurry went below the low end of the DWPF design bases for the SRAT and SME. Much higher levels of dissolved manganese were found in the SRAT and SME products than in recent sludge batches. Closed crucible melts were more reduced than expected. The working hypothesis is that the soluble Mn is less oxidizing than assumed in the REDOX calculations. A change in the coefficient for Mn in the REDOX equation was recommended in a separate report. The DWPF (Hsu) stoichiometric acid equation was examined in detail to better evaluate how to control acid in DWPF. The existing DWPF equation can likely be improved without changing the required sample analyses through a paper study using existing data. The recommended acid stoichiometry for initial SB8 SRAT batches is 115-120% stoichiometry until some processing experience is gained. The conservative range (based on feed properties) of stoichiometric factors derived in this study was from 110-147%, but SRNL recommends using only the lower half of this range, 110-126% even after initial batches provide processing experience. The stoichiometric range for sludge-only processing appears to be suitable for coupled operation based on results from the run in the middle of the range. Catalytic hydrogen was detectable (<0.005 vol%) in all SRAT and SME cycles. Hydrogen reached 30-35% of the SRAT and SME limits at the mid-point of the stoichiometry window (bounding noble metals and acid demand)

  3. Organics Characterization Of DWPF Alternative Reductant Simulants, Glycolic Acid, And Antifoam 747

    Energy Technology Data Exchange (ETDEWEB)

    White, T. L. [Savannah River Site (SRS), Aiken, SC (United States); Wiedenman, B. J. [Savannah River Site (SRS), Aiken, SC (United States); Lambert, D. P. [Savannah River Site (SRS), Aiken, SC (United States); Crump, S. L. [Savannah River Site (SRS), Aiken, SC (United States); Fondeur, F. F. [Savannah River Site (SRS), Aiken, SC (United States); Papathanassiu, A. E. [Catholic University of America Vitreous State Laboratory, Washington, DC (United States); Kot, W. K. [Catholic University of America Vitreous State Laboratory, Washington, DC (United States); Pegg, I. L. [Catholic University of America Vitreous State Laboratory, Washington, DC (United States)

    2013-10-01

    The present study examines the fate of glycolic acid and other organics added in the Chemical Processing Cell (CPC) of the Defense Waste Processing Facility (DWPF) as part of the glycolic alternate flowsheet. Adoption of this flowsheet is expected to provide certain benefits in terms of a reduction in the processing time, a decrease in hydrogen generation, simplification of chemical storage and handling issues, and an improvement in the processing characteristics of the waste stream including an increase in the amount of nitrate allowed in the CPC process. Understanding the fate of organics in this flowsheet is imperative because tank farm waste processed in the CPC is eventually immobilized by vitrification; thus, the type and amount of organics present in the melter feed may affect optimal melt processing and the quality of the final glass product as well as alter flammability calculations on the DWPF melter off gas. To evaluate the fate of the organic compounds added as the part of the glycolic flowsheet, mainly glycolic acid and antifoam 747, samples of simulated waste that was processed using the DWPF CPC protocol for tank farm sludge feed were generated and analyzed for organic compounds using a variety of analytical techniques at the Savannah River National Laboratory (SRNL). These techniques included Ion Chromatography (IC), Gas Chromatography-Mass Spectrometry (GC-MS), Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES), and Nuclear Magnetic Resonance (NMR) Spectroscopy. A set of samples were also sent to the Catholic University of America Vitreous State Laboratory (VSL) for analysis by NMR Spectroscopy at the University of Maryland, College Park. Analytical methods developed and executed at SRNL collectively showed that glycolic acid was the most prevalent organic compound in the supernatants of Slurry Mix Evaporator (SME) products examined. Furthermore, the studies suggested that commercially available glycolic acid contained minor amounts

  4. Organics Characterization Of DWPF Alternative Reductant Simulants, Glycolic Acid, And Antifoam 747

    International Nuclear Information System (INIS)

    White, T. L.; Wiedenman, B. J.; Lambert, D. P.; Crump, S. L.; Fondeur, F. F.; Papathanassiu, A. E.; Kot, W. K.; Pegg, I. L.

    2013-01-01

    The present study examines the fate of glycolic acid and other organics added in the Chemical Processing Cell (CPC) of the Defense Waste Processing Facility (DWPF) as part of the glycolic alternate flowsheet. Adoption of this flowsheet is expected to provide certain benefits in terms of a reduction in the processing time, a decrease in hydrogen generation, simplification of chemical storage and handling issues, and an improvement in the processing characteristics of the waste stream including an increase in the amount of nitrate allowed in the CPC process. Understanding the fate of organics in this flowsheet is imperative because tank farm waste processed in the CPC is eventually immobilized by vitrification; thus, the type and amount of organics present in the melter feed may affect optimal melt processing and the quality of the final glass product as well as alter flammability calculations on the DWPF melter off gas. To evaluate the fate of the organic compounds added as the part of the glycolic flowsheet, mainly glycolic acid and antifoam 747, samples of simulated waste that was processed using the DWPF CPC protocol for tank farm sludge feed were generated and analyzed for organic compounds using a variety of analytical techniques at the Savannah River National Laboratory (SRNL). These techniques included Ion Chromatography (IC), Gas Chromatography-Mass Spectrometry (GC-MS), Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES), and Nuclear Magnetic Resonance (NMR) Spectroscopy. A set of samples were also sent to the Catholic University of America Vitreous State Laboratory (VSL) for analysis by NMR Spectroscopy at the University of Maryland, College Park. Analytical methods developed and executed at SRNL collectively showed that glycolic acid was the most prevalent organic compound in the supernatants of Slurry Mix Evaporator (SME) products examined. Furthermore, the studies suggested that commercially available glycolic acid contained minor amounts

  5. Bounding estimate of DWPF mercury emissions

    International Nuclear Information System (INIS)

    Jacobs, R.A.

    1993-01-01

    Two factors which have substantial impact on predicted Mercury emissions are the air flows in the Chemical Process Cell (CPC) and the exit temperature of the Formic Acid Vent Condenser (FAVC). The discovery in the IDMS (Integrated DWPF Melter System) of H 2 generation by noble metal catalyzed formic acid decomposition and the resultant required dilution air flow has increased the expected instantaneous CPC air flow by as much as a factor of four. In addition, IDMS has experienced higher than design (10 degrees C) FAVC exit temperatures during certain portions of the operating cycle. These temperatures were subsequently attributed to the exothermic reaction of NO to NO 2 . Moreover, evaluation of the DWPF FAVC indicated it was undersized and unless modified or replaced, routine exit temperatures would be in excess of design. Purges required for H 2 flammability control and verification of elevated FAVC exit temperatures due to NO x reactions have lead to significant changes in CPC operating conditions. Accordingly, mercury emissions estimates have been updated based upon the new operating requirements, IDMS experience, and development of an NO x /FAVC model which predicts FAVC exit temperatures. Using very conservative assumptions and maximum purge rates, the maximum calculated Hg emissions is approximately 130 lbs/yr. A range of 100 to 120 lbs/yr is conservatively predicted for other operating conditions. The peak emission rate calculated is 0.027 lbs/hr. The estimated DWPF Hg emissions for the construction permit are 175 lbs/yr (0.02 lbs/hr annual average)

  6. Minimum TI4085D interlock setpoint at 1.0 GPM sludge-only feed rate and 14,000 ppm TOC

    International Nuclear Information System (INIS)

    Choi, A.S.

    1996-01-01

    DWPF-Engineering requested that SRTC determine the minimum indicated melter vapor space temperature that must be maintained in order to minimize the potential for off-gas flammability during a steady sludge-only feeding operation at 1.0 GPM containing 14,000 ppm total organic carbon. The detailed scope of this request is described in the technical task request, HLW-DWPF-TTR-960092 (DWPT Activity No. DWPT-96-0065). In response to this request, a dynamic simulation study was conducted in which the concentration of flammable gases was tracked throughout the course of a simulated 3X off-gas surge using the melter off-gas (MOG) dynamics model. The results of simulation showed that as long as the melter vapor space temperature as indicated on TI4085D is kept at 570 degrees C or higher, the peak concentration of combustible gases in the melter off-gas system is not likely to exceed 60 percent of the lower flammability limit (LFL). The minimum TI4085D of 570 degrees C is valid only when the air purges to FIC3221A and FIC3221B are maintained at or above 850 and 250 lb/hr, respectively. All the key bases and assumptions along with the input data used in the simulation are described in the attached E-7 calculation note

  7. TTP SR1-6-WT-31, Milestone C.3-2 Annual Report on Clemson/INEEL Melter Work

    Energy Technology Data Exchange (ETDEWEB)

    Bickford, D.F.

    1999-10-20

    This work is performed in collaboration with RL37WT31-C and ID77WT31-B. During the first two years of radioactive operation of the DWPF process, several areas for improvement in melter design have been identified. The continuing scope of this task is to address performance limitations and deficiencies identified by the user. SRS will design and test several configurations of the melter pour spout and associated equipment to improve consistency of performance and recommend design improvements.

  8. TTP SR1-6-WT-31, Milestone C.3-2 Annual Report on Clemson/INEEL Melter Work

    International Nuclear Information System (INIS)

    Bickford, D.F.

    1999-01-01

    This work is performed in collaboration with RL37WT31-C and ID77WT31-B. During the first two years of radioactive operation of the DWPF process, several areas for improvement in melter design have been identified. The continuing scope of this task is to address performance limitations and deficiencies identified by the user. SRS will design and test several configurations of the melter pour spout and associated equipment to improve consistency of performance and recommend design improvements

  9. RECENT PROCESS AND EQUIPMENT IMPROVEMENTS TO INCREASE HIGH LEVEL WASTE THROUGHPUT AT THE DEFENSE WASTE PROCESSING FACILITY (DWPF)

    International Nuclear Information System (INIS)

    Smith, M; Allan Barnes, A; Jim Coleman, J; Robert Hopkins, R; Dan Iverson, D; Richard Odriscoll, R; David Peeler, D

    2006-01-01

    The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF), the world's largest operating high level waste (HLW) vitrification plant, began stabilizing about 35 million gallons of SRS liquid radioactive waste by-product in 1996. The DWPF has since filled over 2000 canisters with about 4000 pounds of radioactive glass in each canister. In the past few years there have been several process and equipment improvements at the DWPF to increase the rate at which the waste can be stabilized. These improvements have either directly increased waste processing rates or have desensitized the process and therefore minimized process upsets and thus downtime. These improvements, which include glass former optimization, increased waste loading of the glass, the melter glass pump, the melter heated bellows liner, and glass surge protection software, will be discussed in this paper

  10. Preliminary Evaluation Of DWPF Impacts Of Boric Acid Use In Cesium Strip FOR SWPF And MCU

    International Nuclear Information System (INIS)

    Stone, M.

    2010-01-01

    A new solvent system is being evaluated for use in the Modular Caustic-Side Solvent Extraction Unit (MCU) and in the Salt Waste Processing Facility (SWPF). The new system includes the option to replace the current dilute nitric acid strip solution with boric acid. To support this effort, the impact of using 0.01M, 0.1M, 0.25M and 0.5M boric acid in place of 0.001M nitric acid was evaluated for impacts on the DWPF facility. The evaluation only covered the impacts of boric acid in the strip effluent and does not address the other changes in solvents (i.e., the new extractant, called MaxCalix, or the new suppressor, guanidine). Boric acid additions may lead to increased hydrogen generation during the SRAT and SME cycles as well as change the rheological properties of the feed. The boron in the strip effluent will impact glass composition and could require each SME batch to be trimmed with boric acid to account for any changes in the boron from strip effluent additions. Addition of boron with the strip effluent will require changes in the frit composition and could lead to changes in melt behavior. The severity of the impacts from the boric acid additions is dependent on the amount of boric acid added by the strip effluent. The use of 0.1M or higher concentrations of boric acid in the strip effluent was found to significantly impact DWPF operations while the impact of 0.01M boric acid is expected to be relatively minor. Experimental testing is required to resolve the issues identified during the preliminary evaluation. The issues to be addressed by the testing are: (1) Impact on SRAT acid addition and hydrogen generation; (2) Impact on melter feed rheology; (3) Impact on glass composition control; (4) Impact on frit production; and (5) Impact on melter offgas. A new solvent system is being evaluated for use in the Modular Caustic-Side Solvent Extraction Unit (MCU) and in the Salt Waste Processing Facility (SWPF). The new system includes the option to replace the

  11. Melter Technologies Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Perez, J.M. Jr. [Pacific Northwest National Lab., Richland, WA (United States); Schumacher, R.F. [Savannah River Technology Center, Aiken, SC (United States); Forsberg, C.W. [Oak Ridge National Lab., TN (United States)

    1996-05-01

    The problem of controlling and disposing of surplus fissile material, in particular plutonium, is being addressed by the US Department of Energy (DOE). Immobilization of plutonium by vitrification has been identified as a promising solution. The Melter Evaluation Activity of DOE`s Plutonium Immobilization Task is responsible for evaluating and selecting the preferred melter technologies for vitrification for each of three immobilization options: Greenfield Facility, Adjunct Melter Facility, and Can-In-Canister. A significant number of melter technologies are available for evaluation as a result of vitrification research and development throughout the international communities for over 20 years. This paper describes an evaluation process which will establish the specific requirements of performance against which candidate melter technologies can be carefully evaluated. Melter technologies that have been identified are also described.

  12. Vitrification melter study

    International Nuclear Information System (INIS)

    Jones, J.A.

    1995-04-01

    This report presents the results of a study performed to identify the most promising vitrification melter technologies that the Department of Energy (EM-50) might pursue with available funding. The primary focus was on plasma arc systems and graphite arc melters. The study was also intended to assist EM-50 in evaluating competing technologies, formulating effective technology strategy, developing focused technology development projects, and directing the work of contractors involved in vitrification melter development

  13. Control of high level radioactive waste-glass melters

    International Nuclear Information System (INIS)

    Bickford, D.F.; Choi, A.S.

    1991-01-01

    Slurry Fed Melters (SFM) are being developed in the United States, Europe and Japan for the conversion of high-level radioactive waste to borosilicate glass for permanent disposal. The high transition metal, noble metal, nitrate, organic, and sulfate contents of these wastes lead to unique melter redox control requirements. Pilot waste-glass melter operations have indicated the possibility of nickel sulfide or noble-metal fission-product accumulation on melter floors, which can lead to distortion of electric heating patterns, and decrease melter life. Sulfide formation is prevented by control of the redox chemistry of the melter feed. The redox state of waste-glass melters is determined by balance between the reducing potential of organic compounds in the feed, and the oxidizing potential of gases above the melt, and nitrates and polyvalent elements in the waste. Semiquantitative models predicting limitations of organic content have been developed based on crucible testing. Computerized thermodynamic computations are being developed to predict the sequence and products of redox reactions and is assessing process variations. Continuous melter test results have been compared to improved computer staged-thermodynamic-models of redox behavior. Feed chemistry control to prevent sulfide and moderate noble metal accumulations are discussed. 17 refs., 3 figs

  14. Off-gas system data summary for the ninth run of the large slurry fed melter

    International Nuclear Information System (INIS)

    Colven, W.P.

    1983-01-01

    The ninth melter campaign successfully demonstrated extended operation of both melter and off-gas systems. Two critical problem areas associated with the handling of melter off-gases were resolved leading to firm definition of the DWPF Off-Gas Treatment System. These two concerns, wet scrubber decontamination efficiency and the reduction of solids deposition at the off-gas line entrance, were the primary focus of off-gas system studies during the 63-day run (LSFM-9). The Hydro-Sonic Scrubber was confirmed to be the superior candidate for wet scrubbing by outperforming all other scrubbers tested at the Equipment Test Facility (ETF). The two stage, steam-driven scrubber achieved consistent decontamination factors for cesium exceeding the required DWPF flowsheet DF of 50. As a result, the device was selected as the reference wet scrubber for the DWPF. The Off-Gas Film Cooling device continued to show promising results for reducing three accumulation of solid deposits at the entrance to the off-gas line. In addition, a rotating wire brush cleaning device provided easy and efficient removal of deposits which had accumulated. The combination of the two has adequately resolved the deposit accumulation problem and both devices have been incorporated in the DWPF design

  15. A summary report on feed preparation offgas and glass redox data for Hanford waste vitrification plant: Letter report

    International Nuclear Information System (INIS)

    Merz, M.D.

    1996-03-01

    Tests to evaluate feed processing options for the Hanford Waste Vitrification Plant (HWVP) were conducted by a number of investigators, and considerable data were acquired for tests of different scale, including recent full-scale tests. In this report, a comparison was made of the characteristics of feed preparation observed in tests of scale ranging from 57 ml to full-scale of 28,000 liters. These tests included Pacific Northwest Laboratory (PNL) laboratory-scale tests, Kernforschungszentrums Karlsruhe (KfK) melter feed preparation, Research Scale Melter (RSM) feed preparation, Integrated DWPF Melter System (IDMS) feed preparation, Slurry Integrated Performance Testing (SIPT) feed preparation, and formic acid addition to Hanford Neutralized Current Acid Waste (NCAW) care samples.' The data presented herein were drawn mainly from draft reports and include system characteristics such as slurry volume and depth, sweep gas flow rate, headspace, and heating and stirring characteristics. Operating conditions such as acid feed rate, temperature, starting pH, final pH, quantities and type of frit, nitrite, nitrate, and carbonate concentrations, noble metal content, and waste oxide loading were tabulated. Offgas data for CO 2 , NO x , N 2 O, NO 2 , H 2 and NH 3 were tabulated on a common basis. Observation and non-observation of other species were also noted

  16. Evaluation of liquid-fed ceramic melter scale-up correlations

    International Nuclear Information System (INIS)

    Koegler, S.S.; Mitchell, S.J.

    1988-08-01

    This study was conducted to determine the parameters governing factors of scale for liquid-fed ceramic melters (LFCMs) in order to design full-scale melters using smaller-scale melter data. Results of melter experiments conducted at Pacific Northwest Laboratory (PNL) and Savannah River Laboratory (SRL) are presented for two feed compositions and five different liquid-fed ceramic melters. The melter performance data including nominal feed rate and glass melt rate are correlated as a function of melter surface area. Comparisons are made between the actual melt rate data and melt rates predicted by a cold cap heat transfer model. The heat transfer model could be used in scale-up calculations, but insufficient data are available on the cold cap characteristics. Experiments specifically designed to determine heat transfer parameters are needed to further develop the model. 17 refs

  17. Final Report - Glass Formulation Development and Testing for DWPF High AI2O3 HLW Sludges, VSL-10R1670-1, Rev. 0, dated 12/20/10

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Pegg, I. L.; Kot, W. K.; Gan, H.; Matlack, K. S.

    2013-11-13

    The principal objective of the work described in this Final Report is to develop and identify glass frit compositions for a specified DWPF high-aluminum based sludge waste stream that maximizes waste loading while maintaining high production rate for the waste composition provided by ORP/SRS. This was accomplished through a combination of crucible-scale, vertical gradient furnace, and confirmation tests on the DM100 melter system. The DM100-BL unit was selected for these tests. The DM100-BL was used for previous tests on HLW glass compositions that were used to support subsequent tests on the HLW Pilot Melter. It was also used to process compositions with waste loadings limited by aluminum, bismuth, and chromium, to investigate the volatility of cesium and technetium during the vitrification of an HLW AZ-102 composition, to process glass formulations at compositional and property extremes, and to investigate crystal settling on a composition that exhibited one percent crystals at 963{degrees}C (i.e., close to the WTP limit). The same melter was selected for the present tests in order to maintain comparisons between the previously collected data. The tests provide information on melter processing characteristics and off-gas data, including formation of secondary phases and partitioning. Specific objectives for the melter tests are as follows: Determine maximum glass production rates without bubbling for a simulated SRS Sludge Batch 19 (SB19). Demonstrate a feed rate equivalent to 1125 kg/m{sup 2}/day glass production using melt pool bubbling. Process a high waste loading glass composition with the simulated SRS SB19 waste and measure the quality of the glass product. Determine the effect of argon as a bubbling gas on waste processing and the glass product including feed processing rate, glass redox, melter emissions, etc.. Determine differences in feed processing and glass characteristics for SRS SB19 waste simulated by the co-precipitated and direct

  18. Canister disposition plan for the DWPF Startup Test Program

    International Nuclear Information System (INIS)

    Harbour, J.R.; Payne, C.H.

    1990-01-01

    This report details the disposition of canisters and the canistered waste forms produced during the DWPF Startup Test Program. The six melter campaigns (DWPF Startup Tests FA-13, WP-14, WP-15, WP-16, WP-17, and FA-18) will produce 126 canistered waste forms. In addition, up to 20 additional canistered waste forms may be produced from glass poured during the transition between campaigns. In particular, this canister disposition plan (1) assigns (by alpha-numeric code) a specific canister to each location in the six campaign sequences, (2) describes the method of access for glass sampling on each canistered waste form, (3) describes the nature of the specific tests which will be carried out, (4) details which tests will be carried out on each canistered waste form, (5) provides the sequence of these tests for each canistered waste form, and (6) assigns a storage location for each canistered waste form. The tests are designed to provide evidence, as detailed in the Waste Form Compliance Plan (WCP 1 ), that the DWPF product will comply with the Waste Acceptance Product Specifications (WAPS 2 ). The WAPS must be met before the canistered waste form is accepted by DOE for ultimate disposal at the Federal Repository. The results of these tests will be included in the Waste Form Qualification Report (WQR)

  19. Literature review of arc/plasma, combustion, and joule-heated melter vitrification systems

    International Nuclear Information System (INIS)

    Freeman, C.J.; Abrigo, G.P.; Shafer, P.J.; Merrill, R.A.

    1995-07-01

    This report provides reviews of papers and reports for three basic categories of melters: arc/plasma-heated melters, combustion-heated melters, and joule-heated melters. The literature reviewed here represents those publications which may lend insight to phase I testing of low-level waste vitrification being performed at the Hanford Site in FY 1995. For each melter category, information from those papers and reports containing enough information to determine steady-state mass balance data is tabulated at the end of each section. The tables show the composition of the feed processed, the off-gas measured via decontamination factors, gross energy consumptions, and processing rates, among other data

  20. TTP SR1-6-WT-31, Milestone C.3-2 annual report on Clemson/INEEL melter work. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Bickford, D.F.

    1999-12-17

    This work is performed in collaboration with RL37WT31-C and ID77WT31-B. During the first two years of radioactive operation of the DWPF process, several areas for improvement in melter design have been identified. The continuing scope of this task is to address performance limitations and deficiencies identified by the user. SRS will design and test several configurations of the melter pour spout and associated equipment to improve consistency of performance and recommend design improvements.

  1. TTP SR1-6-WT-31, Milestone C.3-2 annual report on Clemson/INEEL melter work. Revision 1

    International Nuclear Information System (INIS)

    Bickford, D.F.

    1999-01-01

    This work is performed in collaboration with RL37WT31-C and ID77WT31-B. During the first two years of radioactive operation of the DWPF process, several areas for improvement in melter design have been identified. The continuing scope of this task is to address performance limitations and deficiencies identified by the user. SRS will design and test several configurations of the melter pour spout and associated equipment to improve consistency of performance and recommend design improvements

  2. DWPF process control

    International Nuclear Information System (INIS)

    Heckendoin, F.M. II

    1983-01-01

    The Defense Waste Processing Facility (DWPF) for waste vitrification at the Savannah River Plant (SRP) is in the final design stage. Instrumentation to provide the parameter sensing required to assure the quality of the two-foot-diameter, ten-foot-high waste canister is in the final stage of development. All step of the process and instrumentation are now operating as nearly full-scale prototypes at SRP. Quality will be maintained by assuring that only the intended material enters the canisters, and by sensing the resultant condition of the filled canisters. Primary emphasis will be on instrumentation of the process

  3. DC plasma arc melter technology for waste vitrification

    International Nuclear Information System (INIS)

    Hamilton, R.A.; Wittle, J.K.; Trescot, J.

    1995-01-01

    This paper describes the features and benefits of a breakthrough DC Arc Melter for the permanent treatment of all types of solid wastes including nonhazardous, hazardous and radioactive. This DC Arc Furnace system, now commercially available, is the low cost permanent solution for solid waste pollution prevention and remediation. Concern over the effective disposal of wastes generated by the industrial society, worldwide, has prompted development of technologies to address the problem. For the most part these technologies have resulted in niche solutions with limited application. The only solution that has the ability to process almost all wastes, and to recover/recycle metallic and inorganic matter, is the group of technologies known as melters. Melters have distinct advantages over traditional technologies such as incineration because melters operate at higher temperatures, are relatively unaffected by changes in the waste stream, produce a vitrified stable product, and have the capability to recover/recycle slag, metals and gas. The system, DC Plasma Arc Melter, has the lowest capital, maintenance and operating cost of any melter technology because of its patented DC Plasma Arc with graphite electrode. DC Plasma Arc Melter systems are commercially available in sizes from 50 kg/batch or 250--3,000 kg/hr on a continuous feed basis. This paper examines the design and operating benefits of a DC Plasma Arc Melter System

  4. The DWPF: Results of full scale qualification runs leading to radioactive operations

    International Nuclear Information System (INIS)

    Marra, S.L.; Elder, H.H.; Occhipinti, J.H.; Snyder, D.E.

    1996-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site in Aiken, SC will immobilize high-level radioactive liquid waste, currently stored in underground carbon steel tanks, in borosilicate glass. The radioactive waste is transferred to the DWPF in two forms: precipitate slurry and sludge slurry. The radioactive waste is pretreated and then combined with a borosilicate glass frit in the DWPF. This homogeneous slurry is fed to a Joule-heated melter which operates at approximately 1150 degrees C. The glass is poured into stainless steel canisters for eventual disposal in a geologic repository. The DWPF product (i.e. the canistered waste form) must comply with the Waste Acceptance Product Specifications (WAPS) in order to be acceptable for disposal. The DWPF has completed Waste Qualification Runs which demonstrate the facility's ability to comply with the waste acceptance specifications. During the Waste Qualification Runs seventy-one canisters of simulated waste glass were produced in preparation for Radioactive Operations. These canisters of simulated waste glass were produced during five production campaigns which also exercised the facility prior to beginning Radioactive Operations. The results of the Waste Qualification Runs are presented

  5. Efficient particulate scrubber for glass melter off-gas

    International Nuclear Information System (INIS)

    Wright, G.T.

    1983-01-01

    Operation of joule-heated, continuous slurry-fed melters has demonstrated that off-gas aerosols are generated by entrainment of feed slurry and vaporization of volatile species from the melt. Effective off-gas stream decontamination for these aerosols can be obtained by utilizing a suitably designed and operated wet scrubber system. Results are presented for performance tests conducted with an air aspirating-type venturi scrubber processing a simulated melter off-gas aerosol. Mass overall removal efficiencies ranged from 99.5 to 99.8%. Details of the testing program and applications for melter off-gas system design are discussed

  6. Liquid-fed ceramic melter: a general description report

    International Nuclear Information System (INIS)

    Buelt, J.L.; Chapman, C.C.

    1978-10-01

    The Pacific Northwest Laboratory is conducting several research and development programs for the solidification of high-level wastes. The liquid-fed ceramic melter (LFCM) is a major component in the solidification process. This melter can solidify liquid high-level waste, as well as melt calcined waste with glass additives and then solidify the mixture. This report describes the LFCM system and shows the main features of the refractories, electrodes and power systems, melter box and lid, draining system, feeding system, and off-gas system

  7. DWPF Development Plan. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Holtzscheiter, E.W.

    1994-05-09

    The DWPF Development Plan is based on an evaluation process flowsheet and related waste management systems. The scope is shown in Figure 1 entitled ``DWPF Process Development Systems.`` To identify the critical development efforts, each system has been analyzed to determine: The identification of unresolved technology issues. A technology issue (TI) is one that requires basic development to resolve a previously unknown process or equipment problem and is managed via the Technology Assurance Program co-chaired by DWPF and SRTC. Areas that require further work to sufficiently define the process basis or technical operating envelop for DWPF. This activity involves the application of sound engineering and development principles to define the scope of work required to complete the technical data. The identification of the level of effort and expertise required to provide process technical consultation during the start-up and demonstration of this first of a kind plant.

  8. DWPF Development Plan. Revision 1

    International Nuclear Information System (INIS)

    Holtzscheiter, E.W.

    1994-01-01

    The DWPF Development Plan is based on an evaluation process flowsheet and related waste management systems. The scope is shown in Figure 1 entitled ''DWPF Process Development Systems.'' To identify the critical development efforts, each system has been analyzed to determine: The identification of unresolved technology issues. A technology issue (TI) is one that requires basic development to resolve a previously unknown process or equipment problem and is managed via the Technology Assurance Program co-chaired by DWPF and SRTC. Areas that require further work to sufficiently define the process basis or technical operating envelop for DWPF. This activity involves the application of sound engineering and development principles to define the scope of work required to complete the technical data. The identification of the level of effort and expertise required to provide process technical consultation during the start-up and demonstration of this first of a kind plant

  9. Evaluation of melter technologies for vitrification of Hanford site low-level tank waste - phase 1 testing summary report

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, C.N., Westinghouse Hanford

    1996-06-27

    Following negotiation of the fourth amendment to the Tri- Party Agreement for Hanford Site cleanup, commercially available melter technologies were tested during 1994 and 1995 for vitrification of the low-level waste (LLW) stream to be derived from retrieval and pretreatment of the radioactive defense wastes stored in 177 underground tanks. Seven vendors were selected for Phase 1 testing to demonstrate vitrification of a high-sodium content liquid LLW simulant. The tested melter technologies included four Joule-heated melters, a carbon electrode melter, a combustion melter, and a plasma melter. Various dry and slurry melter feed preparation processes also were tested. The technologies and Phase 1 testing results were evaluated and a preliminary technology down-selection completed. This report describes the Phase 1 LLW melter vendor testing and the tested technologies, and summarizes the testing results and the preliminary technology recommendations.

  10. Laboratory Optimization Tests of Technetium Decontamination of Hanford Waste Treatment Plant Direct Feed Low Activity Waste Melter Off-Gas Condensate Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-12-23

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable less integrated operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste.

  11. Defense Waste Processing Facility (DWPF) Durability-Composition Models and the Applicability of the Associated Reduction of Constraints (ROC) Criteria for High TiO2 Containing Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Trivelpiece, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-30

    Radioactive high-level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the DWPF since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it has been poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than relying on statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to determine, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository. The DWPF SPC system is known as the Product Composition Control System (PCCS). One of the process models within PCCS is known as the Thermodynamic Hydration Energy Reaction MOdel (THERMO™). The DWPF will soon be receiving increased concentrations of TiO2-, Na2O-, and Cs2O-enriched wastes from the Salt Waste Processing Facility (SWPF). The SWPF has been built to pretreat the high-curie fraction of the salt waste to be removed from the HLW tanks in the F- and H-Area Tank Farms at the SRS. In order to validate the existing TiO2 term in THERMO™ beyond 2.0 wt% in the DWPF, new durability data were developed over the target range of 2.00 to 6.00 wt% TiO2 and evaluated against the 1995 durability model. The durability was measured by the 7-day Product Consistency Test. This study documents the adequacy of the existing THERMO™ terms. It is recommended that the modified THERMO™ durability models and

  12. Hanford Waste Vitrification Program process development: Melt testing subtask, pilot-scale ceramic melter experiment, run summary

    International Nuclear Information System (INIS)

    Nakaoka, R.K.; Bates, S.O.; Elmore, M.R.; Goles, R.W.; Perez, J.M.; Scott, P.A.; Westsik, J.H.

    1996-03-01

    Hanford Waste Vitrification Program (HWVP) activities for FY 1985 have included engineering and pilot-scale melter experiments HWVP-11/HBCM-85-1 and HWVP-12/PSCM-22. Major objectives designated by HWVP fo these tests were to evaluate the processing characteristics of the current HWVP melter feed during actual melter operation and establish the product quality of HW-39 borosilicate glass. The current melter feed, defined during FY 85, consists of reference feed (HWVP-RF) and glass-forming chemicals added as frit

  13. Integration of the Uncertainties of Anion and TOC Measurements into the Flammability Control Strategy for Sludge Batch 8 at the DWPF

    International Nuclear Information System (INIS)

    Edwards, T. B.

    2013-01-01

    The Savannah River National Laboratory (SRNL) has been working with the Savannah River Remediation (SRR) Defense Waste Processing Facility (DWPF) in the development and implementation of a flammability control strategy for DWPF's melter operation during the processing of Sludge Batch 8 (SB8). SRNL's support has been in response to technical task requests that have been made by SRR's Waste Solidification Engineering (WSE) organization. The flammability control strategy relies on measurements that are performed on Slurry Mix Evaporator (SME) samples by the DWPF Laboratory. Measurements of nitrate, oxalate, formate, and total organic carbon (TOC) standards generated by the DWPF Laboratory are presented in this report, and an evaluation of the uncertainties of these measurements is provided. The impact of the uncertainties of these measurements on DWPF's strategy for controlling melter flammability also is evaluated. The strategy includes monitoring each SME batch for its nitrate content and its TOC content relative to the nitrate content and relative to the antifoam additions made during the preparation of the SME batch. A linearized approach for monitoring the relationship between TOC and nitrate is developed, equations are provided that integrate the measurement uncertainties into the flammability control strategy, and sample calculations for these equations are shown to illustrate the impact of the uncertainties on the flammability control strategy

  14. Spray nozzle pattern test for the DWPF HEME Task QA Plan

    International Nuclear Information System (INIS)

    Lee, L.

    1991-01-01

    The DWPF melter off-gas systems have two High Efficiency Mist Eliminators (HEME) upstream of the High-Efficiency Particulates Air filters (HEPA) to remove fine mists and particulates from the off-gas. To have an acceptable filter life and an efficient operation, an air atomized water is spray on the HEME. The water spray keeps the HEME wet and dissolves the soluble particulates and enhances and HEME efficiency. DWPF Technical asked SRL to determine the conditions which will give satisfactory atomization and distribution of water so that the HEME will operate efficiently. The purpose of this document is to identify, QA controls to be applied in the pursuit of this task (WSRC-RP-91-1151)

  15. High level waste vitrification at the SRP [Savannah River Plant] (DWPF [Defense Waste Processing Facility] summary)

    International Nuclear Information System (INIS)

    Weisman, A.F.; Knight, J.R.; McIntosh, D.L.; Papouchado, L.M.

    1988-01-01

    The Savannah River Plant has been operating a nuclear fuel cycle since the early 1950's. Fuel and target elements are fabricated and irradiated to produce nuclear materials. After removal from the reactors, the fuel elements are processed to extract the products, and waste is stored. During the thirty years of operation including evaporation, about 30 million gallons of high level radioactive waste has accumulated. The Defense Waste Processing Facility (DWPF) under construction at Savannah River will process this waste into a borosilicate glass for long-term geologic disposal. The construction of the DWPF is about 70% complete; this paper will describe the status of the project, including design demonstrations, with an emphasis on the melter system. 9 figs

  16. Results of a pilot scale melter test to attain higher production rates

    International Nuclear Information System (INIS)

    Elliott, M.L.; Perez, J.M. Jr.; Chapman, C.C.

    1991-01-01

    A pilot-scale melter test was completed as part of the effort to enhance glass production rates. The experiment was designed to evaluate the effects of bulk glass temperature and feed oxide loading. The maximum glass production rate obtained, 86 kg/hr-m 2 , was over 200% better than the previous record for the melter used

  17. Vectra GSI, Inc. low-level waste melter testing Phase 1 test report

    Energy Technology Data Exchange (ETDEWEB)

    Stegen, G.E.; Wilson, C.N.

    1996-02-21

    A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Vectra GSI, Inc. was one of seven vendors selected for Phase 1 of the melter demonstration tests using simulated LLW that were completed during fiscal year 1995. The attached report prepared by Vectra GSI, Inc. describes results of melter testing using slurry feed and dried feeds. Results of feed drying and prereaction tests using a fluid bed calciner and rotary dryer also are described.

  18. Vectra GSI, Inc. low-level waste melter testing Phase 1 test report

    International Nuclear Information System (INIS)

    Stegen, G.E.; Wilson, C.N.

    1996-01-01

    A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Vectra GSI, Inc. was one of seven vendors selected for Phase 1 of the melter demonstration tests using simulated LLW that were completed during fiscal year 1995. The attached report prepared by Vectra GSI, Inc. describes results of melter testing using slurry feed and dried feeds. Results of feed drying and prereaction tests using a fluid bed calciner and rotary dryer also are described

  19. Lid heater for glass melter

    International Nuclear Information System (INIS)

    Phillips, T.D.

    1993-01-01

    A glass melter having a lid electrode for heating the glass melt radiantly. The electrode comprises a series of INCONEL 690 tubes running above the melt across the melter interior and through the melter walls and having nickel cores inside the tubes beginning where the tubes leave the melter interior and nickel connectors to connect the tubes electrically in series. An applied voltage causes the tubes to generate heat of electrical resistance for melting frit injected onto the melt. The cores limit heat generated as the current passes through the walls of the melter. Nickel bus connection to the electrical power supply minimizes heat transfer away from the melter that would occur if standard copper or water-cooled copper connections were used between the supply and the INCONEL 690 heating tubes. 3 figures

  20. Advanced Mixed Waste Treatment Project melter system preliminary design technical review meeting

    International Nuclear Information System (INIS)

    Eddy, T.L.; Raivo, B.D.; Soelberg, N.R.; Wiersholm, O.

    1995-02-01

    The Idaho National Engineering Laboratory Advanced Mixed Waste Treatment Project sponsored a plasma are melter technical design review meeting to evaluate high-temperature melter system configurations for processing heterogeneous alpha-contaminated low-level radioactive waste (ALLW). Thermal processing experts representing Department of Energy contractors, the Environmental Protection Agency, and private sector companies participated in the review. The participants discussed issues and evaluated alternative configurations for three areas of the melter system design: plasma torch melters and graphite arc melters, offgas treatment options, and overall system configuration considerations. The Technical Advisory Committee for the review concluded that graphite arc melters are preferred over plasma torch melters for processing ALLW. Initiating involvement of stakeholders was considered essential at this stage of the design. For the offgas treatment system, the advisory committee raised the question whether to a use wet-dry or a dry-wet system. The committee recommended that the waste stream characterization, feed preparation, and the control system are essential design tasks for the high-temperature melter treatment system. The participants strongly recommended that a complete melter treatment system be assembled to conduct tests with nonradioactive surrogate waste material. A nonradioactive test bed would allow for inexpensive design and operational changes prior to assembling a system for radioactive waste treatment operations

  1. Advanced Mixed Waste Treatment Project melter system preliminary design technical review meeting

    Energy Technology Data Exchange (ETDEWEB)

    Eddy, T.L.; Raivo, B.D.; Soelberg, N.R.; Wiersholm, O.

    1995-02-01

    The Idaho National Engineering Laboratory Advanced Mixed Waste Treatment Project sponsored a plasma are melter technical design review meeting to evaluate high-temperature melter system configurations for processing heterogeneous alpha-contaminated low-level radioactive waste (ALLW). Thermal processing experts representing Department of Energy contractors, the Environmental Protection Agency, and private sector companies participated in the review. The participants discussed issues and evaluated alternative configurations for three areas of the melter system design: plasma torch melters and graphite arc melters, offgas treatment options, and overall system configuration considerations. The Technical Advisory Committee for the review concluded that graphite arc melters are preferred over plasma torch melters for processing ALLW. Initiating involvement of stakeholders was considered essential at this stage of the design. For the offgas treatment system, the advisory committee raised the question whether to a use wet-dry or a dry-wet system. The committee recommended that the waste stream characterization, feed preparation, and the control system are essential design tasks for the high-temperature melter treatment system. The participants strongly recommended that a complete melter treatment system be assembled to conduct tests with nonradioactive surrogate waste material. A nonradioactive test bed would allow for inexpensive design and operational changes prior to assembling a system for radioactive waste treatment operations.

  2. DWPF recycle minimization: Brainstorming session

    International Nuclear Information System (INIS)

    Jacobs, R.A.; Poirier, M.R.

    1993-01-01

    The recycle stream from the DWPF constitutes a major source of water addition to the High Level Waste evaporator system. As now designed, the entire flow of 3.5 to 6.5 gal/min (at sign 25% and 75% attainment, respectively), or 2 gal/min during idling, flow to the 2H evaporator system (Tank 43). Substantial improvement in the HLW water balance and tank volume management is expected if the DWPF recycle to the HLW evaporator system can be significantly reduced. A task team has been appointed to study alternatives for reducing the flow to the HLW evaporator system and make recommendations for implementation and/or further study and evaluation. The brainstorming session detailed in this report was designed to produce the first cut options for the task team to further evaluate

  3. Statistical analysis of the DWPF prototypic sampler

    International Nuclear Information System (INIS)

    Postles, R.L.; Reeve, C.P.; Jenkins, W.J.; Bickford, D.F.

    1991-01-01

    The DWPF process will be controlled using assay measurements on samples of feed slurry. These slurries are radioactive, and thus will be sampled remotely. A Hydraguard trademark pump-driven sampler system will be used as the remote sampling device. A prototype Hydraguard trademark sampler has been studied in a full-scale mock-up of a DWPF process vessel. Two issues were of dominant interest: (1) what accuracy and precision can be provided by such a pump-driven sampler in the face of the slurry rheology; and, if the Hydraguard trademark sample accurately represents the slurry in its local area, (2) is the slurry homogeneous enough throughout for it to represent the entire vessel? To determine Hydraguard trademark Accuracy, a Grab Sampler of simpler mechanism was used as reference. This (Low) Grab Sampler was located as near to the intake port of the Hydraguard trademark as could be arranged. To determine Homogeneity, a second (High) Grab Sampler was located above the first. The data necessary to these determinations comes from the measurement system, so its important variables also affect the results. Thus, the design of the test involved not just Sampling variables, but also some of the Measurement variables as well. However, the main concern was the Sampler and not the Measurement System, so the test design included only such measurement variables as could not be circumvented (Vials, Dissolution Method, and Aliquoting). The test was executed by, or under the direct oversight of, expert technologists. It thus did not explore the many important particulars of ''routine'' plant operations (such as Remote Sample Preparation or Laboratory Shift Operation)

  4. INTEGRATED DM 1200 MELTER TESTING OF HLW C-106/AY-102 COMPOSITION USING BUBBLERS VSL-03R3800-1 REV 0 9/15/03

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of simulated HLW C-106/AY-102 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW C-106/AY-102 feed; determine the effect of bubbling rate on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post test inspections of system components.

  5. Integrated DM 1200 Melter Testing Of HLW C-106/AY-102 Composition Using Bubblers VSL-03R3800-1, Rev. 0, 9/15/03

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Bardakci, T.; Gong, W.; D'Angelo, N.A.; Pegg, I.L.

    2011-01-01

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of simulated HLW C-106/AY-102 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW C-106/AY-102 feed; determine the effect of bubbling rate on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post test inspections of system components.

  6. Corrosion impact of reductant on DWPF and downstream facilities

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Imrich, K. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Murphy, T. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Wilderman, J. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-12-01

    Glycolic acid is being evaluated as an alternate reductant in the preparation of high level waste for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). During processing, the glycolic acid is not completely consumed and small quantities of the glycolate anion are carried forward to other high level waste (HLW) facilities. The impact of the glycolate anion on the corrosion of the materials of construction throughout the waste processing system has not been previously evaluated. A literature review had revealed that corrosion data in glycolate-bearing solution applicable to SRS systems were not available. Therefore, testing was recommended to evaluate the materials of construction of vessels, piping and components within DWPF and downstream facilities. The testing, conducted in non-radioactive simulants, consisted of both accelerated tests (electrochemical and hot-wall) with coupons in laboratory vessels and prototypical tests with coupons immersed in scale-up and mock-up test systems. Eight waste or process streams were identified in which the glycolate anion might impact the performance of the materials of construction. These streams were 70% glycolic acid (DWPF feed vessels and piping), SRAT/SME supernate (Chemical Processing Cell (CPC) vessels and piping), DWPF acidic recycle (DWPF condenser and recycle tanks and piping), basic concentrated recycle (HLW tanks, evaporators, and transfer lines), salt processing (ARP, MCU, and Saltstone tanks and piping), boric acid (MCU separators), and dilute waste (HLW evaporator condensate tanks and transfer line and ETF components). For each stream, high temperature limits and worst-case glycolate concentrations were identified for performing the recommended tests. Test solution chemistries were generally based on analytical results of actual waste samples taken from the various process facilities or of prototypical simulants produced in the laboratory. The materials of construction for most vessels

  7. Vitrification of noble metals containing NCAW simulant with an engineering scale melter (ESM): Campaign report

    Energy Technology Data Exchange (ETDEWEB)

    Grunewald, W.; Roth, G.; Tobie, W.; Weisenburger, S.; Weiss, K.; Elliott, M.; Eyler, L.L.

    1996-03-01

    ESM has been designed as a 10th-scale model of the DWPF-type melter, currently the reference melter for nitrification of Hanford double shell tankwaste. ESM and related equipment have been integrated to the existing mockup vitrification plant VA-WAK at KfK. On June 2-July 10, 1992, a shakedown test using 2.61 m{sup 3} of NCAW (neutralized current acid waste) simulant without noble metals was performed. On July 11-Aug. 30, 1992, 14.23 m{sup 3} of the same simulant with nominal concentrations of Ru, Rh, and Pd were vitrified. Objective was to investigate the behavior of such a melter with respect to discharge of noble metals with routine glass pouring via glass overflow. Results indicate an accumulation of noble metals in the bottom area of the flat-bottomed ESM. About 65 wt% of the noble metals fed to the melter could be drained out, whereas 35 wt% accumulated in the melter, based on analysis of glass samples from glass pouring stream in to the canisters. After the melter was drained at the end of the campaign through a bottom drain valve, glass samples were taken from the residual bottom layer. The samples had significantly increased noble metals content (factor of 20-45 to target loading). They showed also a significant decrease of the specific electric resistance compared to bulk glass (factor of 10). A decrease of 10- 15% of the resistance between he power electrodes could be seen at the run end, but the total amount of noble metals accumulated was not yet sufficient enough to disturb the Joule heating of the glass tank severely.

  8. Induction melter apparatus

    Science.gov (United States)

    Roach, Jay A [Idaho Falls, ID; Richardson, John G [Idaho Falls, ID; Raivo, Brian D [Idaho Falls, ID; Soelberg, Nicholas R [Idaho Falls, ID

    2008-06-17

    Apparatus and methods of operation are provided for a cold-crucible-induction melter for vitrifying waste wherein a single induction power supply may be used to effect a selected thermal distribution by independently energizing at least two inductors. Also, a bottom drain assembly may be heated by an inductor and may include an electrically resistive heater. The bottom drain assembly may be cooled to solidify molten material passing therethrough to prevent discharge of molten material therefrom. Configurations are provided wherein the induction flux skin depth substantially corresponds with the central longitudinal axis of the crucible. Further, the drain tube may be positioned within the induction flux skin depth in relation to material within the crucible or may be substantially aligned with a direction of flow of molten material within the crucible. An improved head design including four shells forming thermal radiation shields and at least two gas-cooled plenums is also disclosed.

  9. Remote process cell mercury transfer pumps for DWPF

    International Nuclear Information System (INIS)

    Nielsen, M.G.; Vaughn, V.G.

    1986-01-01

    Final design and the results of the testing performed thus far show that the water displacement of mercury to a height of 40 feet is feasible with just 6 gallons of motive water. Control of the transfer is achieved by monitoring the pump discharge pressure. An air actuated plug valve configuration successfully contained the required discharge pressure of 260 psi. The requirements of low flow and maximum separation of mercury from particulates are achieved due to the configuration of the pressure canister. The pump is capable of transferring a discrete amount of mercury with little additional slurry particulates. The success of this new pumping configuration is highlighted by the fact that it was the inspiration for other remote transfer applications tested at SRP. These application include the dual canister sample pump shown in Figure 7, as well as a successful prototype pump designed at Pacific Northwest Laboratories (PNL). The PNL pump was designed for the purpose of metering waste slurries to an electric melter. Upon completion of final pump fabrication, the Defense Waste Processing facility (DWPF) facility will have a simple and highly reliable method of remotely transferring small discrete batches of mercury as required from radioactive process vessels. 3 refs., 7 figs., 1 tab

  10. Small-Scale High Temperature Melter-1 (SSHTM-1) Data Package. Appendix B

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-01

    This appendix provides the data for Alternate HTM Flowsheet 2 (Glycolic Acid) melter feed preparation activities in both the laboratory- and small-scale testing. The first section provides an outline of this appendix. The melter feed preparation data are presented in the next two main sections, laboratory melter feed preparation data and small-scale melter feed preparation data. Section 3.0 provides the laboratory data which is discussed in the main body of the Small-Scale High Temperature-1 (SSHTM-1) Data Package, milestone C95-02.02Y. Section 3.1 gives the flowsheet in outline form as used in the laboratory-scale tests. This section also includes the ``Laboratory Melter Feed Preparation Activity Log`` which gives A chronological account of the test in terms of time, temperature, slurry pH, and specific observations about slurry appearance, acid addition rates, and samples taken. The ``Laboratory Melter Feed Preparation Activity Log`` provides a road map to the reader by which all the activity and data from the laboratory can be easily accessed. A summary of analytical data is presented next, section 3.2, which covers starting materials and progresses to the analysis of the melter feed. The next section, 3.3, characterizes the off-gas generation that occurs during the slurry processing. The following section, 3.4, provides the rheology data gathered including gram waste oxide loading information for the various slurries tested. The final section, 3.5, includes data from standard crucible redox testing. Section 4.0 provides the small-scale data in parallel form to section 3.0. Section 5.0 concludes with the references for this appendix.

  11. Nuclear waste glass melter design including the power and control systems

    International Nuclear Information System (INIS)

    Chapman, C.C.

    1982-01-01

    An energy balance of a joule-heated nuclear waste glass melter is used to discuss the problems in the design of the melter geometry and in the specifications of the power and control systems. The relationships between geometry, electrode current density, production rate, load voltage, and load power are presented graphically. The influence of liquid feeding on the surface of the glass and the variability of nuclear waste glass on the design and control during operation is discussed. 10 refs

  12. LFCM [liquid-fed ceramic melter] vitrification technology: Quarterly progress report, January--March 1987

    International Nuclear Information System (INIS)

    Brouns, R. A.; Allen, C. R.; Powell, J. A.

    1988-05-01

    This report is compiled by the Nuclear Waste Treatment Program and the Hanford Waste Vitrification Program at Pacific Northwest Laboratory to describe the progress in developing, testing, applying and documenting liquid-fed ceramic melter vitrification technology. Progress in the following technical subject areas during the second quarter of FY 1987 is discussed: melting process chemistry and glass development, feed preparation and transfer systems, melter systems, canister filling and handling systems, and process/product modeling. 23 refs., 14 figs., 10 tabs

  13. U.S. Bureau of Mines, Phase 1 Hanford low-level waste melter tests. Final report

    International Nuclear Information System (INIS)

    Eaton, W.C.; Oden, L.L.; O'Connor, W.K.

    1995-11-01

    A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Phase 1 of the melter demonstration tests using simulated LLW was completed during fiscal year 1995. This document is the melter offgas report on testing performed by the U.S. Department of the Interior, Bureau of Mines, Albany Research Center in Albany, Oregon. The Bureau of Mines (one of the seven vendors selected) was chosen to demonstrate carbon electrode melter technology (also called carbon arc or electric arc) under WHC Subcontract number MMI-SVV-384216. The report contains description of the tests, observation, test data and some analysis of the data as it pertains to application of this technology for LLW vitrification. Testing consisted of melter feed preparation and three melter tests, the first of which was to fulfill the requirements of the statement of work (WHC-SD-EM-RD-044), and the second and third were to address issues identified during the first test. The document also contains summaries of the melter offgas report issued as a separate document U.S. Bureau of Mines, Phase 1 Hanford Low-Level Waste Melter Tests: Melter Offgas Report (WHC-SD-WM-VI-032)

  14. U.S. Bureau of Mines, Phase 1 Hanford low-level waste melter tests. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Eaton, W.C. [Westinghouse Hanford Co., Richland, WA (United States); Oden, L.L.; O`Connor, W.K. [Bureau of Mines, Albany, OR (United States). Albany Research Center

    1995-11-01

    A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Phase 1 of the melter demonstration tests using simulated LLW was completed during fiscal year 1995. This document is the melter offgas report on testing performed by the U.S. Department of the Interior, Bureau of Mines, Albany Research Center in Albany, Oregon. The Bureau of Mines (one of the seven vendors selected) was chosen to demonstrate carbon electrode melter technology (also called carbon arc or electric arc) under WHC Subcontract number MMI-SVV-384216. The report contains description of the tests, observation, test data and some analysis of the data as it pertains to application of this technology for LLW vitrification. Testing consisted of melter feed preparation and three melter tests, the first of which was to fulfill the requirements of the statement of work (WHC-SD-EM-RD-044), and the second and third were to address issues identified during the first test. The document also contains summaries of the melter offgas report issued as a separate document U.S. Bureau of Mines, Phase 1 Hanford Low-Level Waste Melter Tests: Melter Offgas Report (WHC-SD-WM-VI-032).

  15. Compilation of information on melter modeling

    International Nuclear Information System (INIS)

    Eyler, L.L.

    1996-03-01

    The objective of the task described in this report is to compile information on modeling capabilities for the High-Temperature Melter and the Cold Crucible Melter and issue a modeling capabilities letter report summarizing existing modeling capabilities. The report is to include strategy recommendations for future modeling efforts to support the High Level Waste (BLW) melter development

  16. DWPF integrated cold runs revised technical bases for precipitate hydrolysis

    International Nuclear Information System (INIS)

    Landon, L.F.

    1992-01-01

    The report defines new precipitate hydrolysis process operating parameters for DWPF Chemical runs assuming the precipitate feed simulants to be processed reflect the decision to implement a final wash of the tetraphenylborate slurry before transfer to DWPF (i.e. the Late Wash Facility). Control of the nitrite content of the tetraphenylborate slurry to 0.01M or less has eliminated the need for hydroxylamine nitrate (HAN) during hydrolysis. Consequently, the oxidant nitrous oxide will not be generated. However, nitric oxide (NO) is expected to be generated (reaction of formic acid with nitrite) and some fraction of the NO can be expected to be oxidized to nitrogen dioxide. The rate of NO generation with low nitrite feed has not been quantified at this time nor is the extent to which the NO is oxidized to NO 2 known. A mass spectrometer is being installed in the Precipitate Hydrolysis Experimental Facility (PHEF) which will enable the NO generation rate to be defined as well as the extent to which the NO is oxidized to NO 2 . There is some undocumented data available for C 6 H 6 /NO and C 6 H 6 /NO 2 with N 2 as the diluent but no similar data for CO 2 . Development of test data in the required time frame is not possible. However, MOC's will be estimated for benzene/NO/NO 2 /CO 2 gas mixtures (the MOC is expected to be approximately 60% less than for the HAN process). Once these data are obtained, and NO/NO 2 concentration profiles are obtained from PHEF hydrolysis process demonstrations, a flammability control strategy for the DWPF Salt Processing Cell will be developed. Implementation of the HAN process purge strategy upon startup of the SPC with the late wash process would be conservative

  17. Noble metals-compatible melter features development Phase 1: Establishing functional and design criteria and design concepts

    International Nuclear Information System (INIS)

    Elmore, M.R.; Siemens, D.H.; Chapman, C.C.

    1996-03-01

    Premature failures have occurred in melters at Japan's Tokai Mockup Facility and at the Federal Republic of Germany (FRG) PAMELA plant during processing of feeds with high levels of noble metals. Melter failure was due to the accumulation of an electrically conductive, noble metals-containing precipitates in the glass, that then resulted in short circuiting of the electrodes. A comparison was made of the anticipated Hanford Waste Vitrification Plant (HWVP) feed with the feeds processed in the FRG and Japanese melters. The evaluation showed that comparable levels of noble metals and other potential precipitate-forming components (e.g. Cr/Fe/Ni-spinels) exist in the HWVP feed. As a result, the HWVP project made a decision to modify the present reference melter design to include features to prevent the precipitation and accumulation or otherwise accommodate precipitated phases on a routine basis without loss of production capacity

  18. Initial demonstration of DWPF process and product control strategy using actual radioactive waste

    International Nuclear Information System (INIS)

    Andrews, M.K.; Bibler, N.E.; Jantzen, C.M.; Beam, D.C.

    1991-01-01

    The Defense Waste Processing Facility at the Savannah River Site (SRS) will vitrify high-level nuclear waste into borosilicate glass. The waste will be mixed with properly formulated glass-making frit and fed to a melter at 1150 degrees C. Process control and product quality are ensured by proper control of the melter feed composition. Algorithms have been developed to predict the processability of the melt and the durability of the final glass based on this feed composition. To test these algorithms, an actual radioactive waste contained in a shielded facility at SRS was analyzed and a frit composition formulated using a simple computer spreadsheet which contained the algorithms. This frit was then mixed with the waste and the resulting slurry fed to a research scale joule-heated melter operated remotely. Approximately 24 kg of glass were successfully prepared. This paper will describe the frit formulation, the vitrification process, and the glass durability

  19. Technical bases for the DWPF testing program

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1990-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) will be the first production facility in the United States for the immobilization of high-level nuclear waste. Production of DWPF canistered wasteforms will begin prior to repository licensing, so decisions on facility startup will have to be made before the final decisions on repository design are made. The Department of Energy's Office of Civilian Radioactive Waste Management (RW) has addressed this discrepancy by defining a Waste Acceptance Process. This process provides assurance that the borosilicate-glass wasteform, in a stainless-steel canister, produced by the DWPF will be acceptable for permanent storage in a federal repository. As part of this process, detailed technical specifications have been developed for the DWPF product. SRS has developed detailed strategies for demonstrating compliance with each of the Waste Acceptance Process specifications. An important part of the compliance is the testing which will be carried out in the DWPF. In this paper, the bases for each of the tests to be performed in the DWPF to establish compliance with the specifications are described, and the tests are detailed. The results of initial tests relating to characterization of sealed canisters are reported

  20. Americium/Curium Melter 2A Pilot Tests

    International Nuclear Information System (INIS)

    Smith, M.E.; Fellinger, A.P.; Jones, T.M.; Miller, C.B.; Miller, D.H.; Snyder, T.K.; Stone, M.E.; Witt, D.C.

    1998-05-01

    Isotopes of americium (Am) and curium (Cm) were produced in the past at the Savannah River Site (SRS) for research, medical, and radiological applications. These highly radioactive and valuable isotopes have been stored in an SRS reprocessing facility for a number of years. Vitrification of this solution will allow the material to be more safely stored until it is transported to the DOE Oak Ridge Reservation for use in research and medical applications. To this end, the Am/Cm Melter 2A pilot system, a full-scale non- radioactive pilot plant of the system to be installed at the reprocessing facility, was designed, constructed and tested. The full- scale pilot system has a frit and aqueous feed delivery system, a dual zone bushing melter, and an off-gas treatment system. The main items which were tested included the dual zone bushing melter, the drain tube with dual heating and cooling zones, glass compositions, and the off-gas system which used for the first time a film cooler/lower melter plenum. Most of the process and equipment were proven to function properly, but several problems were found which will need further work. A system description and a discussion of test results will be given

  1. Incorporating Cold Cap Behavior in a Joule-heated Waste Glass Melter Model

    Energy Technology Data Exchange (ETDEWEB)

    Varija Agarwal; Donna Post Guillen

    2013-08-01

    In this paper, an overview of Joule-heated waste glass melters used in the vitrification of high level waste (HLW) is presented, with a focus on the cold cap region. This region, in which feed-to-glass conversion reactions occur, is critical in determining the melting properties of any given glass melter. An existing 1D computer model of the cold cap, implemented in MATLAB, is described in detail. This model is a standalone model that calculates cold cap properties based on boundary conditions at the top and bottom of the cold cap. Efforts to couple this cold cap model with a 3D STAR-CCM+ model of a Joule-heated melter are then described. The coupling is being implemented in ModelCenter, a software integration tool. The ultimate goal of this model is to guide the specification of melter parameters that optimize glass quality and production rate.

  2. DWPF nitric-glycolic flowsheet chemical process cell chemistry. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Zamecnik, J. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-02-01

    The conversions of nitrite to nitrate, the destruction of glycolate, and the conversion of glycolate to formate and oxalate were modeled for the Nitric-Glycolic flowsheet using data from Chemical Process Cell (CPC) simulant runs conducted by SRNL from 2011 to 2015. The goal of this work was to develop empirical correlations for these variables versus measureable variables from the chemical process so that these quantities could be predicted a-priori from the sludge composition and measurable processing variables. The need for these predictions arises from the need to predict the REDuction/OXidation (REDOX) state of the glass from the Defense Waste Processing Facility (DWPF) melter. This report summarizes the initial work on these correlations based on the aforementioned data. Further refinement of the models as additional data is collected is recommended.

  3. Inorganic analyses of volatilized and condensed species within prototypic Defense Waste Processing Facility (DWPF) canistered waste

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1992-01-01

    The high-level radioactive waste currently stored in carbon steel tanks at the Savannah River Site (SRS) will be immobilized in a borosilicate glass in the Defense Waste Processing Facility (DWPF). The canistered waste will be sent to a geologic repository for final disposal. The Waste Acceptance Preliminary Specifications (WAPS) require the identification of any inorganic phases that may be present in the canister that may lead to internal corrosion of the canister or that could potentially adversely affect normal canister handling. During vitrification, volatilization of mixed (Na, K, Cs)Cl, (Na, K, Cs) 2 SO 4 , (Na, K, Cs)BF 4 , (Na, K) 2 B 4 O 7 and (Na,K)CrO 4 species from glass melt condensed in the melter off-gas and in the cyclone separator in the canister pour spout vacuum line. A full-scale DWPF prototypic canister filled during Campaign 10 of the SRS Scale Glass Melter was sectioned and examined. Mixed (NaK)CI, (NaK) 2 SO 4 , (NaK) borates, and a (Na,K) fluoride phase (either NaF or Na 2 BF 4 ) were identified on the interior canister walls, neck, and shoulder above the melt pour surface. Similar deposits were found on the glass melt surface and on glass fracture surfaces. Chromates were not found. Spinel crystals were found associated with the glass pour surface. Reference amounts of the halides and sulfates were found retained in the glass and the glass chemistry, including the distribution of the halides and sulfates, was homogeneous. In all cases where rust was observed, heavy metals (Zn, Ti, Sn) from the cutting blade/fluid were present indicating that the rust was a reaction product of the cutting fluid with glass and heat sensitized canister or with carbon-steel contamination on canister interior. Only minimal water vapor is present so that internal corrosion of the canister, will not occur

  4. Melter Disposal Strategic Planning Document

    Energy Technology Data Exchange (ETDEWEB)

    BURBANK, D.A.

    2000-09-25

    This document describes the proposed strategy for disposal of spent and failed melters from the tank waste treatment plant to be built by the Office of River Protection at the Hanford site in Washington. It describes program management activities, disposal and transportation systems, leachate management, permitting, and safety authorization basis approvals needed to execute the strategy.

  5. Assessment of water/glass interactions in waste glass melter operation

    International Nuclear Information System (INIS)

    Postma, A.K.; Chapman, C.C.; Buelt, J.L.

    1980-04-01

    A study was made to assess the possibility of a vapor explosion in a liquid-fed glass melter and during off-standard conditions for other vitrification processes. The glass melter considered is one designed for the vitrification of high-level nuclear wastes and is comprised of a ceramic-lined cavity with electrodes for joule heating and processing equipment required to add feed and withdraw glass. Vapor explosions needed to be considered because experience in other industrial processes has shown that violent interactions can occur if a hot liquid is mixed with a cooler, vaporizable liquid. Available experimental evidence and theoretical analyses indicate that destructive glass/water interactions are low probability events, if they are possible at all. Under standard conditions, aspects of liquid-fed melter operation which work against explosive interactions include: (1) the aqueous feed is near its boiling point; (2) the feed contains high concentrations of suspended particles; (3) molten glass has high viscosity (greater than 20 poise); and (4) the glass solidifies before film boiling can collapse. While it was concluded that vapor explosions are not expected in a liquid-fed melter, available information does not allow them to be ruled out altogether. Several precautionary measures which are easily incorporated into melter operation procedures were identified and additional experiments were recommended

  6. Hazards analysis of TNX Large Melter-Off-Gas System

    International Nuclear Information System (INIS)

    Randall, C.T.

    1982-03-01

    Analysis of the potential safety hazards and an evaluation of the engineered safety features and administrative controls indicate that the LMOG System can be operated without undue hazard to employees or the public, or damage to equipment. The safety features provided in the facility design coupled with the planned procedural and administrative controls make the occurrence of serious accidents very improbable. A set of recommendations evolved during this analysis that was judged potentially capable of further reducing the probability of personnel injury or further mitigating the consequences of potential accidents. These recommendations concerned areas such as formic acid vapor hazards, hazard of feeding water to the melter at an uncontrolled rate, prevention of uncontrolled glass pours due to melter pressure excursions and additional interlocks. These specific suggestions were reviewed with operational and technical personnel and are being incorporated into the process. The safeguards provided by these recommendations are discussed in this report

  7. Temperature control system for liquid-fed ceramic melters

    International Nuclear Information System (INIS)

    Westsik, J.H. Jr.

    1986-10-01

    A temperature-feedback system has been developed for controlling electrical power to liquid-fed ceramic melters (LFCM). Software, written for a microcomputer-based data acquisition and process monitoring system, compares glass temperatures with a temperature setpoint and adjusts the electrical power accordingly. Included in the control algorithm are steps to reject failed thermocouples, spatially average the glass temperatures, smooth the averaged temperatures over time using a digital filter, and detect foaming in the glass. The temperature control system has proved effective during all phases of melter operation including startup, steady operation, loss of feed, and shutdown. This system replaces current, power, and resistance feedback control systems used previously in controlling the LFCM process

  8. Technology of off-gas treatment for liquid-fed ceramic melters

    Energy Technology Data Exchange (ETDEWEB)

    Scott, P.A.; Goles, R.W.; Peters, R.D.

    1985-05-01

    The technology for treating off gas from liquid-fed ceramic melters (LFCMs) has been under development at the Pacific Northwest Laboratory since 1977. This report presents the off-gas technology as developed at PNL and by others to establish a benchmark of development and to identify technical issues. Tests conducted on simulated (nonradioactive) wastes have provided data that allow estimation of melter off-gas composition for a given waste. Mechanisms controlling volatilization of radionuclides and noxious gases are postulated, and correlations between melter operation and emissions are presented. This report is directed to those familiar with LFCM operation. Off-gas treatment systems always require primary quench scrubbers, aerosol scrubbers, and final particulate filters. Depending on the composition of the off gas, equipment for removal of ruthenium, iodine, tritium, and noxious gases may also be needed. Nitrogen oxides are the most common noxious gases requiring treatment, and can be controlled by aqueous absorption or catalytic conversion with ammonia. High efficiency particulate air (HEPA) filters should be used for final filtration. The design criteria needed for an off-gas system can be derived from emission regulations and composition of the melter feed. Conservative values for melter off-gas composition can be specified by statistical treatment of reported off-gas data. Statistical evaluation can also be used to predict the frequency and magnitude of normal surge events that occur in the melter. 44 refs., 28 figs., 17 tabs.

  9. Steam Explosions in Slurry-fed Ceramic Melters

    Energy Technology Data Exchange (ETDEWEB)

    Carter, J.T.

    2001-03-28

    This report assesses the potential and consequences of a steam explosion in Slurry Feed Ceramic Melters (SFCM). The principles that determine if an interaction is realistically probable within a SFCM are established. Also considered are the mitigating effects due to dissolved, non-condensable gas(es) and suspended solids within the slurry feed, radiation, high glass viscosity, and the existence of a cold cap. The report finds that, even if any explosion were to occur, however, it would not be large enough to compromise vessel integrity.

  10. Examination Of Sulfur Measurements In DWPF Sludge Slurry And SRAT Product Materials

    International Nuclear Information System (INIS)

    Bannochie, C. J.; Wiedenman, B. J.

    2012-01-01

    Savannah River National Laboratory (SRNL) was asked to re-sample the received SB7b WAPS material for wt. % solids, perform an aqua regia digestion and analyze the digested material by inductively coupled plasma - atomic emission spectroscopy (ICP-AES), as well as re-examine the supernate by ICP-AES. The new analyses were requested in order to provide confidence that the initial analytical subsample was representative of the Tank 40 sample received and to replicate the S results obtained on the initial subsample collected. The ICP-AES analyses for S were examined with both axial and radial detection of the sulfur ICP-AES spectroscopic emission lines to ascertain if there was any significant difference in the reported results. The outcome of this second subsample of the Tank 40 WAPS material is the first subject of this report. After examination of the data from the new subsample of the SB7b WAPS material, a team of DWPF and SRNL staff looked for ways to address the question of whether there was in fact insoluble S that was not being accounted for by ion chromatography (IC) analysis. The question of how much S is reaching the melter was thought best addressed by examining a DWPF Slurry Mix Evaporator (SME) Product sample, but the significant dilution of sludge material, containing the S species in question, that results from frit addition was believed to add additional uncertainty to the S analysis of SME Product material. At the time of these discussions it was believed that all S present in a Sludge Receipt and Adjustment Tank (SRAT) Receipt sample would be converted to sulfate during the course of the SRAT cycle. A SRAT Product sample would not have the S dilution effect resulting from frit addition, and hence, it was decided that a DWPF SRAT Product sample would be obtained and submitted to SRNL for digestion and sample preparation followed by a round-robin analysis of the prepared samples by the DWPF Laboratory, F/H Laboratories, and SRNL for S and sulfate. The

  11. Final Report Integrated DM1200 Melter Testing Using AZ-102 And C-106/AY-102 HLW Simulants: HLW Simulant Verification VSL-05R5800-1, Rev. 0, 6/27/05

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Gong, W.; Bardakci, T.; D'Angelo, N.A.; Brandys, M.; Kot, W.K.; Pegg, I.L.

    2011-01-01

    The principal objectives of the DM1200 melter tests were to determine the effects of feed rheology, feed solid content, and bubbler configuration on glass production rate and off-gas system performance while processing the HLW AZ-101 and C-106/AY-102 feed compositions; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components, as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and post test inspections of system components. The specific objectives (including test success criteria) of this testing, along with how each objective was met, are outlined in a table. The data provided in this Final Report address the impacts of HLW melter feed rheology on melter throughput and validation of the simulated HLW melter feeds. The primary purpose of this testing is to further validate/verify the HLW melter simulants that have been used for previous melter testing and to support their continued use in developing melter and off-gas related processing information for the Project. The primary simulant property in question is rheology. Simulants and melter feeds used in all previous melter tests were produced by direct addition of chemicals; these feed tend to be less viscous than rheological the upper-bound feeds made from actual wastes. Data provided here compare melter processing for the melter feed used in all previous DM100 and DM1200 tests (nominal melter feed) with feed adjusted by the feed vendor (NOAH Technologies) to be more viscous, thereby simulating more closely the upperbounding feed produced from actual waste. This report provides results of tests that are described in the Test Plan for this work. The Test Plan is responsive to one of several test objectives covered in the WTP Test Specification for this work; consequently, only part of the scope described in the Test Specification was addressed in this particular Test Plan. For the purpose of

  12. FINAL REPORT INTEGRATED DM1200 MELTER TESTING USING AZ 102 AND C 106/AY-102 HLW SIMULANTS: HLW SIMULANT VERIFICATION VSL-05R5800-1 REV 0 6/27/05

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    The principal objectives of the DM1200 melter tests were to determine the effects of feed rheology, feed solid content, and bubbler configuration on glass production rate and off-gas system performance while processing the HLW AZ-101 and C-106/AY-102 feed compositions; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components, as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and post test inspections of system components. The specific objectives (including test success criteria) of this testing, along with how each objective was met, are outlined in a table. The data provided in this Final Report address the impacts of HLW melter feed rheology on melter throughput and validation of the simulated HLW melter feeds. The primary purpose of this testing is to further validate/verify the HLW melter simulants that have been used for previous melter testing and to support their continued use in developing melter and off-gas related processing information for the Project. The primary simulant property in question is rheology. Simulants and melter feeds used in all previous melter tests were produced by direct addition of chemicals; these feed tend to be less viscous than rheological the upper-bound feeds made from actual wastes. Data provided here compare melter processing for the melter feed used in all previous DM100 and DM1200 tests (nominal melter feed) with feed adjusted by the feed vendor (NOAH Technologies) to be more viscous, thereby simulating more closely the upperbounding feed produced from actual waste. This report provides results of tests that are described in the Test Plan for this work. The Test Plan is responsive to one of several test objectives covered in the WTP Test Specification for this work; consequently, only part of the scope described in the Test Specification was addressed in this particular Test Plan. For the purpose of

  13. DWPF Simulant CPC Studies For SB8

    Energy Technology Data Exchange (ETDEWEB)

    Newell, J. D.

    2013-09-25

    Prior to processing a Sludge Batch (SB) in the Defense Waste Processing Facility (DWPF), flowsheet studies using simulants are performed. Typically, the flowsheet studies are conducted based on projected composition(s). The results from the flowsheet testing are used to 1) guide decisions during sludge batch preparation, 2) serve as a preliminary evaluation of potential processing issues, and 3) provide a basis to support the Shielded Cells qualification runs performed at the Savannah River National Laboratory (SRNL). SB8 was initially projected to be a combination of the Tank 40 heel (Sludge Batch 7b), Tank 13, Tank 12, and the Tank 51 heel. In order to accelerate preparation of SB8, the decision was made to delay the oxalate-rich material from Tank 12 to a future sludge batch. SB8 simulant studies without Tank 12 were reported in a separate report.1 The data presented in this report will be useful when processing future sludge batches containing Tank 12. The wash endpoint target for SB8 was set at a significantly higher sodium concentration to allow acceptable glass compositions at the targeted waste loading. Four non-coupled tests were conducted using simulant representing Tank 40 at 110-146% of the Koopman Minimum Acid requirement. Hydrogen was generated during high acid stoichiometry (146% acid) SRAT testing up to 31% of the DWPF hydrogen limit. SME hydrogen generation reached 48% of of the DWPF limit for the high acid run. Two non-coupled tests were conducted using simulant representing Tank 51 at 110-146% of the Koopman Minimum Acid requirement. Hydrogen was generated during high acid stoichiometry SRAT testing up to 16% of the DWPF limit. SME hydrogen generation reached 49% of the DWPF limit for hydrogen in the SME for the high acid run. Simulant processing was successful using previously established antifoam addition strategy. Foaming during formic acid addition was not observed in any of the runs. Nitrite was destroyed in all runs and no N2O was detected

  14. Control of high level radioactive waste-glass melters - Part 5: Modeling of complex redox effects

    International Nuclear Information System (INIS)

    Bickford, D.F.; Choi, A.S.

    1991-01-01

    Computerized thermodynamic computations are useful in predicting the sequence and products of redox reactions and in assessing process variations. The redox state of waste-glass melters is determined by balance between the reducing potential of organic compounds in the feed, and the oxidizing potential of gases above the melt, and nitrates and polyvalent elements in the waste. Semiquantitative models predicting limitations of organic content have been developed based on crucible testing. Continuous melter test results have been compared to this improved staged-thermodynamic model of redox behavior

  15. Technical information report: Plasma melter operation, reliability, and maintenance analysis

    International Nuclear Information System (INIS)

    Hendrickson, D.W.

    1995-01-01

    This document provides a technical report of operability, reliability, and maintenance of a plasma melter for low-level waste vitrification, in support of the Hanford Tank Waste Remediation System (TWRS) Low-Level Waste (LLW) Vitrification Program. A process description is provided that minimizes maintenance and downtime and includes material and energy balances, equipment sizes and arrangement, startup/operation/maintence/shutdown cycle descriptions, and basis for scale-up to a 200 metric ton/day production facility. Operational requirements are provided including utilities, feeds, labor, and maintenance. Equipment reliability estimates and maintenance requirements are provided which includes a list of failure modes, responses, and consequences

  16. Liquidus Temperature Data for DWPF Glass

    International Nuclear Information System (INIS)

    Piepel, G.F.; Vienna, J.D.; Crum, J.V.; Mika, M.; Hrma, P.

    1999-01-01

    This report provides new liquidus temperature (T L ) versus composition data that can be used to reduce uncertainty in T L calculation for DWPF glass. According to the test plan and test matrix design PNNL has measured T L for 53 glasses within and just outside of the current DWPF processing composition window. The T L database generated under this task will directly support developing and enhancing the current T L process-control model. Preliminary calculations have shown a high probability of increasing HLW loading in glass produced at the SRS and Hanford. This increase in waste loading will decrease the life-cycle tank cleanup costs by decreasing process time and the volume of waste glass produced

  17. GTS Duratek, Phase I Hanford low-level waste melter tests: 100-kg melter offgas report

    International Nuclear Information System (INIS)

    Eaton, W.C.

    1995-11-01

    A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Phase 1 of the melter demonstration tests using simulated LLW was completed during fiscal year 1995. This document is the 100-kg melter offgas report on testing performed by GTS Duratek, Inc., in Columbia, Maryland. GTS Duratek (one of the seven vendors selected) was chosen to demonstrate Joule heated melter technology under WHC subcontract number MMI-SVV-384215. The document contains the complete offgas report on the 100-kg melter as prepared by Parsons Engineering Science, Inc. A summary of this report is also contained in the GTS Duratek, Phase I Hanford Low-Level Waste Melter Tests: Final Report (WHC-SD-WM-VI-027)

  18. Melter Throughput Enhancements for High-Iron HLW

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Gan, Hoa [The Catholic University of America, Washington, DC (United States); Joseph, Innocent [The Catholic University of America, Washington, DC (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States); Chaudhuri, Malabika [The Catholic University of America, Washington, DC (United States); Kot, Wing [The Catholic University of America, Washington, DC (United States)

    2012-12-26

    This report describes work performed to develop and test new glass and feed formulations in order to increase glass melting rates in high waste loading glass formulations for HLW with high concentrations of iron. Testing was designed to identify glass and melter feed formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts to assess melt rate using a vertical gradient furnace system and to develop new formulations with enhanced melt rate. Testing evaluated the effects of waste loading on glass properties and the maximum waste loading that can be achieved. The results from crucible-scale testing supported subsequent DuraMelter 100 (DM100) tests designed to examine the effects of enhanced glass and feed formulations on waste processing rate and product quality. The DM100 was selected as the platform for these tests due to its extensive previous use in processing rate determination for various HLW streams and glass compositions.

  19. DWPF waste form compliance plan (Draft Revision)

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Marra, S.L.

    1991-01-01

    The Department of Energy currently has over 100 million liters of high-level radioactive waste in storage at the Savannah River Site (SRS). In the late 1970's, the Department of Energy recognized that there were significant safety and cost advantages associated with immobilizing the high-level waste in a stable solid form. Several alternative waste forms were evaluated in terms of product quality and reliability of fabrication. This evaluation led to a decision to build the Defense Waste Processing Facility (DWPF) at SRS to convert the easily dispersed liquid waste to borosilicate glass. In accordance with the NEPA (National Environmental Policy Act) process, an Environmental Impact Statement was prepared for the facility, as well as an Environmental Assessment of the alternative waste forms, and issuance of a Record of Decision (in December, 1982) on the waste form. The Department of Energy, recognizing that start-up of the DWPF would considerably precede licensing of a repository, instituted a Waste Acceptance Process to ensure that these canistered waste forms would be acceptable for eventual disposal at a federal repository. This report is a revision of the DWPF compliance plan

  20. ENHANCED DOE HIGH LEVEL WASTE MELTER THROUGHPUT STUDIES: SRNL GLASS SELECTION STRATEGY

    Energy Technology Data Exchange (ETDEWEB)

    Raszewski, F; Tommy Edwards, T; David Peeler, D

    2008-01-23

    The Department of Energy has authorized a team of glass formulation and processing experts at the Savannah River National Laboratory (SRNL), the Pacific Northwest National Laboratory (PNNL), and the Vitreous State Laboratory (VSL) at Catholic University of America to develop a systematic approach to increase high level waste melter throughput (by increasing waste loading with minimal or positive impacts on melt rate). This task is aimed at proof-of-principle testing and the development of tools to improve waste loading and melt rate, which will lead to higher waste throughput. Four specific tasks have been proposed to meet these objectives (for details, see WSRC-STI-2007-00483): (1) Integration and Oversight, (2) Crystal Accumulation Modeling (led by PNNL)/Higher Waste Loading Glasses (led by SRNL), (3) Melt Rate Evaluation and Modeling, and (4) Melter Scale Demonstrations. Task 2, Crystal Accumulation Modeling/Higher Waste Loading Glasses is the focus of this report. The objective of this study is to provide supplemental data to support the possible use of alternative melter technologies and/or implementation of alternative process control models or strategies to target higher waste loadings (WLs) for the Defense Waste Processing Facility (DWPF)--ultimately leading to higher waste throughputs and a reduced mission life. The glass selection strategy discussed in this report was developed to gain insight into specific technical issues that could limit or compromise the ability of glass formulation efforts to target higher WLs for future sludge batches at the Savannah River Site (SRS). These technical issues include Al-dissolution, higher TiO{sub 2} limits and homogeneity issues for coupled-operations, Al{sub 2}O{sub 3} solubility, and nepheline formation. To address these technical issues, a test matrix of 28 glass compositions has been developed based on 5 different sludge projections for future processing. The glasses will be fabricated and characterized based on

  1. Final Report - Glass Formulation Testing to Increase Sulfate Volatilization from Melter, VSL-04R4970-1, Rev. 0, dated 2/24/05

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Matlack, K. A.; Pegg, I. L.; Gong, W.

    2013-11-13

    The principal objectives of the DM100 and DM10 tests were to determine the impact of four different organics and one inorganic feed additive on sulfate volatilization and to determine the sulfur partitioning between the glass and the off-gas system. The tests provided information on melter processing characteristics and off-gas data including sulfur incorporation and partitioning. A series of DM10 and DM100 melter tests were conducted using a LAW Envelope A feed. The testing was divided into three parts. The first part involved a series of DM10 melter tests with four different organic feed additives: sugar, polyethylene glycol (PEG), starch, and urea. The second part involved two confirmatory 50-hour melter tests on the DM100 using the best combination of reductants and conditions based on the DM10 results. The third part was performed on the DM100 with feeds containing vanadium oxide (V{sub 2}O{sub 5}) as an inorganic additive to increase sulfur partitioning to the off-gas. Although vanadium oxide is not a reductant, previous testing has shown that vanadium shows promise for partitioning sulfur to the melter exhaust, presumably through its known catalytic effect on the SO{sub 2}/SO{sub 3} reaction. Crucible-scale tests were conducted prior to the melter tests to confirm that the glasses and feeds would be processable in the melter and that the glasses would meet the waste form (ILAW) performance requirements. Thus, the major objectives of these tests were to: Perform screening tests on the DM10 followed by tests on the DM100-WV system using a LAW -Envelope A feed with four organic additives to assess their impact on sulfur volatilization. Perform tests on the DM100-WV system using a LAW -Envelope A feed containing vanadium oxide to assess its impact on sulfur volatilization. Determine feed processability and product quality with the above additives. Collect melter emissions data to determine the effect of additives on sulfur partitioning and melter emissions

  2. DWPF liquid sample station: Status of equipment development

    International Nuclear Information System (INIS)

    Caplan, J.R.

    1987-01-01

    This report summarizes operating experience and equipment status of the DWPF liquid sample cell. Operation hours to date, results of equipment inspections and problems encountered and their solutions are discussed. An equipment and instrumentation status updating DPST-85-592, DWPF LIQUID SAMPLE CELL MOCK-UP, is presented. Remaining development items are also outlined

  3. Measurement of the volatility and glass transition temperatures of glasses produced during the DWPF startup test program

    International Nuclear Information System (INIS)

    Marra, J.C.; Harbour, J.R.

    1995-01-01

    The Defense Waste Processing Facility (DWPF) will immobilize high-level radioactive waste currently stored in underground tanks at the Savannah River Site by incorporating the waste into a glass matrix. The molten waste glass will be poured into stainless steel canisters which will be welded shut to produce the final waste form. One specification requires that any volatiles produced as a result of accidentally heating the waste glass to the glass transition temperature be identified. Glass samples from five melter campaigns, run as part of the DWPF Startup Test Program, were analyzed to determine glass transition temperatures and to examine the volatilization (by weight loss). Glass transition temperatures (T g ) for the glasses, determined by differential scanning calorimetry (DSC), ranged between 445 C and 474 C. Thermogravimetric analysis (TGA) scans showed that no overall weight loss occurred in any of the glass samples when heated to 500 C. Therefore, no volatility will occur in the final glass product when heated up to 500 C

  4. GLYCOLIC-NITRIC ACID FLOWSHEET DEMONSTRATION OF THE DWPF CHEMICAL PROCESSING CELL WITH MATRIX SIMULANTS AND SUPERNATE

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, D.; Stone, M.; Newell, J.; Best, D.

    2012-05-07

    Savannah River Remediation (SRR) is evaluating changes to its current DWPF flowsheet to improve processing cycle times. This will enable the facility to support higher canister production while maximizing waste loading. Higher throughput is needed in the CPC since the installation of the bubblers into the melter has increased melt rate. Due to the significant maintenance required for the DWPF gas chromatographs (GC) and the potential for production of flammable quantities of hydrogen, reducing or eliminating the amount of formic acid used in the CPC is being developed. Earlier work at Savannah River National Laboratory has shown that replacing formic acid with an 80:20 molar blend of glycolic and formic acids has the potential to remove mercury in the SRAT without any significant catalytic hydrogen generation. This report summarizes the research completed to determine the feasibility of processing without formic acid. In earlier development of the glycolic-formic acid flowsheet, one run (GF8) was completed without formic acid. It is of particular interest that mercury was successfully removed in GF8, no formic acid at 125% stoichiometry. Glycolic acid did not show the ability to reduce mercury to elemental mercury in initial screening studies, which is why previous testing focused on using the formic/glycolic blend. The objective of the testing detailed in this document is to determine the viability of the nitric-glycolic acid flowsheet in processing sludge over a wide compositional range as requested by DWPF. This work was performed under the guidance of Task Technical and Quality Assurance Plan (TT and QAP). The details regarding the simulant preparation and analysis have been documented previously.

  5. GLYCOLIC-NITRIC ACID FLOWSHEET DEMONSTRATION OF THE DWPF CHEMICAL PROCESS CELL WITH SLUDGE AND SUPERNATE SIMULANTS

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, D.; Stone, M.; Newell, J.; Best, D.; Zamecnik, J.

    2012-08-28

    Savannah River Remediation (SRR) is evaluating changes to its current Defense Waste Processing Facility (DWPF) flowsheet to improve processing cycle times. This will enable the facility to support higher canister production while maximizing waste loading. Higher throughput is needed in the Chemical Process Cell (CPC) since the installation of the bubblers into the melter has increased melt rate. Due to the significant maintenance required for the DWPF gas chromatographs (GC) and the potential for production of flammable quantities of hydrogen, reducing or eliminating the amount of formic acid used in the CPC is being developed. Earlier work at Savannah River National Laboratory has shown that replacing formic acid with an 80:20 molar blend of glycolic and formic acids has the potential to remove mercury in the SRAT without any significant catalytic hydrogen generation. This report summarizes the research completed to determine the feasibility of processing without formic acid. In earlier development of the glycolic-formic acid flowsheet, one run (GF8) was completed without formic acid. It is of particular interest that mercury was successfully removed in GF8, no formic acid at 125% stoichiometry. Glycolic acid did not show the ability to reduce mercury to elemental mercury in initial screening studies, which is why previous testing focused on using the formic/glycolic blend. The objective of the testing detailed in this document is to determine the viability of the nitric-glycolic acid flowsheet in processing sludge over a wide compositional range as requested by DWPF. This work was performed under the guidance of Task Technical and Quality Assurance Plan (TT&QAP). The details regarding the simulant preparation and analysis have been documented previously.

  6. Analysis of cascade impactor and EPA method 29 data from the americium/curium pilot melter system

    International Nuclear Information System (INIS)

    Zamecnik, J.R.

    1997-11-01

    The offgas system of the Am/Cm pilot melter at TNX was characterized by measuring the particulate evolution using a cascade impactor and EPA Method 29. This sampling work was performed by John Harden of the Clemson Environmental Technologies Laboratory, under SCUREF Task SC0056. Elemental analyses were performed by the SRTC Mobile Laboratory.Operation of the Am/Cm melter with B2000 frit has resulted in deposition of PbO and boron compounds in the offgas system that has contributed to pluggage of the High Efficiency Mist Eliminator (HEME). Sampling of the offgas system was performed to quantify the amount of particulate in the offgas system under several sets of conditions. Particulate concentration and particle size distribution were measured just downstream of the melter pressure control air addition port and at the HEME inlet. At both locations, the particulate was measured with and without steam to the film cooler while the melter was idled at about 1450 degrees Celsius. Additional determinations were made at the melter location during feeding and during idling at 1150 degrees Celsius rather than 1450 degrees Celsius (both with no steam to the film cooler). Deposition of particulates upstream of the melter sample point may have, and most likely did occur in each run, so the particulate concentrations measured do no necessarily reflect the total particulate emission at the melt surface. However, the data may be used in a relative sense to judge the system performance

  7. Characterization of a High-Level Waste Cold Cap in a Laboratory-Scale Melter

    Energy Technology Data Exchange (ETDEWEB)

    Dixona, Derek R; Schweiger, Michael J; Hrma, Pavel [Pacific Northwest National Laboratory, Richland (United States)

    2013-05-15

    The feed, slurry or calcine, is charged to the melter from above. The conversion of the melter feed to molten glass occurs within the cold cap, a several centimeters thin layer of the reacting material blanketing the surface of the melt. Between the cold-cap top, which is covered by boiling slurry, and its bottom, where bubbles separate it from molten glass, the temperature changes by ∼900 .deg. C. The heat is delivered to the cold cap from the melt that is stirred mainly by bubbling. The feed contains oxides, hydroxides, acids, inorganic salts and organic materials. On heating, these components react, releasing copious amounts of gases, while molten salts decompose, glass-forming melt is generated, and crystalline phases precipitate and dissolve in the melt. Most of these processes have been studied in detail and became sufficiently understood for a mathematical model to represent the heat and mass transfer within the cold cap. This allows US to relate the rate of melting to the feed properties. While the melting reactions can be studied, and feed properties, such as heat conductivity and density, measured in the laboratory, the actual cold-cap dynamics, as it evolves in the waste glass melter, is not accessible to direct investigation. Therefore, to bridge the gap between the laboratory crucible and the waste glass melter, we explored the cold cap formation in a laboratory-scale melter (LSM) and studied the structure of quenched cold caps. The LSM is a suitable tool for investigating the cold cap. The cold cap that formed in the LSM experiments exhibited macroscopic features observed in scaled melters, as well as microscopic features accessible through laboratory studies and mathematical modeling. The cold cap consists of two main layers. The top layer contains solid particles dissolving in the glass-forming melt and open shafts through which gases are escaping. The bottom layer contains bubbly melt or foam where bubbles coalesce into larger cavities that move

  8. Sludge Washing And Demonstration Of The DWPF Flowsheet In The SRNL Shielded Cells For Sludge Batch 8 Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Pareizs, J. M.; Crawford, C. L.

    2013-04-26

    The current Waste Solidification Engineering (WSE) practice is to prepare sludge batches in Tank 51 by transferring sludge from other tanks to Tank 51. Tank 51 sludge is washed and transferred to Tank 40, the current Defense Waste Processing Facility (DWPF) feed tank. Prior to transfer of Tank 51 to Tank 40, the Savannah River National Laboratory (SRNL) typically simulates the Tank Farm and DWPF processes using a Tank 51 sample (referred to as the qualification sample). WSE requested the SRNL to perform characterization on a Sludge Batch 8 (SB8) sample and demonstrate the DWPF flowsheet in the SRNL shielded cells for SB8 as the final qualification process required prior to SB8 transfer from Tank 51 to Tank 40. A 3-L sample from Tank 51 (the SB8 qualification sample; Tank Farm sample HTF-51-12-80) was received by SRNL on September 20, 2012. The as-received sample was characterized prior to being washed. The washed material was further characterized and used as the material for the DWPF process simulation including a Sludge Receipt and Adjustment Tank (SRAT) cycle, a Slurry Mix Evaporator (SME) cycle, and glass fabrication and chemical durability measurements.

  9. NEXT GENERATION MELTER(S) FOR VITRIFICATION OF HANFORD WASTE: STATUS AND DIRECTION

    International Nuclear Information System (INIS)

    Ramsey, W.G.; Gray, M.F.; Calmus, R.B.; Edge, J.A.; Garrett, B.G.

    2011-01-01

    Vitrification technology has been selected to treat high-level waste (HLW) at the Hanford Site, the West Valley Demonstration Project and the Savannah River Site (SRS), and low activity waste (LAW) at Hanford. In addition, it may potentially be applied to other defense waste streams such as sodium bearing tank waste or calcine. Joule-heated melters (already in service at SRS) will initially be used at the Hanford Site's Waste Treatment and Immobilization Plant (WTP) to vitrify tank waste fractions. The glass waste content and melt/production rates at WTP are limited by the current melter technology. Significant reductions in glass volumes and mission life are only possible with advancements in melter technology coupled with new glass formulations. The Next Generation Melter (NGM) program has been established by the U.S. Department of Energy's (DOE's), Environmental Management Office of Waste Processing (EM-31) to develop melters with greater production capacity (absolute glass throughput rate) and the ability to process melts with higher waste fractions. Advanced systems based on Joule-Heated Ceramic Melter (JHCM) and Cold Crucible Induction Melter (CCIM) technologies will be evaluated for HLW and LAW processing. Washington River Protection Solutions (WRPS), DOE's tank waste contractor, is developing and evaluating these systems in cooperation with EM-31, national and university laboratories, and corporate partners. A primary NGM program goal is to develop the systems (and associated flowsheets) to Technology Readiness Level 6 by 2016. Design and testing are being performed to optimize waste glass process envelopes with melter and balance of plant requirements. A structured decision analysis program will be utilized to assess the performance of the competing melter technologies. Criteria selected for the decision analysis program will include physical process operations, melter performance, system compatibility and other parameters.

  10. Vitrification of SRP waste by a slurry-fed ceramic melter

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1980-01-01

    Savannah River Plant (SRP) high-level waste (HLW) can be vitrified by feeding a slurry, instead of a calcine, to a joule-heated ceramic melter. Potential advantages of slurry feeding include (1) use of simpler equipment, (2) elimination of handling easily dispersed radioactive powder, (3) simpler process control, (4) effective mixing, (5) reduced off-gas volume, and (6) cost savings. Assessment of advantages and disadvantages of slurry feeding along with experimental studies indicate that slurry feeding is a promising way of vitrifying waste

  11. The corrosion behavior of DWPF glasses

    International Nuclear Information System (INIS)

    Ebert, W.L.; Bates, J.K.

    1995-01-01

    The authors analyzed the corroded surfaces of reference glasses developed for the Defense Waste Processing Facility (DWPF) to characterize their corrosion behavior. The corrosion mechanism of nuclear waste glasses must be known in order to provide source terms describing radionuclide release for performance assessment calculations. Different DWPF reference glasses were corroded under conditions that highlighted various aspects of the corrosion process and led to different extents of corrosion. The glasses corroded by similar mechanisms, and a phenomenological description of their corrosion behavior is presented here. The initial leaching of soluble glass components results in the formation of an amorphous gel layer on the glass surface. The gel layer is a transient phase that transforms into a layer of clay crystallites, which equilibrates with the solution as corrosion continues. The clay layer does not act as a barrier to either water penetration or glass dissolution, which continues beneath it, and may eventually separate from the glass. Solubility limits for glass components may be established by the eventual precipitation of secondary phases; thus, corrosion of the glass becomes controlled by the chemical equilibrium between the solution and the assemblage of secondary phases. In effect, the solution is an intermediate phase through which the glass transforms to an energetically more favorable assemblage of phases. Implications regarding the prediction of long-term glass corrosion behavior are discussed

  12. Evaluation of ISDP Batch 2 Qualification Compliance to 512-S, DWPF, Tank Farm, and Saltstone Waste Acceptance Criteria

    Energy Technology Data Exchange (ETDEWEB)

    Shafer, A.

    2010-05-05

    The purpose of this report is to document the acceptability of the second macrobatch (Salt Batch 2) of Tank 49H waste to H Tank Farm, DWPF, and Saltstone for operation of the Interim Salt Disposition Project (ISDP). Tank 49 feed meets the Waste Acceptance Criteria (WAC) requirements specified by References 11, 12, and 13. Salt Batch 2 material is qualified and ready to be processed through ARP/MCU to the final disposal facilities.

  13. Glass sampling program during DWPF Integrated Cold Runs

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1990-01-01

    The described glass sampling program is designed to achieve two objectives: To demonstrate Defense Waste Processing Facility (DWPF) ability to control and verify the radionuclide release properties of the glass product; To confirm DWPF's readiness to obtain glass samples during production, and SRL's readiness to analyze and test those samples remotely. The DWPF strategy for control of the radionuclide release properties of the glass product, and verification of its acceptability are described in this report. The basic approach of the test program is then defined

  14. Immobilization of high-level defense waste in a slurry-fed electric glass melter

    International Nuclear Information System (INIS)

    Brouns, R.A.; Mellinger, G.B.; Nelson, T.A.; Oma, K.H.

    1980-11-01

    Scoping studies have been performed at the Pacific Northwest Laboratory related to the direct liquid-feeding of a generic high-level defense waste to a joule-heated ceramic melter. Tests beginning on the laboratory scale and progressing to full-scale operation are reported. Laboratory work identified the need for a reducing agent in the feed to help control the foaming tendencies of the waste glass. These tests also indicated that suspension agents were helpful in reducing the tendency of solids to settle out of the liquid feed. Testing was then moved to a larger pilot-scale melter (designed for approx. 2.5 kg/h) where verification of the flowsheet examined in the lab was accomplished. It was found that the reducing agent controlled foaming and did not result in the precipitation of metals. Pumping problems were encountered when slurries with higher than normal solids content were fed. A demonstration (designed for approx. 50 kg/h) in a full-scale melter was then made with the tested flowsheet; however, the amount of reducing agent had to be increased. In addition, it was found that feed control needed further development; however, steady-state operation was achieved giving encouraging results on process capacities. During steady-state operation, ruthenium losses to the offgas system averaged less than 0.16%, while cesium losses were somewhat higher, ranging from 0.91 to 24% and averaging 13%. Particulate decontamination factors from feed to offgas in the melter ranged from 5 x 10 2 to greater than 10 3 without any filtration or treatment. Approximately 1050 kg of glass was produced from 2900 L of waste at rates up to 40 kg/h

  15. Steady state simulation of Joule heated ceramic melter for vitrification of high level liquid waste

    Energy Technology Data Exchange (ETDEWEB)

    Sugilal, G; Wattal, P K; Theyyunni, T K [Process Engineering and Systems Development Division, Bhabha Atomic Research Centre, Mumbai (India); Iyer, K N [Department of Mechanical Engineering, Indian Inst. of Tech., Mumbai (India)

    1994-06-01

    The Joule heated ceramic melter is emerging as an attractive alternative to metallic melters for high level waste vitrification. The inherent limitations with metallic melters viz., low capacity and short melter life, are overcome in a ceramic melter which can be adopted for continuous mode of operation. The ceramic melter has the added advantage of better operational flexibility. This paper describes the three dimensional model used for simulating the complex design conditions of the ceramic melter. (author).

  16. Steady state simulation of Joule heated ceramic melter for vitrification of high level liquid waste

    International Nuclear Information System (INIS)

    Sugilal, G.; Wattal, P.K.; Theyyunni, T.K.; Iyer, K.N.

    1994-01-01

    The Joule heated ceramic melter is emerging as an attractive alternative to metallic melters for high level waste vitrification. The inherent limitations with metallic melters viz., low capacity and short melter life, are overcome in a ceramic melter which can be adopted for continuous mode of operation. The ceramic melter has the added advantage of better operational flexibility. This paper describes the three dimensional model used for simulating the complex design conditions of the ceramic melter. (author)

  17. Vitrification of HLW in cold crucible melter

    International Nuclear Information System (INIS)

    Bordier, G.

    2005-01-01

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel the CEA (French Atomic Energy Commission), COGEMA (Industrial Operator), and SGN (COGEMA's Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities: the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed remotely in one of the R7 vitrification

  18. Baseline tests for arc melter vitrification of INEL buried wastes. Volume 1: Facility description and summary data report

    International Nuclear Information System (INIS)

    Oden, L.L.; O'Connor, W.K.; Turner, P.C.; Soelberg, N.R.; Anderson, G.L.

    1993-01-01

    This report presents field results and raw data from the Buried Waste Integrated Demonstration (BWID) Arc Melter Vitrification Project Phase 1 baseline test series conducted by the Idaho National Engineering Laboratory (INEL) in cooperation with the U.S. Bureau of Mines (USBM). The baseline test series was conducted using the electric arc melter facility at the USBM Albany Research Center in Albany, Oregon. Five different surrogate waste feed mixtures were tested that simulated thermally-oxidized, buried, TRU-contaminated, mixed wastes and soils present at the INEL. The USBM Arc Furnace Integrated Waste Processing Test Facility includes a continuous feed system, the arc melting furnace, an offgas control system, and utilities. The melter is a sealed, 3-phase alternating current (ac) furnace approximately 2 m high and 1.3 m wide. The furnace has a capacity of 1 metric ton of steel and can process as much as 1,500 lb/h of soil-type waste materials. The surrogate feed materials included five mixtures designed to simulate incinerated TRU-contaminated buried waste materials mixed with INEL soil. Process samples, melter system operations data and offgas composition data were obtained during the baseline tests to evaluate the melter performance and meet test objectives. Samples and data gathered during this program included (a) automatically and manually logged melter systems operations data, (b) process samples of slag, metal and fume solids, and (c) offgas composition, temperature, velocity, flowrate, moisture content, particulate loading and metals content. This report consists of 2 volumes: Volume I summarizes the baseline test operations. It includes an executive summary, system and facility description, review of the surrogate waste mixtures, and a description of the baseline test activities, measurements, and sample collection. Volume II contains the raw test data and sample analyses from samples collected during the baseline tests

  19. Chemical compatibility of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    Harbour, J.R.

    1993-01-01

    The Waste Acceptance Preliminary Specifications (WAPS) require that the contents of the canistered waste form are compatible with one another and the stainless steel canister. The canistered waste form is a closed system comprised of a stainless steel vessel containing waste glass, air, and condensate. This system will experience a radiation field and an elevated temperature due to radionuclide decay. This report discusses possible chemical reactions, radiation interactions, and corrosive reactions within this system both under normal storage conditions and after exposure to temperatures up to the normal glass transition temperature, which for DWPF waste glass will be between 440 and 460 degrees C. Specific conclusions regarding reactions and corrosion are provided. This document is based on the assumption that the period of interim storage prior to packaging at the federal repository may be as long as 50 years

  20. Parametric testing of a DWPF glass

    International Nuclear Information System (INIS)

    Bazan, F.; Rego, J.

    1985-03-01

    A series of tests has been performed to characterize the chemical stability of a DWPF borosilicate glass sample as part of the Waste Package Task of the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. This material was prepared at the Savannah River Laboratory for the purpose of testing the 165-frit matrix doped with a simulated nonradioactive waste. All tests were conducted at 90 0 C using deionized water and J-13 water (a tuffaceous formation ground water). In the deionized water tests, both monoliths and crushed glass were tested at various ratios of surface area of the sample to volume of water in order to compare leach rates for different sample geometries or leaching times. Effects on the leach rates as a result of the presence of crushed tuff and stainless steel material were also investigated in the tests with J-13 water. 3 refs., 12 figs., 7 tabs

  1. WTP Waste Feed Qualification: Glass Fabrication Unit Operation Testing Report

    Energy Technology Data Exchange (ETDEWEB)

    Stone, M. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Hanford Missions Programs; Newell, J. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Process Technology Programs; Johnson, F. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Engineering Process Development; Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Engineering Process Development

    2016-07-14

    The waste feed qualification program is being developed to protect the Hanford Tank Waste Treatment and Immobilization Plant (WTP) design, safety basis, and technical basis by assuring waste acceptance requirements are met for each staged waste feed campaign prior to transfer from the Tank Operations Contractor to the feed receipt vessels inside the Pretreatment Facility. The Waste Feed Qualification Program Plan describes the three components of waste feed qualification: 1. Demonstrate compliance with the waste acceptance criteria 2. Determine waste processability 3. Test unit operations at laboratory scale. The glass fabrication unit operation is the final step in the process demonstration portion of the waste feed qualification process. This unit operation generally consists of combining each of the waste feed streams (high-level waste (HLW) and low-activity waste (LAW)) with Glass Forming Chemicals (GFCs), fabricating glass coupons, performing chemical composition analysis before and after glass fabrication, measuring hydrogen generation rate either before or after glass former addition, measuring rheological properties before and after glass former addition, and visual observation of the resulting glass coupons. Critical aspects of this unit operation are mixing and sampling of the waste and melter feeds to ensure representative samples are obtained as well as ensuring the fabrication process for the glass coupon is adequate. Testing was performed using a range of simulants (LAW and HLW simulants), and these simulants were mixed with high and low bounding amounts of GFCs to evaluate the mixing, sampling, and glass preparation steps in shielded cells using laboratory techniques. The tests were performed with off-the-shelf equipment at the Savannah River National Laboratory (SRNL) that is similar to equipment used in the SRNL work during qualification of waste feed for the Defense Waste Processing Facility (DWPF) and other waste treatment facilities at the

  2. Detailed design data package: 3.1a-Film cooler pressure drop data; Item 3.2a - SBS packing selection; Item 3.2b, 3.2c - Pressure drop data for SBS distribution plate; and Item 3.2e - SBS distribution plate and liquid risers. PHTD pilot-scale melter testing system cost account milesonte 1.2.2.04.15A

    International Nuclear Information System (INIS)

    Whyatt, G.A.; Anderson, L.D.; Evans, J. II.

    1996-03-01

    This data package transmits information collected on the Liquid-Fed Ceramic Melter (LFCM) offgas system prior to melter feeding operations. Injection of steam to the melter plenum was used to simulate feeding of the melter. Steam surge cases were studied under steady-state surge conditions. Dynamic surges will be examined under data needs. The Fluor data needs included two blank tables requesting specific information for data needs 3.1 and 3.2. These tables are provided in Tables S.1 and S.2 below with the requested information filled in

  3. Cylindrical Induction Melter Modicon Control System

    International Nuclear Information System (INIS)

    Weeks, G.E.

    1998-04-01

    In the last several years an extensive R ampersand D program has been underway to develop a vitrification system to stabilize Americium (Am) and Curium (Cm) inventories at SRS. This report documents the Modicon control system designed for the 3 inch Cylindrical Induction Melter (CIM)

  4. Americium/curium bushing melter drain tests

    International Nuclear Information System (INIS)

    Smith, M.E.; Hardy, B.J.; Smith, M.E.

    1997-01-01

    Americium and curium were produced in the past at the Savannah River Site (SRS) for research, medical, and radiological applications. They have been stored in a nitric acid solution in an SRS reprocessing facility for a number of years. Vitrification of the americium/curium (Am/Cm) solution will allow the material to be safely stored or transported to the DOE Oak Ridge Reservation. Oak Ridge is responsible for marketing radionuclides for research and medical applications. The bushing melter technology being used in the Am/Cm vitrification research work is also under consideration for the stabilization of other actinides such as neptunium and plutonium. A series of melter drain tests were conducted at the Savannah River Technology Center to determine the relationship between the drain tube assembly operating variables and the resulting pour initiation times, glass flowrates, drain tube temperatures, and stop pour times. Performance criteria such as ability to start and stop pours in a controlled manner were also evaluated. The tests were also intended to provide support of oil modeling of drain tube performance predictions and thermal modeling of the drain tube and drain tube heater assembly. These drain tests were instrumental in the design of subsequent melter drain tube and drain tube heaters for the Am/Cm bushing melter, and therefore in the success of the Am/Cm vitrification and plutonium immobilization programs

  5. Plasma/arc melter review for vitrification of mixed wastes: Results

    Energy Technology Data Exchange (ETDEWEB)

    Eddy, T.L.; Soelberg, N.R.; Raivo, B.D. [MeltTran, Inc., Idaho Falls, ID (United States)

    1995-12-31

    In October of 1994, the Idaho Waste Treatment Program (IWTP) sponsored a workshop to review the results of a plasma/arc melter system preliminary design for treating mixed waste. Attention focused on (1) the melter design, (2) the offgas system design, and (3) the overall system design. The inclusion of feed preparation and handling systems, as well as monitoring and control systems, were considered premature until decisions regarding the melter and offgas treatment were resolved. The evaluation was based on the constraints of the transuranic-contaminated mixed waste in the Radioactive Waste Management Complex (RWMC) at the Idaho National Engineering Laboratory (INEL). Major factors are the retention of the transuranics in the basaltic slag, maintenance in a radioactive environment, reliability of components to prevent any major problems, upsets, or safety concerns, and the collection, elimination, or reduction of hazardous materials for appropriate stabilization. Several modifications were recommended by the group at large, discussed by the subcommittees, and accepted as the preferred options by the design team. Though all questions were not answered, the preferred systems for mixed waste treatment were the arc melters with graphite electrode systems with appropriate cooling which reduced maintenance and the possibility of eruptions that have occurred with plasma torches. Arc melters can also result in the minimum footprint and shielding. The preferred offgas systems were the wet/dry systems, that essentially eliminate the formation of carcinogenic compounds so they do not have to be destroyed down stream. This system also puts all of the particulate matter into one stream, instead of two.

  6. Application of electrical resistance tomography to glass melter

    International Nuclear Information System (INIS)

    Ichijo, Noriaki; Sakai, Taiji; Fujiwara, Hiroaki; Matsuno, Shinsuke; Misumi, Ryuta; Nishi, Kazuhiko; Kaminoyama, Meguru

    2015-01-01

    This paper describes the application of electrical resistance tomography (ERT) to glass melter to monitor the accumulation of the noble metals. To minimize the modification of the melter, existing structures such as thermowells and heating electrodes are used as electrodes of ERT, and the number of electrodes is much fewer than the conventional method. Therefore, Expanding Combination Data Acquisition method (ECDA) is developed and applies to the glass melter. ECDA method uses adjacent method and opposite method as a data acquisition and current injection electrodes are used as voltage measurement electrodes to increase the number of the data. In addition, conductivity images are reconstructed only near the wall to improve the resolution. As a result of applying to the glass melter, the conductivity change inside the melter caused by temperature can be monitored. Furthermore, lower voltage is measured in case of containing the noble metals inside the melter. Therefore, the potential as a monitoring method be confirmed. (author)

  7. Test Plan: Phase 1, Hanford LLW melter tests, GTS Duratek, Inc

    International Nuclear Information System (INIS)

    Eaton, W.C.

    1995-01-01

    This document provides a test plan for the conduct of vitrification testing by a vendor in support of the Hanford Tank Waste Remediation System (TWRS) Low-Level Waste (LLW) Vitrification Program. The vendor providing this test plan and conducting the work detailed within it [one of seven selected for glass melter testing under Purchase Order MMI-SVV-384215] is GTS Duratek, Inc., Columbia, Maryland. The GTS Duratek project manager for this work is J. Ruller. This test plan is for Phase I activities described in the above Purchase Order. Test conduct includes melting of glass with Hanford LLW Double-Shell Slurry Feed waste simulant in a DuraMelter trademark vitrification system

  8. The DWPF strategy for producing an acceptable product

    International Nuclear Information System (INIS)

    Goldston, W.T.; Plodinec, M.J.

    1991-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) will convert the 130 million liters of high-level nuclear waste at SRS into stable borosilicate glass. Production of canistered waste forms by the DWPF is scheduled to begin well before submission of the license application for the first repository. The Department of Energy has defined waste acceptance specifications to ensure that DWPF canistered waste forms will be acceptable for eventual disposal. To ensure that canistered waste forms meet those specifications, a program is being carried out to qualify the waste form and those aspects of the production process which affect product quality. This program includes: Pre-production qualification testing of simulated and actual waste forms; Disciplined demonstrations of the ability to produce an acceptable product during startup testing; and Application of a rigorous product control program during production

  9. The Impact Of The Mcu Life Extension Solvent On Dwpf Glass Formulation Efforts

    International Nuclear Information System (INIS)

    Peeler, D.; Edwards, T.

    2011-01-01

    As a part of the Actinide Removal Process (ARP)/Modular Caustic Side Solvent Extraction Unit (MCU) Life Extension Project, a next generation solvent (NG-CSSX), a new strip acid, and modified monosodium titanate (mMST) will be deployed. The strip acid will be changed from dilute nitric acid to dilute boric acid (0.01 M). Because of these changes, experimental testing with the next generation solvent and mMST is required to determine the impact of these changes in 512-S operations as well as Chemical Process Cell (CPC), Defense Waste Processing Facility (DWPF) glass formulation activities, and melter operations at DWPF. To support programmatic objectives, the downstream impacts of the boric acid strip effluent (SE) to the glass formulation activities and melter operations are considered in this study. More specifically, the impacts of boric acid additions to the projected SB7b operating windows, potential impacts to frit production temperatures, and the potential impact of boron volatility are evaluated. Although various boric acid molarities have been reported and discussed, the baseline flowsheet used to support this assessment was 0.01M boric acid. The results of the paper study assessment indicate that Frit 418 and Frit 418-7D are robust to the implementation of the 0.01M boric acid SE into the SB7b flowsheet (sludge-only or ARP-added). More specifically, the projected operating windows for the nominal SB7b projections remain essentially constant (i.e., 25-43 or 25-44% waste loading (WL)) regardless of the flowsheet options (sludge-only, ARP added, and/or the presence of the new SE). These results indicate that even if SE is not transferred to the Sludge Receipt and Adjustment Tank (SRAT), there would be no need to add boric acid (from a trim tank) to compositionally compensate for the absence of the boric acid SE in either a sludge-only or ARP-added SB7b flowsheet. With respect to boron volatility, the Measurement Acceptability Region (MAR) assessments also

  10. Can-in-canister cold demonstration in DWPF (U)

    International Nuclear Information System (INIS)

    Kuehn, N.H.

    1996-07-01

    The Department of Energy Fissile Materials Disposition Program is evaluating a number of options for disposition of weapons-usable plutonium surplus to national defense needs. One of the immobilization options is the Can-In-Canister approach. In this option small cans of a plutonium glass, which contains a neutron absorber, are placed on a support structure in a large Savannah River Site Defense Waste Processing Facility (DWPF) canister. The top is then welded onto the canister. This canister is filled with High Level Waste (HLW) glass at the DWPF. The HLW glass provides the radiation source for proliferation resistance. These canisters are to be placed in a Federal Repository. To provide information on the technical feasibility of this option prior to the Record of Decision on plutonium disposition, the Department of Energy Fissile Materials Disposition Program funded a demonstration in the DWPF. This demonstration was conducted before the start of radioactive operations. Two test canisters containing cans of surrogate (non- radioactive) plutonium glass were successfully filled with simulated HLW glass at the DWPF using standard pouring procedures. One canister had twenty cans of surrogate plutonium glass. The other had eight cans of surrogate plutonium glass. After the canisters were filled, the contents of the canisters were examined to provide data on the effect of the rack and cans on the filling of the DWPF canister, the effect of the pour on the surrogate plutonium glass and the effect of the rack and cans on the simulated HLW glass. There was no deformation of the support racks during the pour. The simulated HLW glass filled all the regions around the rack and cans and the regions between the cans and the wall of the canister. This report discusses the design of the racks and cans, the modification of the DWPF canisters to accommodate the rack and cans, the conditions during the pours and the results of the post pour analysis

  11. SME Acceptability Determination For DWPF Process Control (U)

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-12

    The statistical system described in this document is called the Product Composition Control System (PCCS). K. G. Brown and R. L. Postles were the originators and developers of this system as well as the authors of the first three versions of this technical basis document for PCCS. PCCS has guided acceptability decisions for the processing at the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) since the start of radioactive operations in 1996. The author of this revision to the document gratefully acknowledges the firm technical foundation that Brown and Postles established to support the ongoing successful operation at the DWPF. Their integration of the glass propertycomposition models, developed under the direction of C. M. Jantzen, into a coherent and robust control system, has served the DWPF well over the last 20+ years, even as new challenges, such as the introduction into the DWPF flowsheet of auxiliary streams from the Actinide Removal Process (ARP) and other processes, were met. The purpose of this revision is to provide a technical basis for modifications to PCCS required to support the introduction of waste streams from the Salt Waste Processing Facility (SWPF) into the DWPF flowsheet. An expanded glass composition region is anticipated by the introduction of waste streams from SWPF, and property-composition studies of that glass region have been conducted. Jantzen, once again, directed the development of glass property-composition models applicable for this expanded composition region. The author gratefully acknowledges the technical contributions of C.M. Jantzen leading to the development of these glass property-composition models. The integration of these models into the PCCS constraints necessary to administer future acceptability decisions for the processing at DWPF is provided by this sixth revision of this document.

  12. Letter report: Cold crucible melter assessment

    International Nuclear Information System (INIS)

    Elliott, M.L.

    1996-03-01

    One of the activities of the PNL Vitrification Technology Development (PVTD) Project is to assist the Tank Waste Remediation Systems (TWRS) Program in determining which melter systems should be performance tested for potential implementation in the high-level waste (HLW) vitrification plant. The Richland Operations Office (RL) has recommended that the Cold Crucible Melter (CCM) be evaluated as a candidate ''next generation'' melter. As a result, the CCM System Evaluation cost account was established under the PVTD Project so that the CCM could be initially assessed on a high-priority basis. This letter report summarizes a brief initial review and assessment of the CCM. Using the recommendations made in this document, Westinghouse Hanford Company (WHC) and RL will make a decision regarding the urgency of performance testing the CCM. If the decision is favorable, a subcontract will be negotiated for performance testing of a CCM using Hanford HLW simulants in a pilot-scale facility. Because of the aggressive nature of the schedule, the CCM evaluation was not rigorous. The evaluation consisted of a literature review and interviews with proponents of the technology during a recent trip to France. This letter report summarizes the evaluation and makes recommendations regarding further work in this area

  13. Rheological Characterization of Unusual DWPF Slurry Samples

    International Nuclear Information System (INIS)

    Koopman, D. C.

    2005-01-01

    A study was undertaken to identify and clarify examples of unusual rheological behavior in Defense Waste Processing Facility (DWPF) simulant slurry samples. Identification was accomplished by reviewing sludge, Sludge Receipt and Adjustment Tank (SRAT) product, and Slurry Mix Evaporator (SME) product simulant rheological results from the prior year. Clarification of unusual rheological behavior was achieved by developing and implementing new measurement techniques. Development of these new methods is covered in a separate report, WSRC-TR-2004-00334. This report includes a review of recent literature on unusual rheological behavior, followed by a summary of the rheological measurement results obtained on a set of unusual simulant samples. Shifts in rheological behavior of slurries as the wt. % total solids changed have been observed in numerous systems. The main finding of the experimental work was that the various unusual DWPF simulant slurry samples exhibit some degree of time dependent behavior. When a given shear rate is applied to a sample, the apparent viscosity of the slurry changes with time rather than remaining constant. These unusual simulant samples are more rheologically complex than Newtonian liquids or more simple slurries, neither of which shows significant time dependence. The study concludes that the unusual rheological behavior that has been observed is being caused by time dependent rheological properties in the slurries being measured. Most of the changes are due to the effect of time under shear, but SB3 SME products were also changing properties while stored in sample bottles. The most likely source of this shear-related time dependence for sludge is in the simulant preparation. More than a single source of time dependence was inferred for the simulant SME product slurries based on the range of phenomena observed. Rheological property changes were observed on the time-scale of a single measurement (minutes) as well as on a time scale of hours

  14. Physical and numerical modeling of Joule-heated melters

    Energy Technology Data Exchange (ETDEWEB)

    Eyler, L.L.; Skarda, R.J.; Crowder, R.S. III; Trent, D.S.; Reid, C.R.; Lessor, D.L.

    1985-10-01

    The Joule-heated ceramic-lined melter is an integral part of the high level waste immobilization process under development by the US Department of Energy. Scaleup and design of this waste glass melting furnace requires an understanding of the relationships between melting cavity design parameters and the furnace performance characteristics such as mixing, heat transfer, and electrical requirements. Developing empirical models of these relationships through actual melter testing with numerous designs would be a very costly and time consuming task. Additionally, the Pacific Northwest Laboratory (PNL) has been developing numerical models that simulate a Joule-heated melter for analyzing melter performance. This report documents the method used and results of this modeling effort. Numerical modeling results are compared with the more conventional, physical modeling results to validate the approach. Also included are the results of numerically simulating an operating research melter at PNL. Physical Joule-heated melters modeling results used for qualiying the simulation capabilities of the melter code included: (1) a melter with a single pair of electrodes and (2) a melter with a dual pair (two pairs) of electrodes. The physical model of the melter having two electrode pairs utilized a configuration with primary and secondary electrodes. The principal melter parameters (the ratio of power applied to each electrode pair, modeling fluid depth, electrode spacing) were varied in nine tests of the physical model during FY85. Code predictions were made for five of these tests. Voltage drops, temperature field data, and electric field data varied in their agreement with the physical modeling results, but in general were judged acceptable. 14 refs., 79 figs., 17 tabs.

  15. Physical and numerical modeling of Joule-heated melters

    International Nuclear Information System (INIS)

    Eyler, L.L.; Skarda, R.J.; Crowder, R.S. III; Trent, D.S.; Reid, C.R.; Lessor, D.L.

    1985-10-01

    The Joule-heated ceramic-lined melter is an integral part of the high level waste immobilization process under development by the US Department of Energy. Scaleup and design of this waste glass melting furnace requires an understanding of the relationships between melting cavity design parameters and the furnace performance characteristics such as mixing, heat transfer, and electrical requirements. Developing empirical models of these relationships through actual melter testing with numerous designs would be a very costly and time consuming task. Additionally, the Pacific Northwest Laboratory (PNL) has been developing numerical models that simulate a Joule-heated melter for analyzing melter performance. This report documents the method used and results of this modeling effort. Numerical modeling results are compared with the more conventional, physical modeling results to validate the approach. Also included are the results of numerically simulating an operating research melter at PNL. Physical Joule-heated melters modeling results used for qualiying the simulation capabilities of the melter code included: (1) a melter with a single pair of electrodes and (2) a melter with a dual pair (two pairs) of electrodes. The physical model of the melter having two electrode pairs utilized a configuration with primary and secondary electrodes. The principal melter parameters (the ratio of power applied to each electrode pair, modeling fluid depth, electrode spacing) were varied in nine tests of the physical model during FY85. Code predictions were made for five of these tests. Voltage drops, temperature field data, and electric field data varied in their agreement with the physical modeling results, but in general were judged acceptable. 14 refs., 79 figs., 17 tabs

  16. Sludge Washing and Demonstration of the DWPF Nitric/Formic Flowsheet in the SRNL Shielded Cells for Sludge Batch 9 Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Pareizs, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Newell, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Martino, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Johnson, F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-11-01

    Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) to qualify the next batch of sludge – Sludge Batch 9 (SB9). Current practice is to prepare sludge batches in Tank 51 by transferring sludge to Tank 51 from other tanks. The sludge is washed and transferred to Tank 40, the current Defense Waste Process Facility (DWPF) feed tank. Prior to sludge transfer from Tank 51 to Tank 40, the Tank 51 sludge must be qualified. SRNL qualifies the sludge in multiple steps. First, a Tank 51 sample is received, then characterized, washed, and again characterized. SRNL then demonstrates the DWPF Chemical Process Cell (CPC) flowsheet with the sludge. The final step of qualification involves chemical durability measurements of glass fabricated in the DWPF CPC demonstrations. In past sludge batches, SRNL had completed the DWPF demonstration with Tank 51 sludge. For SB9, SRNL has been requested to process a blend of Tank 51 and Tank 40 at a targeted ratio of 44% Tank 51 and 56% Tank 40 on an insoluble solids basis.

  17. Spray Calciner/In-Can Melter high-level waste solidification technical manual

    International Nuclear Information System (INIS)

    Larson, D.E.

    1980-09-01

    This technical manual summarizes process and equipment technology developed at Pacific Northwest Laboratory over the last 20 years for vitrification of high-level liquid waste by the Spray Calciner/In-Can Melter process. Pacific Northwest Laboratory experience includes process development and demonstration in laboratory-, pilot-, and full-scale equipment using nonradioactive synthetic wastes. Also, laboratory- and pilot-scale process demonstrations have been conducted using actual high-level radioactive wastes. In the course of process development, more than 26 tonnes of borosilicate glass have been produced in 75 canisters. Four of these canisters contained radioactive waste glass. The associated process and glass chemistry is discussed. Technology areas described include calciner feed treatment and techniques, calcination, vitrification, off-gas treatment, glass containment (the canister), and waste glass chemistry. Areas of optimization and site-specific development that would be needed to adapt this base technology for specific plant application are indicated. A conceptual Spray Calciner/In-Can Melter system design and analyses are provided in the manual to assist prospective users in evaluating the process for plant application, to provide equipment design information, and to supply information for safety analyses and environmental reports. The base (generic) technology for the Spray Calciner/In-Can Melter process has been developed to a point at which it is ready for plant application

  18. Hanford high-level waste melter system evaluation data packages

    International Nuclear Information System (INIS)

    Elliott, M.L.; Shafer, P.J.; Lamar, D.A.; Merrill, R.A.; Grunewald, W.; Roth, G.; Tobie, W.

    1996-03-01

    The Tank Waste Remediation System is selecting a reference melter system for the Hanford High-Level Waste vitrification plant. A melter evaluation was conducted in FY 1994 to narrow down the long list of potential melter technologies to a few for testing. A formal evaluation was performed by a Melter Selection Working Group (MSWG), which met in June and August 1994. At the June meeting, MSWG evaluated 15 technologies and selected six for more thorough evaluation at the Aug. meeting. All 6 were variations of joule-heated or induction-heated melters. Between the June and August meetings, Hanford site staff and consultants compiled data packages for each of the six melter technologies as well as variants of the baseline technologies. Information was solicited from melter candidate vendors to supplement existing information. This document contains the data packages compiled to provide background information to MSWG in support of the evaluation of the six technologies. (A separate evaluation was performed by Fluor Daniel, Inc. to identify balance of plant impacts if a given melter system was selected.)

  19. Melter operation results in chemical test at Rokkasho Reprocessing Plant

    International Nuclear Information System (INIS)

    Kanehira, Norio; Yoshioka, Masahiro; Muramoto, Hitoshi; Oba, Takaaki; Takahashi, Yuji

    2005-01-01

    Chemical Test of the glass melter system of the Vitrification Facility at Rokkasho Reprocessing Plant (RRP) was performed. In this test, basic performance of heating-up of the melter, melting glass, pouring glass was confirmed using simulated materials. Through these tests and operation of all modes, good results were gained, and training of operators was completed. (author)

  20. DC graphite plasma arc melter technology for waste vitrification

    International Nuclear Information System (INIS)

    Hamilton, R.A.; Wittle, J.K.; Trescot, J.; Wilver, P.

    1995-01-01

    This paper describes the features and benefits of a DC Arc Melter for the permanent treatment of all types of solid wastes including nonhazardous, hazardous and radioactive. This DC Arc Melter system is the low cost permanent solution for solid waste pollution prevention and remediation. Concern over the effective disposal of wastes generated by our industrial society, worldwide, has prompted development of technologies to address the problem. The only solution that has the ability to process almost all wastes, and to recover/recycle metallic and inorganic matter, is the group of technologies known as melters. Melters have distinct advantages over traditional technologies such as incineration because melters; operate at higher temperatures, are relatively unaffected by changes in the waste stream, produce a vitrified stable product, reduce gaseous emissions, and have the capability to recover/recycle slag, metals and gas. The system, DC Plasma Arc Melter, has the lowest capital, maintenance and operating cost of any melter technology because of its patented DC Plasma Arc with graphite electrode. DC Plasma Arc Melter systems are available in sizes from 50 kg/batch or 250-3,000 kg/hr on a continuous basis

  1. Estimation of Total Error in DWPF Reported Radionuclide Inventories

    International Nuclear Information System (INIS)

    Edwards, T.B.

    1995-01-01

    This report investigates the impact of random errors due to measurement and sampling on the reported concentrations of radionuclides in DWPF's filled canister inventory resulting from each macro-batch. The objective of this investigation is to estimate the variance of the total error in reporting these radionuclide concentrations

  2. DWPF PCCS version 2.0 test case

    International Nuclear Information System (INIS)

    Brown, K.G.; Pickett, M.A.

    1992-01-01

    To verify the operation of the Product Composition Control System (PCCS), a test case specific to DWPF operation was developed. The values and parameters necessary to demonstrate proper DWPF product composition control have been determined and are presented in this paper. If this control information (i.e., for transfers and analyses) is entered into the PCCS as illustrated in this paper, and the results obtained correspond to the independently-generated results, it can safely be said that the PCCS is operating correctly and can thus be used to control the DWPF. The independent results for this test case will be generated and enumerated in a future report. This test case was constructed along the lines of the normal DWPF operation. Many essential parameters are internal to the PCCS (e.g., property constraint and variance information) and can only be manipulated by personnel knowledgeable of the Symbolics reg-sign hardware and software. The validity of these parameters will rely on induction from observed PCCS results. Key process control values are entered into the PCCS as they would during normal operation. Examples of the screens used to input specific process control information are provided. These inputs should be entered into the PCCS database, and the results generated should be checked against the independent, computed results to confirm the validity of the PCCS

  3. Remotely replaceable jumpers and embedded wiring for the DWPF

    International Nuclear Information System (INIS)

    Heckendorn, F.M. II.

    1984-01-01

    The Defense Waste Processing Facility (DWPF) for radioactive waste vitrification at the Savannah River Plant (SRP) is now under construction. Development of specialized electrical/instrument inter-connectors, or jumpers, is now complete. Remote replacement of the associated through-wall wiring using a standard canyon crane has also been demonstrated. 8 figures

  4. Vitrification of Hanford wastes in a joule-heated ceramic melter and evaluation of resultant canisterized product

    International Nuclear Information System (INIS)

    Chapman, C.C.; Buelt, J.L.; Slate, S.C.; Katayama, Y.B.; Bunnell, L.R.

    1979-08-01

    Experience gained in the week-long vitrification test and characterization of the glass produced in the run support the following conclusions: The Hanford waste simulated in this test can be readily vitrified in a joule-heated ceramic melter. Physical properties of the molten glass were entirely compatible with melter operation. The average feed rate of 106 kg/h is high enough to make the ceramic melter a feasible piece of equipment for vitrifying Hanford wastes. The glass produced in this trial had good chemical durability, 6(10) -5 g/cm 2 -d. When one of the canisters was purposely dropped onto a steel pad, the damage was limited to deformation of the steel can in the impact area, cracking of a weld, and fracturing of glass in the immediate vicinity of the impact area. No glass was released from the canister as a result of the drop test. The results of this vitrification test support the technical feasibility of vitrifying Hanford wastes by means of a joule-heated ceramic melter. Surface area for large glass castings is equivalent to the mass median particle diameters between 4.27 cm (1.75 in.) and 8.91 cm (3.51 in.) even when allowed to cool rapidly by standing in ambient air. Large canisters (up to 0.91 m in dia) can be cast without large voids while standing in air if the fill rate is over 100 kg/h. 34 figures, 10 tables

  5. MASBAL: A computer program for predicting the composition of nuclear waste glass produced by a slurry-fed ceramic melter

    International Nuclear Information System (INIS)

    Reimus, P.W.

    1987-07-01

    This report is a user's manual for the MASBAL computer program. MASBAL's objectives are to predict the composition of nuclear waste glass produced by a slurry-fed ceramic melter based on a knowledge of process conditions; to generate simulated data that can be used to estimate the uncertainty in the predicted glass composition as a function of process uncertainties; and to generate simulated data that can be used to provide a measure of the inherent variability in the glass composition as a function of the inherent variability in the feed composition. These three capabilities are important to nuclear waste glass producers because there are constraints on the range of compositions that can be processed in a ceramic melter and on the range of compositions that will be acceptable for disposal in a geologic repository. MASBAL was developed specifically to simulate the operation of the West Valley Component Test system, a commercial-scale ceramic melter system that will process high-level nuclear wastes currently stored in underground tanks at the site of the Western New York Nuclear Services Center (near West Valley, New York). The program is flexible enough, however, to simulate any slurry-fed ceramic melter system. 4 refs., 16 figs., 5 tabs

  6. Cullet Manufacture Using the Cylindrical Induction Melter

    International Nuclear Information System (INIS)

    Miller, D. H.

    2000-01-01

    The base process for vitrification of the Am/Cm solution stored in F-canyon uses 25SrABS cullet as the glass former. A small portion of the cullet used in the SRTC development work was purchased from Corning while the majority was made in the 5 inch Cylindrical Induction Melter (CIM5). Task 1.01 of TTR-NMSS/SE-006, Additional Am-Cm Process Development Studies, requested that a process for the glass former (cullet) fabrication be specified. This report provides the process details for 25SrAB cullet production thereby satisfying Task 1.01

  7. Equipment experience in a radioactive LFCM [liquid-fed ceramic melter] vitrification facility

    International Nuclear Information System (INIS)

    Holton, L.K. Jr.; Dierks, R.D.; Sevigny, G.J.; Goles, R.W.; Surma, J.E.; Thomas, N.M.

    1986-11-01

    Since October 1984, the Pacific Northwest Laboratory (PNL) has operated a pilot-scale radioactive liquid-fed ceramic melter (RLFCM) vitrification process in shielded manipulator hot cells. This vitrification facility is being operated for the Department of Energy (DOE) to remotely test vitrification equipment components in a radioactive environment and to develop design and operation data that can be applied to production-scale projects. This paper summarizes equipment and process experience obtained from the operations of equipment systems for waste feeding, waste vitrification, canister filling, canister handling, and vitrification off-gas treatment

  8. Energy Efficient Glass Melting - The Next Generation Melter

    Energy Technology Data Exchange (ETDEWEB)

    David Rue

    2008-03-01

    The objective of this project is to demonstrate a high intensity glass melter, based on the submerged combustion melting technology. This melter will serve as the melting and homogenization section of a segmented, lower-capital cost, energy-efficient Next Generation Glass Melting System (NGMS). After this project, the melter will be ready to move toward commercial trials for some glasses needing little refining (fiberglass, etc.). For other glasses, a second project Phase or glass industry research is anticipated to develop the fining stage of the NGMS process.

  9. Rheological Properties of Defense Waste processing Facility Melter Feeds

    International Nuclear Information System (INIS)

    Ebadian, M.A.; Mao, F.

    1998-01-01

    In the present investigation, viscosity measurements have been carried out for two types of simulated Defense waste slurries, a Savannah River slurry and a Hanford slurry. The measurements were conducted in two experimental options. A rotational viscometer was used to measure viscosity under well-defined temperature and pH value operating conditions. The solids concentration used for this option was lower than 15 wt.%. Both the slurries have been investigated using this experimental option. The Savannah River slurry has also been investigated in a pipeline flow system, which measured the pressure drop as the slurry flowed through the pipe. The slurry's viscosity can be extracted from the pressure drop information. These investigations have been performed in relatively wide parameter ranges. The solids concentration of the slurry tested in the pipeline system was as high as 25 wt.%.The slurry pH in both experimental options covered a range of 4 to 13.5. The highest operating temperature was 66 C for the rotational viscometer and 55 C for the pipeline system. In FY97, the experiments for the Hanford slurry in the pipeline system will be performed

  10. Translating DWPF design criteria into an engineered facility design

    International Nuclear Information System (INIS)

    Kemp, J.B.

    1986-01-01

    The Defense Waste Processing Facility (DWPF) takes radioactive defense waste sludge and the radioactive nuclides, cesium and strontium, from the salt solution, and incorporates them in borosilicate glass in stainless steel canisters, for subsequent disposal in a deep geologic repository. The facility was designed by Bechtel National, Inc. under a subcontract from E.I. DuPont de Nemurs and Co., the prime contractor for the Department of Energy, for the design, construction and commissioning of the plant. The design criteria were specified by the DuPont Company, based upon their extensive experience as designer, and operator since the early 1950's, of the existing Savannah River Plant facilities. Some of the design criteria imposed unusual or new requirements on the detailed design of the facilities. This paper describes some of these criteria, encompassing several engineering disciplines, and discusses the solutions and designs which were developed for the DWPF

  11. Hydrogen generation during treatment of simulated high-level radioactive waste with formic acid

    International Nuclear Information System (INIS)

    Ritter, J.A.; Zamecnik, J.R.; Hsu, C.W.

    1992-01-01

    The Integrated Defense Waste Processing Facility (DWPF) Melter System (IDMS), operated by the Savannah River Laboratory, is a one-fifth scale pilot facility used in support of the start-up and operation of the Department of Energy's DWPF. Five IDMS runs determined the effect of the presence of noble metals in HLW sludge on the H 2 generation rate during the preparation of melter feed with formic acid. Overall, the results clearly showed that H 2 generation in the DWPF SRAT could, at times, exceed the lower flammable limit of H 2 in air (4 vol%), depending on such factors as offgas generation and air inleakage of the DWPF vessels. Therefore, the installation of a forced air purge system and H 2 monitors were recommended to the DWPF to control the generation of H 2 during melter feed preparation by fuel dilution

  12. Modified IRC bench-scale arc melter for waste processing

    International Nuclear Information System (INIS)

    Eddy, T.L.; Sears, J.W.; Grandy, J.D.; Kong, P.C.; Watkins, A.D.

    1994-03-01

    This report describes the INEL Research Center (IRC) arc melter facility and its recent modifications. The arc melter can now be used to study volatilization of toxic and high vapor pressure metals and the effects of reducing and oxidizing (redox) states in the melt. The modifications include adding an auger feeder, a gas flow control and monitoring system, an offgas sampling and exhaust system, and a baghouse filter system, as well as improving the electrode drive, slag sampling system, temperature measurement and video monitoring and recording methods, and oxidation lance. In addition to the volatilization and redox studies, the arc melter facility has been used to produce a variety of glass/ceramic waste forms for property evaluation. Waste forms can be produced on a daily basis. Some of the melts performed are described to illustrate the melter's operating characteristics

  13. High-Level Waste Melter Study Report

    Energy Technology Data Exchange (ETDEWEB)

    Perez, Joseph M.; Bickford, Dennis F.; Day, Delbert E.; Kim, Dong-Sang; Lambert, Steven L.; Marra, Sharon L.; Peeler, David K.; Strachan, Denis M.; Triplett, Mark B.; Vienna, John D.; Wittman, Richard S.

    2001-07-13

    At the Hanford Site in Richland, Washington, the path to site cleanup involves vitrification of the majority of the wastes that currently reside in large underground tanks. A Joule-heated glass melter is the equipment of choice for vitrifying the high-level fraction of these wastes. Even though this technology has general national and international acceptance, opportunities may exist to improve or change the technology to reduce the enormous cost of accomplishing the mission of site cleanup. Consequently, the U.S. Department of Energy requested the staff of the Tanks Focus Area to review immobilization technologies, waste forms, and modifications to requirements for solidification of the high-level waste fraction at Hanford to determine what aspects could affect cost reductions with reasonable long-term risk. The results of this study are summarized in this report.

  14. Next Generation Melter Optioneering Study - Interim Report

    International Nuclear Information System (INIS)

    Gray, M.F.; Calmus, R.B.; Ramsey, G.; Lomax, J.; Allen, H.

    2010-01-01

    The next generation melter (NOM) development program includes a down selection process to aid in determining the recommended vitrification technology to implement into the WTP at the first melter change-out which is scheduled for 2025. This optioneering study presents a structured value engineering process to establish and assess evaluation criteria that will be incorporated into the down selection process. This process establishes an evaluation framework that will be used progressively throughout the NGM program, and as such this interim report will be updated on a regular basis. The workshop objectives were achieved. In particular: (1) Consensus was reached with stakeholders and technology providers represented at the workshop regarding the need for a decision making process and the application of the D 2 0 process to NGM option evaluation. (2) A framework was established for applying the decision making process to technology development and evaluation between 2010 and 2013. (3) The criteria for the initial evaluation in 2011 were refined and agreed with stakeholders and technology providers. (4) The technology providers have the guidance required to produce data/information to support the next phase of the evaluation process. In some cases it may be necessary to reflect the data/information requirements and overall approach to the evaluation of technology options against specific criteria within updated Statements of Work for 2010-2011. Access to the WTP engineering data has been identified as being very important for option development and evaluation due to the interface issues for the NGM and surrounding plant. WRPS efforts are ongoing to establish precisely data that is required and how to resolve this Issue. It is intended to apply a similarly structured decision making process to the development and evaluation of LAW NGM options.

  15. Numerical analysis of historical change of the electric resistance in the TVF glass melter

    International Nuclear Information System (INIS)

    Kawamura, Takumi; Sakai, Takaaki

    2004-09-01

    Concerning to the TVF glass melter in the Tokai reprocessing center, it is being planned to detect the deposition of the noble metals in a glass melter and remove them periodically to extend the melter lifetime. Numerical analysis has been performed for the electric resistance evaluation in order to estimate the sedimentation situation and current density distribution from the melter resistance. Electric field analysis was carried out by using MAGNA-FIM code and the influence factors to melter resistance was evaluated concerning to the sedimentation situation and glass temperature. In addition, transitions of the sedimentation and melter resistances were estimated from the operation history of the TVF-1 melter. As a result, the followings were obtained. From the evaluation of the influence factors to melter resistance, it turns out that the volume and the noble metals concentration of a sediment influence notably to melter resistance when the sediment contacts to electrodes. The sediment temperature at the melter bottom has small sensitivity in case of the non-contact situation. The glass temperature in the melter upper part, however, has big sensitivity in melter resistance irrespective of the existence of contact. Based on the above sensitivity evaluation, Numerical analysis was carried out supposing the sedimentation process which suits to a melter resistance fall during the operation history of the TVF-1 melter. As input conditions, the voltage between electrodes and the temperature in the melter were referred from the operation history data. It was assumed that the noble metals concentration in a sediment increased constantly for every operation batch. As a result, the characteristics of melter resistance history was reproduced successfully in general. Thereby, it became prospective to predict the sedimentation situation by using the new resistance analysis model for the glass melter. (author)

  16. U.S. Bureau of Mines, phase I Hanford low-level waste melter tests: Melter offgas report

    International Nuclear Information System (INIS)

    Eaton, W.C.

    1995-01-01

    A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Phase 1 of the melter demonstration tests using simulated LLW was completed during fiscal year 1995. This document is the melter offgas report on testing performed by the U.S. Department of the Interior, Bureau of Mines, Albany Research Center in Albany, Oregon. The Bureau of Mines (one of the seven vendors selected) was chosen to demonstrate carbon electrode melter technology (also called carbon arc or electric arc) under WHC subcontract number MMI-SVV-384216. The document contains the complete offgas report for the first 24-hour melter test (WHC-1) as prepared by Entropy Inc. A summary of this report is also contained in the''U.S. Bureau of Mines, Phase 1 Hanford Low-Level Waste Melter Tests: Final Report'' (WHC-SD-WM-VI-030)

  17. Statistical process control applied to the liquid-fed ceramic melter process

    International Nuclear Information System (INIS)

    Pulsipher, B.A.; Kuhn, W.L.

    1987-09-01

    In this report, an application of control charts to the apparent feed composition of a Liquid-Fed Ceramic Melter (LFCM) is demonstrated by using results from a simulation of the LFCM system. Usual applications of control charts require the assumption of uncorrelated observations over time. This assumption is violated in the LFCM system because of the heels left in tanks from previous batches. Methods for dealing with this problem have been developed to create control charts for individual batches sent to the feed preparation tank (FPT). These control charts are capable of detecting changes in the process average as well as changes in the process variation. All numbers reported in this document were derived from a simulated demonstration of a plausible LFCM system. In practice, site-specific data must be used as input to a simulation tailored to that site. These data directly affect all variance estimates used to develop control charts. 64 refs., 3 figs., 2 tabs

  18. Estimation of total error in DWPF reported radionuclide inventories. Revision 1

    International Nuclear Information System (INIS)

    Edwards, T.B.

    1995-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site is required to determine and report the radionuclide inventory of its glass product. For each macro-batch, the DWPF will report both the total amount (in curies) of each reportable radionuclide and the average concentration (in curies/gram of glass) of each reportable radionuclide. The DWPF is to provide the estimated error of these reported values of its radionuclide inventory as well. The objective of this document is to provide a framework for determining the estimated error in DWPF's reporting of these radionuclide inventories. This report investigates the impact of random errors due to measurement and sampling on the total amount of each reportable radionuclide in a given macro-batch. In addition, the impact of these measurement and sampling errors and process variation are evaluated to determine the uncertainty in the reported average concentrations of radionuclides in DWPF's filled canister inventory resulting from each macro-batch

  19. Nitric-glycolic flowsheet reduction/oxidation (redox) model for the defense waste processing facility (DWPF)

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Trivelpiece, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Ramsey, W. G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-14

    Control of the REDuction/OXidation (REDOX) state of glasses containing high concentrations of transition metals, such as High Level Waste (HLW) glasses, is critical in order to eliminate processing difficulties caused by overly reduced or overly oxidized melts. Operation of a HLW melter at Fe+2/ΣFe ratios of between 0.09 and 0.33, retains radionuclides in the melt and thus the final glass. Specifically, long-lived radioactive 99Tc species are less volatile in the reduced Tc4+ state as TcO2 than as NaTcO4 or Tc2O7, and ruthenium radionuclides in the reduced Ru4+ state are insoluble RuO2 in the melt which are not as volatile as NaRuO4 where the Ru is in the +7 oxidation state. Similarly, hazardous volatile Cr6+ occurs in oxidized melt pools as Na2CrO4 or Na2Cr2O7, while the Cr+3 state is less volatile and remains in the melt as NaCrO2 or precipitates as chrome rich spinels. The melter REDOX control balances the oxidants and reductants from the feed and from processing additives such as antifoam.

  20. Review of continuous ceramic-lined melter and associated experience at PNL

    International Nuclear Information System (INIS)

    Buelt, J.L.; Chapman, C.C.; Barnes, S.M.; Dierks, R.D.

    1979-01-01

    Development of continuous, ceramic-lined melters applicable to immobilization of radioactive wastes began at PNL in 1973. A comprehensive program is curretly in progress. The melters constructed at PNL have incorporated remote and reliable design features necessary for radioactive use. The extensive experience with vitrification of simulated wastes has proven the continuous melter's applicability to radioactive waste immobilization

  1. Graphite electrode arc melter demonstration Phase 2 test results

    International Nuclear Information System (INIS)

    Soelberg, N.R.; Chambers, A.G.; Anderson, G.L.; O'Connor, W.K.; Oden, L.L.; Turner, P.C.

    1996-06-01

    Several U.S. Department of Energy organizations and the U.S. Bureau of Mines have been collaboratively conducting mixed waste treatment process demonstration testing on the near full-scale graphite electrode submerged arc melter system at the Bureau's Albany (Oregon) Research Center. An initial test series successfully demonstrated arc melter capability for treating surrogate incinerator ash of buried mixed wastes with soil. The conceptual treatment process for that test series assumed that buried waste would be retrieved and incinerated, and that the incinerator ash would be vitrified in an arc melter. This report presents results from a recently completed second series of tests, undertaken to determine the ability of the arc melter system to stably process a wide range of open-quotes as-receivedclose quotes heterogeneous solid mixed wastes containing high levels of organics, representative of the wastes buried and stored at the Idaho National Engineering Laboratory (INEL). The Phase 2 demonstration test results indicate that an arc melter system is capable of directly processing these wastes and could enable elimination of an up-front incineration step in the conceptual treatment process

  2. Graphite electrode arc melter demonstration Phase 2 test results

    Energy Technology Data Exchange (ETDEWEB)

    Soelberg, N.R.; Chambers, A.G.; Anderson, G.L.; O`Connor, W.K.; Oden, L.L.; Turner, P.C.

    1996-06-01

    Several U.S. Department of Energy organizations and the U.S. Bureau of Mines have been collaboratively conducting mixed waste treatment process demonstration testing on the near full-scale graphite electrode submerged arc melter system at the Bureau`s Albany (Oregon) Research Center. An initial test series successfully demonstrated arc melter capability for treating surrogate incinerator ash of buried mixed wastes with soil. The conceptual treatment process for that test series assumed that buried waste would be retrieved and incinerated, and that the incinerator ash would be vitrified in an arc melter. This report presents results from a recently completed second series of tests, undertaken to determine the ability of the arc melter system to stably process a wide range of {open_quotes}as-received{close_quotes} heterogeneous solid mixed wastes containing high levels of organics, representative of the wastes buried and stored at the Idaho National Engineering Laboratory (INEL). The Phase 2 demonstration test results indicate that an arc melter system is capable of directly processing these wastes and could enable elimination of an up-front incineration step in the conceptual treatment process.

  3. Heat Transfer Model of a Small-Scale Waste Glass Melter with Cold Cap Layer

    Energy Technology Data Exchange (ETDEWEB)

    Abboud, Alexander; Guillen, Donna Post; Pokorny, Richard

    2016-09-01

    At the Hanford site in the state of Washington, more than 56 million gallons of radioactive waste is stored in underground tanks. The cleanup plan for this waste is vitrification at the Waste Treatment Plant (WTP), currently under construction. At the WTP, the waste will be blended with glass-forming materials and heated to 1423K, then poured into stainless steel canisters to cool and solidify. A fundamental understanding of the glass batch melting process is needed to optimize the process to reduce cost and decrease the life cycle of the cleanup effort. The cold cap layer that floats on the surface of the glass melt is the primary reaction zone for the feed-to-glass conversion. The conversion reactions include water release, melting of salts, evolution of batch gases, dissolution of quartz and the formation of molten glass. Obtaining efficient heat transfer to this region is crucial to achieving high rates of glass conversion. Computational fluid dynamics (CFD) modeling is being used to understand the heat transfer dynamics of the system and provide insight to optimize the process. A CFD model was developed to simulate the DM1200, a pilot-scale melter that has been extensively tested by the Vitreous State Laboratory (VSL). Electrodes are built into the melter to provide Joule heating to the molten glass. To promote heat transfer from the molten glass into the reactive cold cap layer, bubbling of the molten glass is used to stimulate forced convection within the melt pool. A three-phase volume of fluid approach is utilized to model the system, wherein the molten glass and cold cap regions are modeled as separate liquid phases, and the bubbling gas and plenum regions are modeled as one lumped gas phase. The modeling of the entire system with a volume of fluid model allows for the prescription of physical properties on a per-phase basis. The molten glass phase and the gas phase physical properties are obtained from previous experimental work. Finding representative

  4. Arc melter demonstration baseline test results

    International Nuclear Information System (INIS)

    Soelberg, N.R.; Chambers, A.G.; Anderson, G.L.; Oden, L.L.; O'Connor, W.K.; Turner, P.C.

    1994-07-01

    This report describes the test results and evaluation for the Phase 1 (baseline) arc melter vitrification test series conducted for the Buried Waste Integrated Demonstration program (BWID). Phase 1 tests were conducted on surrogate mixtures of as-incinerated wastes and soil. Some buried wastes, soils, and stored wastes at the INEL and other DOE sites, are contaminated with transuranic (TRU) radionuclides and hazardous organics and metals. The high temperature environment in an electric arc furnace may be used to process these wastes to produce materials suitable for final disposal. An electric arc furnace system can treat heterogeneous wastes and contaminated soils by (a) dissolving and retaining TRU elements and selected toxic metals as oxides in the slag phase, (b) destroying organic materials by dissociation, pyrolyzation, and combustion, and (c) capturing separated volatilized metals in the offgas system for further treatment. Structural metals in the waste may be melted and tapped separately for recycle or disposal, or these metals may be oxidized and dissolved into the slag. The molten slag, after cooling, will provide a glass/ceramic final waste form that is homogeneous, highly nonleachable, and extremely durable. These features make this waste form suitable for immobilization of TRU radionuclides and toxic metals for geologic timeframes. Further, the volume of contaminated wastes and soils will be substantially reduced in the process

  5. Review of Statistical Analyses Resulting from Performance of HLDWD-DWPF-005

    International Nuclear Information System (INIS)

    Beck, R.S.

    1997-01-01

    The Engineering Department at the Defense Waste Processing Facility (DWPF) has reviewed two reports from the Statistical Consulting Section (SCS) involving the statistical analysis of test results for analysis of small sample inserts (references 1 ampersand 2). The test results cover two proposed analytical methods, a room temperature hydrofluoric acid preparation (Cold Chem) and a sodium peroxide/sodium hydroxide fusion modified for insert samples (Modified Fusion). The reports support implementation of the proposed small sample containers and analytical methods at DWPF. Hydragard sampler valve performance was typical of previous results (reference 3). Using an element from each major feed stream. lithium from the frit and iron from the sludge, the sampler was determined to deliver a uniform mixture in either sample container.The lithium to iron ratios were equivalent for the standard 15 ml vial and the 3 ml insert.The proposed method provide equivalent analyses as compared to the current methods. The biases associated with the proposed methods on a vitrified basis are less than 5% for major elements. The sum of oxides for the proposed method compares favorably with the sum of oxides for the conventional methods. However, the average sum of oxides for the Cold Chem method was 94.3% which is below the minimum required recovery of 95%. Both proposed methods, cold Chem and Modified Fusion, will be required at first to provide an accurate analysis which will routinely meet the 95% and 105% average sum of oxides limit for Product Composition Control System (PCCS).Issued to be resolved during phased implementation are as follows: (1) Determine calcine/vitrification factor for radioactive feed; (2) Evaluate covariance matrix change against process operating ranges to determine optimum sample size; (3) Evaluate sources for low sum of oxides; and (4) Improve remote operability of production versions of equipment and instruments for installation in 221-S.The specifics of

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

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-04-30

    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.

  7. SLUDGE WASHING AND DEMONSTRATION OF THE DWPF FLOWSHEET IN THE SRNL SHIELDED CELLS FOR SLUDGE BATCH 7A QUALIFICATION

    Energy Technology Data Exchange (ETDEWEB)

    Pareizs, J.; Billings, A.; Click, D.

    2011-07-08

    Waste Solidification Engineering (WSE) has requested that characterization and a radioactive demonstration of the next batch of sludge slurry (Sludge Batch 7a*) be completed in the Shielded Cells Facility of the Savannah River National Laboratory (SRNL) via a Technical Task Request (TTR). This characterization and demonstration, or sludge batch qualification process, is required prior to transfer of the sludge from Tank 51 to the Defense Waste Processing Facility (DWPF) feed tank (Tank 40). The current WSE practice is to prepare sludge batches in Tank 51 by transferring sludge from other tanks. Discharges of nuclear materials from H Canyon are often added to Tank 51 during sludge batch preparation. The sludge is washed and transferred to Tank 40, the current DWPF feed tank. Prior to transfer of Tank 51 to Tank 40, SRNL simulates the Tank Farm and DWPF processes with a Tank 51 sample (referred to as the qualification sample). Sludge Batch 7a (SB7a) is composed of portions of Tanks 4, 7, and 12; the Sludge Batch 6 heel in Tank 51; and a plutonium stream from H Canyon. SRNL received the Tank 51 qualification sample (sample ID HTF-51-10-125) following sludge additions to Tank 51. This report documents: (1) The washing (addition of water to dilute the sludge supernate) and concentration (decanting of supernate) of the SB7a - Tank 51 qualification sample to adjust sodium content and weight percent insoluble solids to Tank Farm projections. (2) The performance of a DWPF Chemical Process Cell (CPC) simulation using the washed Tank 51 sample. The simulation included a Sludge Receipt and Adjustment Tank (SRAT) cycle, where acid was added to the sludge to destroy nitrite and reduce mercury, and a Slurry Mix Evaporator (SME) cycle, where glass frit was added to the sludge in preparation for vitrification. The SME cycle also included replication of five canister decontamination additions and concentrations. Processing parameters were based on work with a non

  8. Gaseous and particulate emissions from a DC arc melter.

    Science.gov (United States)

    Overcamp, Thomas J; Speer, Matthew P; Griner, Stewart J; Cash, Douglas M

    2003-01-01

    Tests treating soils contaminated with metal compounds and radionuclide surrogates were conducted in a DC arc melter. The soil melted, and glassy or ceramic waste forms with a separate metal phase were produced. Tests were run in the melter plenum with either air or N2 purge gases. In addition to nitrogen, the primary emissions of gases were CO2, CO, oxygen, methane, and oxides of nitrogen (NO(x)). Although the gas flow through the melter was low, the particulate concentrations ranged from 32 to 145 g/m3. Cerium, a nonradioactive surrogate for plutonium and uranium, was not enriched in the particulate matter (PM). The PM was enriched in cesium and highly enriched in lead.

  9. Burst Test Qualification Analysis of DWPF Canister-Plug Weld

    International Nuclear Information System (INIS)

    Gupta, N.K.; Gong, Chung.

    1995-02-01

    The DWPF canister closure system uses resistance welding for sealing the canister nozzle and plug to ensure leak tightness. The welding group at SRTC is using the burst test to qualify this seal weld in lieu of the shear test in ASME B ampersand PV Code, Section IX, paragraph QW-196. The burst test is considered simpler and more appropriate than the shear test for this application. Although the geometry, loading and boundary conditions are quite different in the two tests, structural analyses show similarity in the failure mode of the shear test in paragraph QW-196 and the burst test on the DWPF canister nozzle Non-linear structural analyses are performed using finite element techniques to study the failure mode of the two tests. Actual test geometry and realistic stress strain data for the 304L stainless steel and the weld material are used in the analyses. The finite element models are loaded until failure strains are reached. The failure modes in both tests are shear at the failure points. Based on these observations, it is concluded that the use of a burst test in lieu of the shear test for qualifying the canister-plug weld is acceptable. The burst test analysis for the canister-plug also yields the burst pressures which compare favorably with the actual pressure found during burst tests. Thus, the analysis also provides an estimate of the safety margins in the design of these vessels

  10. Lot No. 1 of Frit 202 for DWPF cold runs

    International Nuclear Information System (INIS)

    Schumacher, R.F.

    1993-01-01

    This report was prepared at the end of 1992 and summarizes the evaluation of the first lot sample of DWPF Frit 202 from Cataphote Inc. Publication of this report was delayed until the results from the carbon analyses could be included. To avoid confusion the frit specifications presented in this report were those available at the end of 1992. The specifications were slightly modified early in 1993. The frit was received and evaluated for moisture, particle size distribution, organic-inorganic carbon and chemical composition. Moisture content and particle size distribution were determined on a representative sample at SRTC. These properties were within the DWPF specifications for Frit 202. A representative sample was submitted to Corning Engineering Laboratory Services for chemical analyses. The sample was split and two dissolutions prepared. Each dissolution was analyzed on two separate days. The results indicate that there is a high probability (>95%) that the silica content of this frit is below the specification limit of 77.0 ± 1.0 wt %. The average of the four analyzed values was 75.1 wt % with a standard deviation of 0.28 wt %. All other oxides were within the elliptical two sigma limits. Control standard frit samples were submitted and analyzed at the same time and the results were very similar to previous analyses of these materials

  11. ISOLOK VALVE ACCEPTANCE TESTING FOR DWPF SME SAMPLING PROCESS

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, T.; Hera, K.; Coleman, C.; Jones, M.; Wiedenman, B.

    2011-12-05

    Evaluation of the Defense Waste Processing Facility (DWPF) Chemical Process Cell (CPC) cycle time identified several opportunities to improve the CPC processing time. Of the opportunities, a focus area related to optimizing the equipment and efficiency of the sample turnaround time for DWPF Analytical Laboratory was identified. The Mechanical Systems & Custom Equipment Development (MS&CED) Section of the Savannah River National Laboratory (SRNL) evaluated the possibility of using an Isolok{reg_sign} sampling valve as an alternative to the Hydragard{reg_sign} valve for taking process samples. Previous viability testing was conducted with favorable results using the Isolok sampler and reported in SRNL-STI-2010-00749 (1). This task has the potential to improve operability, reduce maintenance time and decrease CPC cycle time. This report summarizes the results from acceptance testing which was requested in Task Technical Request (TTR) HLW-DWPF-TTR-2010-0036 (2) and which was conducted as outlined in Task Technical and Quality Assurance Plan (TTQAP) SRNL-RP-2011-00145 (3). The Isolok to be tested is the same model which was tested, qualified, and installed in the Sludge Receipt Adjustment Tank (SRAT) sample system. RW-0333P QA requirements apply to this task. This task was to qualify the Isolok sampler for use in the DWPF Slurry Mix Evaporator (SME) sampling process. The Hydragard, which is the current baseline sampling method, was used for comparison to the Isolok sampling data. The Isolok sampler is an air powered grab sampler used to 'pull' a sample volume from a process line. The operation of the sampler is shown in Figure 1. The image on the left shows the Isolok's spool extended into the process line and the image on the right shows the sampler retracted and then dispensing the liquid into the sampling container. To determine tank homogeneity, a Coliwasa sampler was used to grab samples at a high and low location within the mixing tank. Data from

  12. Savannah River Laboratory's operating experience with glass melters

    International Nuclear Information System (INIS)

    Brown, F.H.; Randall, C.T.; Cosper, M.B.; Moseley, J.P.

    1982-01-01

    The Department of Energy, with recommendations from the Du Pont Company, is proposing that a Defense Waste Processing Facility be constructed at the Savannah River Plant to immobilize radioactive The immobilization process is designed around the solidification of waste sludge in borosilicate glass. The Savannah River Laboratory, who is responsible for the solidification process development program, has completed an experimental program with one large-scale glass melter and just started up another melter. Experimental data indicate that process requirements can easily be met with the current design. 7 figures

  13. DWPF glass transition temperatures: What they are and why they are important

    International Nuclear Information System (INIS)

    Marra, S.L.; Jantzen, C.M.; Ramsey, A.A.

    1991-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site will immobilize high-level radioactive liquid waste in borosilicate glass. The glass will be poured into stainless steel canisters for eventual disposal in a geologic repository. The Department of Energy has defined a set of requirements for the DWPF canistered waste form which must be met in order to assure compatibility with, and acceptance by, the repository. These requirements are the Waste Acceptance Preliminary Specifications (WAPS). The WAPS require DWPF to report glass transition temperatures for the projected range of compositions. This information will be used by the repository to establish waste package design limits

  14. Design and operation of small-scale glass melters for immobilizing radioactive waste

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Chismar, P.H.

    1980-01-01

    A small-scale (3-kg), joule-heated, continuous melter has been designed to study vitrification of Savannah River Plant radioactive waste. The first melter built has been in nonradioactive service for nearly three years. This melter had Inconel 690 electrodes and uses Monofrax K-3 for the contact refractory. Several problems seem in this melter have had an impact on the design of a full-scale system. Problems include uncontrolled electric currents passing through the throat, and formation of a slag layer at the bottom of the melter. The performance of a similar melter in a low-maintenance, radioactive environment is also described. Problems such as halide refluxing, and hot streaking, first observed in this melter, are also discussed

  15. DM100 AND DM1200 MELTER TESTING WITH HIGH WASTE LOADING GLASS FORMULATIONS FOR HANFORD HIGH-ALUMINUM HLW STREAMS

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; KOT WK; PEGG IL; JOSEPH I

    2009-12-30

    This Test Plan describes work to support the development and testing of high waste loading glass formulations that achieve high glass melting rates for Hanford high aluminum high level waste (HLW). In particular, the present testing is designed to evaluate the effect of using low activity waste (LAW) waste streams as a source of sodium in place ofchemical additives, sugar or cellulose as a reductant, boehmite as an aluminum source, and further enhancements to waste processing rate while meeting all processing and product quality requirements. The work will include preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM 100) tests designed to examine the effects of enhanced glass formulations, glass processing temperature, incorporation of the LAW waste stream as a sodium source, type of organic reductant, and feed solids content on waste processing rate and product quality. Also included is a confirmatory test on the HLW Pilot Melter (DM1200) with a composition selected from those tested on the DM100. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy's (DOE's) Office of River Protection (ORP) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same waste composition. This Test Plan is prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is about 12,500. This estimate is based upon the inventory ofthe tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat

  16. ELIMINATION OF THE CHARACTERIZATION OF DWPF POUR STREAM SAMPLE AND THE GLASS FABRICATION AND TESTING OF THE DWPF SLUDGE BATCH QUALIFICATION SAMPLE

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J.; Peeler, D.; Edwards, T.

    2012-05-11

    A recommendation to eliminate all characterization of pour stream glass samples and the glass fabrication and Product Consistency Test (PCT) of the sludge batch qualification sample was made by a Six-Sigma team chartered to eliminate non-value-added activities for the Defense Waste Processing Facility (DWPF) sludge batch qualification program and is documented in the report SS-PIP-2006-00030. That recommendation was supported through a technical data review by the Savannah River National Laboratory (SRNL) and is documented in the memorandums SRNL-PSE-2007-00079 and SRNL-PSE-2007-00080. At the time of writing those memorandums, the DWPF was processing sludge-only waste but, has since transitioned to a coupled operation (sludge and salt). The SRNL was recently tasked to perform a similar data review relevant to coupled operations and re-evaluate the previous recommendations. This report evaluates the validity of eliminating the characterization of pour stream glass samples and the glass fabrication and Product Consistency Test (PCT) of the sludge batch qualification samples based on sludge-only and coupled operations. The pour stream sample has confirmed the DWPF's ability to produce an acceptable waste form from Slurry Mix Evaporator (SME) blending and product composition/durability predictions for the previous sixteen years but, ultimately the pour stream analysis has added minimal value to the DWPF's waste qualification strategy. Similarly, the information gained from the glass fabrication and PCT of the sludge batch qualification sample was determined to add minimal value to the waste qualification strategy since that sample is routinely not representative of the waste composition ultimately processed at the DWPF due to blending and salt processing considerations. Moreover, the qualification process has repeatedly confirmed minimal differences in glass behavior from actual radioactive waste to glasses fabricated from simulants or batch chemicals. In

  17. Elimination Of The Characterization Of DWPF Pour Stream Sample And The Glass Fabrication And Testing Of The DWPF Sludge Batch Qualification Sample

    International Nuclear Information System (INIS)

    Amoroso, J.; Peeler, D.; Edwards, T.

    2012-01-01

    A recommendation to eliminate all characterization of pour stream glass samples and the glass fabrication and Product Consistency Test (PCT) of the sludge batch qualification sample was made by a Six-Sigma team chartered to eliminate non-value-added activities for the Defense Waste Processing Facility (DWPF) sludge batch qualification program and is documented in the report SS-PIP-2006-00030. That recommendation was supported through a technical data review by the Savannah River National Laboratory (SRNL) and is documented in the memorandums SRNL-PSE-2007-00079 and SRNL-PSE-2007-00080. At the time of writing those memorandums, the DWPF was processing sludge-only waste but, has since transitioned to a coupled operation (sludge and salt). The SRNL was recently tasked to perform a similar data review relevant to coupled operations and re-evaluate the previous recommendations. This report evaluates the validity of eliminating the characterization of pour stream glass samples and the glass fabrication and Product Consistency Test (PCT) of the sludge batch qualification samples based on sludge-only and coupled operations. The pour stream sample has confirmed the DWPF's ability to produce an acceptable waste form from Slurry Mix Evaporator (SME) blending and product composition/durability predictions for the previous sixteen years but, ultimately the pour stream analysis has added minimal value to the DWPF's waste qualification strategy. Similarly, the information gained from the glass fabrication and PCT of the sludge batch qualification sample was determined to add minimal value to the waste qualification strategy since that sample is routinely not representative of the waste composition ultimately processed at the DWPF due to blending and salt processing considerations. Moreover, the qualification process has repeatedly confirmed minimal differences in glass behavior from actual radioactive waste to glasses fabricated from simulants or batch chemicals. In contrast, the

  18. Pilot scale processing of simulated Savannah River Site high level radioactive waste

    International Nuclear Information System (INIS)

    Hutson, N.D.; Zamecnik, J.R.; Ritter, J.A.; Carter, J.T.

    1991-01-01

    The Savannah River Laboratory operates the Integrated DWPF Melter System (IDMS), which is a pilot-scale test facility used in support of the start-up and operation of the US Department of Energy's Defense Waste Processing Facility (DWPF). Specifically, the IDMS is used in the evaluation of the DWPF melter and its associated feed preparation and offgass treatment systems. This article provides a general overview of some of the test work which has been conducted in the IDMS facility. The chemistry associated with the chemical treatment of the sludge (via formic acid adjustment) is discussed. Operating experiences with simulated sludge containing high levels of nitrite, mercury, and noble metals are summarized

  19. Development Of Remote Hanford Connector Gasket Replacement Tooling For DWPF

    International Nuclear Information System (INIS)

    Krementz, D.; Coughlin, Jeffrey

    2009-01-01

    The Defense Waste Processing Facility (DWPF) requested the Savannah River National Laboratory (SRNL) to develop tooling and equipment to remotely replace gaskets in mechanical Hanford connectors to reduce personnel radiation exposure as compared to the current hands-on method. It is also expected that radiation levels will continually increase with future waste streams. The equipment is operated in the Remote Equipment Decontamination Cell (REDC), which is equipped with compressed air, two master-slave manipulators (MSM's) and an electro-mechanical manipulator (EMM) arm for operation of the remote tools. The REDC does not provide access to electrical power, so the equipment must be manually or pneumatically operated. The MSM's have a load limit at full extension of ten pounds, which limited the weight of the installation tool. In order to remotely replace Hanford connector gaskets several operations must be performed remotely, these include: removal of the spent gasket and retaining ring (retaining ring is also called snap ring), loading the new snap ring and gasket into the installation tool and installation of the new gasket into the Hanford connector. SRNL developed and tested tools that successfully perform all of the necessary tasks. Removal of snap rings from horizontal and vertical connectors is performed by separate air actuated retaining ring removal tools and is manipulated in the cell by the MSM. In order install a new gasket, the snap ring loader is used to load a new snap ring into a groove in the gasket installation tool. A new gasket is placed on the installation tool and retained by custom springs. An MSM lifts the installation tool and presses the mounted gasket against the connector block. Once the installation tool is in position, the gasket and snap ring are installed onto the connector by pneumatic actuation. All of the tools are located on a custom work table with a pneumatic valve station that directs compressed air to the desired tool and

  20. Glass melter assembly for the Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Chen, A.E.; Russell, A.; Shah, K.R.; Kalia, J.

    1993-01-01

    The Hanford Waste Vitrification Plant (HWVP) is designed to solidify high level radioactive waste by converting it into stable borosilicate after mixing with glass frit and water. The heart of this conversion process takes place in the glass melter. The life span of the existing melter is limited by the possible premature failure of the heater assembly, which is not remotely replaceable, in the riser and pour spout. A goal of HWVP Project is to design remotely replaceable riser and pour spout heaters so that the useful life of the melter can be prolonged. The riser pour spout area is accessible only by the canyon crane and impact wrench. It is also congested with supporting frame members, service piping, electrode terminals, canister positioning arm and other various melter components. The visibility is low and the accessibility is limited. The problem is further compounded by the extreme high temperature in the riser core and the electrical conductive nature of the molten glass that flows through it

  1. DATA SUMMARY REPORT SMALL SCALE MELTER TESTING OF HLW ALGORITHM GLASSES MATRIX1 TESTS VSL-07S1220-1 REV 0 7/25/07

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; PEGG IL

    2011-12-29

    Eight tests using different HLW feeds were conducted on the DM100-BL to determine the effect of variations in glass properties and feed composition on processing rates and melter conditions (off-gas characteristics, glass processing, foaming, cold cap, etc.) at constant bubbling rate. In over seven hundred hours of testing, the property extremes of glass viscosity, electrical conductivity, and T{sub 1%}, as well as minimum and maximum concentrations of several major and minor glass components were evaluated using glass compositions that have been tested previously at the crucible scale. Other parameters evaluated with respect to glass processing properties were +/-15% batching errors in the addition of glass forming chemicals (GFCs) to the feed, and variation in the sources of boron and sodium used in the GFCs. Tests evaluating batching errors and GFC source employed variations on the HLW98-86 formulation (a glass composition formulated for HLW C-106/AY-102 waste and processed in several previous melter tests) in order to best isolate the effect of each test variable. These tests are outlined in a Test Plan that was prepared in response to the Test Specification for this work. The present report provides summary level data for all of the tests in the first test matrix (Matrix 1) in the Test Plan. Summary results from the remaining tests, investigating minimum and maximum concentrations of major and minor glass components employing variations on the HLW98-86 formulation and glasses generated by the HLW glass formulation algorithm, will be reported separately after those tests are completed. The test data summarized herein include glass production rates, the type and amount of feed used, a variety of measured melter parameters including temperatures and electrode power, feed sample analysis, measured glass properties, and gaseous emissions rates. More detailed information and analysis from the melter tests with complete emission chemistry, glass durability, and

  2. DWPF glass transition temperatures - What they are and why they are important

    International Nuclear Information System (INIS)

    Marra, S.L.; Applewhite-Ramsey, A.L.; Jantzen, C.M.

    1991-01-01

    The Department of Energy has defined a set of requirements for the DWPF canistered waste form which must be met in order to assure compatibility with, and acceptance by, the first geologic repository. These requirements are the Waste Acceptance Preliminary Specifications (WAPS). The WAPS require DWPF to report glass transition temperatures for the projected range of compositions. This information will be used by the repository to establish waste package design limits

  3. DWPF upgrade, immobilization Programmatic Environmental Impact Statement input. Revision 1

    International Nuclear Information System (INIS)

    Sullivan, I.K.; Bignell, D.

    1994-01-01

    This Programmatic Environmental Impact Statement (PEIS) addresses the immobilization of plutonium by vitrification. Existing engineering documents, analyses, EIS, and technical publications were used and incorporated wherever possible to provide a timely response to this support effort. Although the vitrification technology is proven for the immobilization of high-level radioactive waste, more study and technical detail will be necessary to provide a comprehensive EIS that fully addresses all aspects of introduction of plutonium to the vitrification process. This document describes the concept(s) of plutonium processing as it relates to the upgrade of the DWPF and is therefore conceptual in nature. These concepts are based on technical data and experience at the Savannah River Site and will be detailed and finalized to support execution of this immobilization option

  4. Final Report Tests On The Duramelter 1200 HLW Pilot Melter System Using AZ-101 HLW Simulants VSL-02R0100-2, Rev. 1, 2/17/03

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Bardakci, T.; Gong, W.; D'Angelo, N.A.; Schatz, T.R.; Pegg, I.L.

    2011-01-01

    This document provides the final report on data and results obtained from a series of nine tests performed on the one-third scale DuraMelter(trademark) 1200 (DM1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part B1 (1). Both melters have similar melt surface areas (1.2 m 2 ) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plans. The nine tests reported here were preceded by an initial series of short-duration tests conducted to support the start-up and commissioning of this system. This report is a followup to the previously issued Preliminary Data Summary Reports. The DM1200 system was deployed for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. These tests include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The primary objective of the present series of tests was to determine the effects of a variety of parameters on the glass production rate in comparison to the RPP-WTP HL W design basis of 400 kg/m 2 /d. Previous testing on the DMIOOO system (1) concluded that achievement of that rate with simulants of projected WTP melter feeds (AZ-101 and C-106/AY-102) was unlikely without the use of bubblers. As part of those tests, the same feed that was used during the cold-commissioning of the West Valley Demonstration Project (WVDP) HLW vitrification system was run on the DM1000 system. The DM1000 tests reproduced the rates that were obtained at the larger

  5. FINAL REPORT TESTS ON THE DURAMELTER 1200 HLW PILOT MELTER SYSTEM USING AZ-101 HLW SIMULANTS VSL-02R0100-2 REV 1 2/17/03

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; KOT WK; BARDAKCI T; GONG W; D' ANGELO NA; SCHATZ TR; PEGG IL

    2011-12-29

    This document provides the final report on data and results obtained from a series of nine tests performed on the one-third scale DuraMelter{trademark} 1200 (DM1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part B1 [1]. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plans. The nine tests reported here were preceded by an initial series of short-duration tests conducted to support the start-up and commissioning of this system. This report is a followup to the previously issued Preliminary Data Summary Reports. The DM1200 system was deployed for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. These tests include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The primary objective of the present series of tests was to determine the effects of a variety of parameters on the glass production rate in comparison to the RPP-WTP HL W design basis of 400 kg/m{sup 2}/d. Previous testing on the DMIOOO system [1] concluded that achievement of that rate with simulants of projected WTP melter feeds (AZ-101 and C-106/AY-102) was unlikely without the use of bubblers. As part of those tests, the same feed that was used during the cold-commissioning of the West Valley Demonstration Project (WVDP) HLW vitrification system was run on the DM1000 system. The DM1000 tests reproduced the rates that were obtained at the

  6. Crystal accumulation in the Hanford Waste Treatment Plant high level waste melter: Summary of FY2016 experiements

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. [Savannah River Site (SRS), Aiken, SC (United States); Fowley, M. [Savannah River Site (SRS), Aiken, SC (United States); Miller, D. [Savannah River Site (SRS), Aiken, SC (United States)

    2016-12-01

    Five experiments were completed with the full-scale, room temperature Hanford Waste Treatment and Immobilization Plant (WTP) high-level waste (HLW) melter riser test system to observe particle flow and settling in support of a crystal tolerant approach to melter operation. A prototypic pour rate was maintained based on the volumetric flow rate. Accumulation of particles was observed at the bottom of the riser and along the bottom of the throat after each experiment. Measurements of the accumulated layer thicknesses showed that the settled particles at the bottom of the riser did not vary in thickness during pouring cycles or idle periods. Some of the settled particles at the bottom of the throat were re-suspended during subsequent pouring cycles, and settled back to approximately the same thickness after each idle period. The cause of the consistency of the accumulated layer thicknesses is not year clear, but was hypothesized to be related to particle flow back to the feed tank. Additional experiments reinforced the observation of particle flow along a considerable portion of the throat during idle periods. Limitations of the system are noted in this report and may be addressed via future modifications. Follow-on experiments will be designed to evaluate the impact of pouring rate on particle re-suspension, the influence of feed tank agitation on particle accumulation, and the effect of changes in air lance positioning on the accumulation and re-suspension of particles at the bottom of the riser. A method for sampling the accumulated particles will be developed to support particle size distribution analyses. Thicker accumulated layers will be intentionally formed via direct addition of particles to select areas of the system to better understand the ability to continue pouring and re-suspend particles. Results from the room temperature system will be correlated with observations and data from the Research Scale Melter (RSM) at Pacific Northwest National Laboratory

  7. Comparison of the rotary calciner-metallic melter and the slurry-fed ceramic melter technologies for vitrifying West Valley high-level wastes

    International Nuclear Information System (INIS)

    Chapman, C.C.

    1983-01-01

    Two processes which are believed applicable and available for vitrification of West Valley's high-level (HLW) wastes were technically evaluated and compared. The rotary calciner-metallic melter (AVH) and the slurry-fed ceramic melter (SFCM) were evaluated under the following general categories: process flow sheet, remote operability, safety and environmental considerations, and estimated cost and schedules

  8. Statistical process control support during Defense Waste Processing Facility chemical runs

    International Nuclear Information System (INIS)

    Brown, K.G.

    1994-01-01

    The Product Composition Control System (PCCS) has been developed to ensure that the wasteforms produced by the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) will satisfy the regulatory and processing criteria that will be imposed. The PCCS provides rigorous, statistically-defensible management of a noisy, multivariate system subject to multiple constraints. The system has been successfully tested and has been used to control the production of the first two melter feed batches during DWPF Chemical Runs. These operations will demonstrate the viability of the DWPF process. This paper provides a brief discussion of the technical foundation for the statistical process control algorithms incorporated into PCCS, and describes the results obtained and lessons learned from DWPF Cold Chemical Run operations. The DWPF will immobilize approximately 130 million liters of high-level nuclear waste currently stored at the Site in 51 carbon steel tanks. Waste handling operations separate this waste into highly radioactive sludge and precipitate streams and less radioactive water soluble salts. (In a separate facility, soluble salts are disposed of as low-level waste in a mixture of cement slag, and flyash.) In DWPF, the precipitate steam (Precipitate Hydrolysis Aqueous or PHA) is blended with the insoluble sludge and ground glass frit to produce melter feed slurry which is continuously fed to the DWPF melter. The melter produces a molten borosilicate glass which is poured into stainless steel canisters for cooling and, ultimately, shipment to and storage in a geologic repository

  9. High-Temperature Corrosion Study for the RPP Low Activity Waste Melter

    International Nuclear Information System (INIS)

    Marshall, K.M.

    2003-01-01

    The River Protection Program (RPP) low activity waste (LAW) melter design incorporates a series of bubblers used to increase convection in the molten glass. Through runs of a pilot melter at Duratek, Inc. in Columbia, Maryland, the bubblers have been identified as the major component limiting LAW melter availability, requiring frequent replacement due to corrosive degradation, primarily at the melt line. Laboratory experiments were performed to evaluate the performance of several alloys and coatings in simulated RPP low activity waste melter vapor space and molten glass environments. The performance of the alloys and coatings was studied in order to advance our understanding of how these materials react at the melt/air interface inside the melter. The ultimate goal was to identify a material with superior performance compared to that of Inconel 693, and to deliver a bubbler sub-assembly made of that material to the RPP LAW melter pilot facility for further testing

  10. GTS Duratek, phase I Hanford low-level waste melter tests: Final report

    International Nuclear Information System (INIS)

    Eaton, W.C.

    1995-01-01

    A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense waste stored in underground tanks at the Hanford Site in southeastern Washington State. Phase 1 of the melter demonstration tests using simulated LLW was completed during fiscal year 1995. This document is the final report on testing performed by GTS Duratek Inc. in Columbia, Maryland. GTS Duratek (one of the seven vendors selected) was chosen to demonstrate Joule heated melter technology under WHC subcontract number MMI-SVV-384215. The report contains description of the tests, observations, test data and some analysis of the data as it pertains to application of this technology for LLW vitrification. The document also contains summaries of the melter offgas reports issued as separate documents for the 100 kg melter (WHC-SD-WM-VI-028) and for the 1000 kg melter (WHC-SD-WM-VI-029)

  11. Advanced waste form and melter development for treatment of troublesome high-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-02

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these "troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approached to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.

  12. Cold crucible induction melter studies for making glass ceramic waste forms: A feasibility assessment

    International Nuclear Information System (INIS)

    Crum, Jarrod; Maio, Vince; McCloy, John; Scott, Clark; Riley, Brian; Benefiel, Brad; Vienna, John; Archibald, Kip; Rodriguez, Carmen; Rutledge, Veronica; Zhu, Zihua; Ryan, Joe; Olszta, Matthew

    2014-01-01

    Glass ceramics are being developed to immobilize fission products, separated from used nuclear fuel by aqueous reprocessing, into a stable waste form suitable for disposal in a geological repository. This work documents the glass ceramic formulation at bench scale and for a scaled melter test performed in a pilot-scale (∼1/4 scale) cold crucible induction melter (CCIM). Melt viscosity, electrical conductivity, and crystallization behavior upon cooling were measured on a small set of compositions to select a formulation for melter testing. Property measurements also identified a temperature range for melter operation and cooling profiles necessary to crystallize the targeted phases in the waste form. Bench scale and melter run results successfully demonstrate the processability of the glass ceramic using the CCIM melter technology

  13. Joule-Heated Ceramic-Lined Melter to Vitrify Liquid Radioactive Wastes Containing Am241 Generated From MOX Fuel Fabrication in Russia

    International Nuclear Information System (INIS)

    Smith, E C; Bowan II, B W; Pegg, I; Jardine, L J

    2004-01-01

    The governments of the United Stated of America and the Russian Federation (RF) signed an Agreement September 1, 2000 to dispose of weapons plutonium that has been designated as no longer required for defense purposes. The Agreement declares that each country will disposition 34MT of excess weapons grade plutonium from their stockpiles. The preferred disposition technology is the fabrication of mixed oxide (MOx) fuel for use or burning in pressurized water reactors to destroy the plutonium. Implementation of this Agreement will require the conversion of plutonium metal to oxide and the fabrication of MOx fuel within the Russian Federation. The MOx fuel fabrication and metal to oxide conversion processes will generate solid and liquid radioactive wastes containing trace amounts of plutonium, neptunium, americium, and uranium requiring treatment, storage, and disposal. Unique to the Russian MOx fuel fabrication facility's flow-sheet is a liquid waste stream with high concentrations (∼1 g/l) of 241 Am and non radioactive silver. The silver is used to dissolve PuO 2 feed materials to the MOx fabrication facility. Technical solutions are needed to treat and solidify this liquid waste stream. Alternative treatment technologies for this liquid waste stream are being evaluated by a Russian engineering team. The technologies being evaluated include borosilicate and phosphate vitrification alternatives. The evaluations are being performed at a conceptual design level of detail under a Lawrence Livermore National Laboratory (LLNL) contract with the Russian organization TVEL using DOE NA-26 funding. As part of this contract, the RF team is evaluating the technical and economic feasibility of the US borosilicate glass vitrification technology based on a Duratek melter to solidify this waste stream into a form acceptable for storage and geologic disposal. The composition of the glass formed from treating the waste is dictated by the concentration of silver and americium it

  14. Materials and design experience in a slurry-fed electric glass melter

    International Nuclear Information System (INIS)

    Barnes, S.M.; Larson, D.E.

    1981-08-01

    The design of a slurry-fed electric gas melter and an examination of the performance and condition of the construction materials were completed. The joule-heated, ceramic-lined melter was constructed to test the applicability of materials and processes for high-level waste vitrification. The developmental Liquid-Fed Ceramic Melter (LFCM) was operated for three years with simulated high-level waste and was subjected to conditions more severe than those expected for a nuclear waste vitrification plant

  15. Electrical service and controls for Joule heating of a defense waste experimental glass melter

    International Nuclear Information System (INIS)

    Erickson, C.J.; Haideri, A.Q.

    1983-01-01

    Vitrification of radioactive liquid waste in a glass matrix is a leading candidate for long-term storage of high-level waste. This paper describes the electrical service and control system for an experimental electrically heated, nonradioactive glass melter installed at Savannah River Laboratory. Data accumulated, and design/operating experience acquired in operating this melter, are being used to design a modified melter to be installed in a processing area for use with radioactive materials

  16. Computer modeling of ceramic melters to assess impacts of process and design variables on performance

    International Nuclear Information System (INIS)

    Eyler, L.L.; Elliott, M.L.; Lowery, P.S.; Lessor, D.L.

    1991-01-01

    Numerical and physical simulation of existing and advanced melter designs conducted to assess impacts of process and design variables on performance of ceramic melters are presented. Coupled equations of flow, thermal, and electric fields were numerically solved in time-dependent three dimensional finite volume form. Recent simulation results of a three electrode melter design with sloped walls indicate the presence of bi-modal stable flow patterns dominated by boundary conditions

  17. Melter system technology testing for Hanford Site low-level tank waste vitrification

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1996-01-01

    Following revisions to the Tri-Party Agreement for Hanford Site cleanup, which specified vitrification for Complete melter feasibility and system operability immobilization of the low-level waste (LLW) tests, select reference melter(s), and establish reference derived from retrieval and pretreatment of the radioactive LLW glass formulation that meets complete systems defense wastes stored in 177 underground tanks, commercial requirements (June 1996). Available melter technologies were tested during 1994 to 1995 as part of a multiphase program to select reference Submit conceptual design and initiate definitive design technologies for the new LLW vitrification mission

  18. Physical modeling of joule heated ceramic glass melters for high level waste immobilization

    International Nuclear Information System (INIS)

    Quigley, M.S.; Kreid, D.K.

    1979-03-01

    This study developed physical modeling techniques and apparatus suitable for experimental analysis of joule heated ceramic glass melters designed for immobilizing high level waste. The physical modeling experiments can give qualitative insight into the design and operation of prototype furnaces and, if properly verified with prototype data, the physical models could be used for quantitative analysis of specific furnaces. Based on evaluation of the results of this study, it is recommended that the following actions and investigations be undertaken: It was not shown that the isothermal boundary conditions imposed by this study established prototypic heat losses through the boundaries of the model. Prototype wall temperatures and heat fluxes should be measured to provide better verification of the accuracy of the physical model. The VECTRA computer code is a two-dimensional analytical model. Physical model runs which are isothermal in the Y direction should be made to provide two-dimensional data for more direct comparison to the VECTRA predictions. The ability of the physical model to accurately predict prototype operating conditions should be proven before the model can become a reliable design tool. This will require significantly more prototype operating and glass property data than were available at the time of this study. A complete set of measurements covering power input, heat balances, wall temperatures, glass temperatures, and glass properties should be attempted for at least one prototype run. The information could be used to verify both physical and analytical models. Particle settling and/or sludge buildup should be studied directly by observing the accumulation of the appropriate size and density particles during feeding in the physical model. New designs should be formulated and modeled to minimize the potential problems with melter operation identifed by this study

  19. High-level waste melter alternatives assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Calmus, R.B.

    1995-02-01

    This document describes the Tank Waste Remediation System (TWRS) High-Level Waste (HLW) Program`s (hereafter referred to as HLW Program) Melter Candidate Assessment Activity performed in fiscal year (FY) 1994. The mission of the TWRS Program is to store, treat, and immobilize highly radioactive Hanford Site waste (current and future tank waste and encapsulated strontium and cesium isotopic sources) in an environmentally sound, safe, and cost-effective manner. The goal of the HLW Program is to immobilize the HLW fraction of pretreated tank waste into a vitrified product suitable for interim onsite storage and eventual offsite disposal at a geologic repository. Preparation of the encapsulated strontium and cesium isotopic sources for final disposal is also included in the HLW Program. As a result of trade studies performed in 1992 and 1993, processes planned for pretreatment of tank wastes were modified substantially because of increasing estimates of the quantity of high-level and transuranic tank waste remaining after pretreatment. This resulted in substantial increases in needed vitrification plant capacity compared to the capacity of original Hanford Waste Vitrification Plant (HWVP). The required capacity has not been finalized, but is expected to be four to eight times that of the HWVP design. The increased capacity requirements for the HLW vitrification plant`s melter prompted the assessment of candidate high-capacity HLW melter technologies to determine the most viable candidates and the required development and testing (D and T) focus required to select the Hanford Site HLW vitrification plant melter system. An assessment process was developed in early 1994. This document describes the assessment team, roles of team members, the phased assessment process and results, resulting recommendations, and the implementation strategy.

  20. High-level waste melter alternatives assessment report

    International Nuclear Information System (INIS)

    Calmus, R.B.

    1995-02-01

    This document describes the Tank Waste Remediation System (TWRS) High-Level Waste (HLW) Program's (hereafter referred to as HLW Program) Melter Candidate Assessment Activity performed in fiscal year (FY) 1994. The mission of the TWRS Program is to store, treat, and immobilize highly radioactive Hanford Site waste (current and future tank waste and encapsulated strontium and cesium isotopic sources) in an environmentally sound, safe, and cost-effective manner. The goal of the HLW Program is to immobilize the HLW fraction of pretreated tank waste into a vitrified product suitable for interim onsite storage and eventual offsite disposal at a geologic repository. Preparation of the encapsulated strontium and cesium isotopic sources for final disposal is also included in the HLW Program. As a result of trade studies performed in 1992 and 1993, processes planned for pretreatment of tank wastes were modified substantially because of increasing estimates of the quantity of high-level and transuranic tank waste remaining after pretreatment. This resulted in substantial increases in needed vitrification plant capacity compared to the capacity of original Hanford Waste Vitrification Plant (HWVP). The required capacity has not been finalized, but is expected to be four to eight times that of the HWVP design. The increased capacity requirements for the HLW vitrification plant's melter prompted the assessment of candidate high-capacity HLW melter technologies to determine the most viable candidates and the required development and testing (D and T) focus required to select the Hanford Site HLW vitrification plant melter system. An assessment process was developed in early 1994. This document describes the assessment team, roles of team members, the phased assessment process and results, resulting recommendations, and the implementation strategy

  1. Melter development needs assessment for RWMC buried wastes

    International Nuclear Information System (INIS)

    Donaldson, A.D.; Carpenedo, R.J.; Anderson, G.L.

    1992-02-01

    This report presents a survey and initial assessment of the existing state-of-the-art melter technology necessary to thermally treat (stabilize) buried TRU waste, by producing a highly leach resistant glass/ceramic waste form suitable for final disposal. Buried mixed transuranic (TRU) waste at the Idaho National Engineering Laboratory (INEL) represents an environmental hazard requiring remediation. The Environmental Protection Agency (EPA) placed the INEL on the National Priorities List in 1989. Remediation of the buried TRU-contaminated waste via the CERCLA decision process is required to remove INEL from the National Priorities List. A Waste Technology Development (WTD) Preliminary Systems Design and Thermal Technologies Screening Study identified joule-heated and plasma-heated melters as the most probable thermal systems technologies capable of melting the INEL soil and waste to produce the desired final waste form [Iron-Enriched Basalt (IEB) glass/ceramic]. The work reported herein then surveys the state of existing melter technology and assesses it within the context of processing INEL buried TRU wastes and contaminated soils. Necessary technology development work is recommended

  2. Remote Fiber Laser Cutting System for Dismantling Glass Melter - 13071

    Energy Technology Data Exchange (ETDEWEB)

    Mitsui, Takashi; Miura, Noriaki [IHI Corporation, 1 Shin-Nakahara-cho, Isogo-ku, Yokohama, Kanagawa (Japan); Oowaki, Katsura; Kawaguchi, Isao [IHI Inspection and Instrumentation Co., Ltd, 1 Shin-Nakahara-cho, Isogo-ku, Yokohama, Kanagawa (Japan); Miura, Yasuhiko; Ino, Tooru [Japan Nuclear Fuel Limited, 4-108, Aza Okitsuke, Oaza Obuchi, Rokkasho-Mura, Kamikita-gun, Aomori (Japan)

    2013-07-01

    Since 2008, the equipment for dismantling the used glass melter has been developed in High-level Liquid Waste (HLW) Vitrification Facility in the Japanese Rokkasho Reprocessing Plant (RRP). Due to the high radioactivity of the glass melter, the equipment requires a fully-remote operation in the vitrification cell. The remote fiber laser cutting system was adopted as one of the major pieces of equipment. An output power of fiber laser is typically higher than other types of laser and so can provide high-cutting performance. The fiber laser can cut thick stainless steel and Inconel, which are parts of the glass melter such as casings, electrodes and nozzles. As a result, it can make the whole of the dismantling work efficiently done for a shorter period. Various conditions of the cutting test have been evaluated in the process of developing the remote fiber cutting system. In addition, the expected remote operations of the power manipulator with the laser torch have been fully verified and optimized using 3D simulations. (authors)

  3. Characterization of projected DWPF glasses heat treated to simulate canister centerline cooling

    International Nuclear Information System (INIS)

    Marra, S.L.; Jantzen, C.M.

    1992-05-01

    Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). Eventually these canistered waste forms will be sent to a geologic repository for final disposal. In order to assure acceptability by the repository, the Department of Energy has defined requirements which DWPF canistered waste forms must meet. These requirements are the Waste Acceptance Preliminary Specifications (WAPS). The WAPS require DWPF to identify the crystalline phases expected to be present in the final glass product. Knowledge of the thermal history of the borosilicate glass during filling and cooldown of the canister is necessary to determine the amount and type of crystalline phases present in the final glass product. Glass samples of seven projected DWPF compositions were cooled following the same temperature profile as that of glass at the centerline of the full-scale DWPF canister. The glasses were characterized by x-ray diffraction and scanning electron microscopy to identify the crystalline phases present The volume percents of each crystalline phase present were determined by quantitative x-ray diffraction. The Product Consistency Test (PCI) was used to determine the durability of the heat-treated glasses

  4. Startup and operation of a plant-scale continuous glass melter for vitrification of Savannah River Plant simulated waste

    International Nuclear Information System (INIS)

    Willis, T.A.

    1980-01-01

    The reference process for disposal of radioactive waste from the Savannah River Plant is vitrification of the waste in borosilicate glass in a continuous glass melter. Design, startup, and operation of a plant-scale developmental melter system are discussed

  5. Dissolution rates of DWPF glasses from long-term PCT

    International Nuclear Information System (INIS)

    Ebert, W.L.; Tam, S.W.

    1996-01-01

    We have characterized the corrosion behavior of several Defense Waste Processing Facility (DWPF) reference waste glasses by conducting static dissolution tests with crushed glasses. Glass dissolution rates were calculated from measured B concentrations in tests conducted for up to five years. The dissolution rates of all glasses increased significantly after certain alteration phases precipitated. Calculation of the dissolution rates was complicated by the decrease in the available surface area as the glass dissolves. We took the loss of surface area into account by modeling the particles to be spheres, then extracting from the short-term test results the dissolution rate corresponding to a linear decrease in the radius of spherical particles. The measured extent of dissolution in tests conducted for longer times was less than predicted with this linear dissolution model. This indicates that advanced stages of corrosion are affected by another process besides dissolution, which we believe to be associated with a decrease in the precipitation rate of the alteration phases. These results show that the dissolution rate measured soon after the formation of certain alteration phases provides an upper limit for the long-term dissolution rate, and can be used to determine a bounding value for the source term for radionuclide release from waste glasses. The long-term dissolution rates measured in tests at 20,000 per m at 90 degrees C in tuff groundwater at pH values near 12 for the Environmental Assessment glass and glasses made with SRL 131 and SRL 202 frits, respectively

  6. Erosion Modeling Analysis For Modified DWPF SME Tank

    International Nuclear Information System (INIS)

    LEE, SI

    2004-01-01

    In support of an erosion evaluation for the modified cooling coil guide and its supporting structure in the DWPF SME vessel, a computational model was developed to identify potential sites of high erosion using the same methodology established by previous work. The erosion mechanism identified in the previous work was applied to the evaluation of high erosion locations representative of the actual flow process in the modified coil guide of the SME vessel, abrasive erosion which occurs by high wall shear of viscous liquid. The results show that primary locations of the highest erosion due to the abrasive wall erosion are at the leading edge of the guide, external surface of the insert plate, the tank floor next to the insert plate of the coil guide support, and the upstream lead-in plate. The present modeling results show a good comparison between the original and the modified cases in terms of high erosion sites, as well as the degree of erosion and the calculated shear stress. Wall she ar of the tank floor is reduced by about 30 per cent because of the new coil support plate. Calculations for the impeller speed lower than 103 rpm in the SME showed similar erosion patterns but significantly reduced wall shear stresses and reduced overall erosion. Comparisons of the 103 rpm results with SME measurements indicated that no significant erosion of the tank floor in the SME is to be expected. Thus, it is recommended that the agitator speed of SME does not exceed 103 rpm

  7. Cold-Crucible Design Parameters for Next Generation HLW Melters

    International Nuclear Information System (INIS)

    Gombert, D.; Richardson, J.; Aloy, A.; Day, D.

    2002-01-01

    The cold-crucible induction melter (CCIM) design eliminates many materials and operating constraints inherent in joule-heated melter (JHM) technology, which is the standard for vitrification of high-activity wastes worldwide. The cold-crucible design is smaller, less expensive, and generates much less waste for ultimate disposal. It should also allow a much more flexible operating envelope, which will be crucial if the heterogeneous wastes at the DOE reprocessing sites are to be vitrified. A joule-heated melter operates by passing current between water-cooled electrodes through a molten pool in a refractory-lined chamber. This design is inherently limited by susceptibility of materials to corrosion and melting. In addition, redox conditions and free metal content have exacerbated materials problems or lead to electrical short-circuiting causing failures in DOE melters. In contrast, the CCIM design is based on inductive coupling of a water-cooled high-frequency electrical coil with the glass, causing eddycurrents that produce heat and mixing. A critical difference is that inductance coupling transfers energy through a nonconductive solid layer of slag coating the metal container inside the coil, whereas the jouleheated design relies on passing current through conductive molten glass in direct contact with the metal electrodes and ceramic refractories. The frozen slag in the CCIM design protects the containment and eliminates the need for refractory, while the corrosive molten glass can be the limiting factor in the JH melter design. The CCIM design also eliminates the need for electrodes that typically limit operating temperature to below 1200 degrees C. While significant marketing claims have been made by French and Russian technology suppliers and developers, little data is available for engineering and economic evaluation of the technology, and no facilities are available in the US to support testing. A currently funded project at the Idaho National Engineering

  8. FINAL REPORT INTEGRATED DM1200 MELTER TESTING OF REDOX EFFECTS USING HLW AZ-101 AND C-106/AY-102 SIMULANTS VSL-04R4800-1 REV 0 5/6/

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; LUTZE W; BIZOT PM; CALLOW RA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 and C-106/AY-102 HLW simulants. The tests reported herein are a subset of three tests from a larger series of tests described in the Test Plan for the work; results from the remaining tests will be reported separately. Three nine day tests, one with AZ-101 and two with C-106/AY-102 feeds were conducted with variable amounts of added sugar to address the effects of redox. The test with AZ-101 included ruthenium spikes to also address the effects of redox on ruthenium volatility. One of tests addressed the effects of increased flow-sheet nitrate levels using C-106/AY-102 feeds. With high nitrate/nitrite feeds (such as WTP LAW feeds), reductants are required to prevent melt foaming and deleterious effects on glass production rates. Sugar is the baseline WTP reductant for this purpose. WTP HLW feeds typically have relatively low nitrate/nitrite content in comparison to the organic carbon content and, therefore, have typically not required sugar additions. However, HLW feed variability, particularly with respect to nitrate levels, may necessitate the use of sugar in some instances. The tests reported here investigate the effects of variable sugar additions to the melter feed as well as elevated nitrate levels in the waste. Variables held constant to the extent possible included melt temperature, bubbling rate, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW feeds with variable amounts of added sugar and increased nitrate levels; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and

  9. Final Report Integrated DM1200 Melter Testing Of Redox Effects Using HLW AZ-101 And C-106/AY-102 Simulants VSL-04R4800-1, Rev. 0, 5/6/04

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Gong, W.; Bardakci, T.; D'Angelo, N.A.; Lutze, W.; Bizot, P.M.; Callow, R.A.; Brandys, M.; Kot, W.K.; Pegg, I.L.

    2011-01-01

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 and C-106/AY-102 HLW simulants. The tests reported herein are a subset of three tests from a larger series of tests described in the Test Plan for the work; results from the remaining tests will be reported separately. Three nine day tests, one with AZ-101 and two with C-106/AY-102 feeds were conducted with variable amounts of added sugar to address the effects of redox. The test with AZ-101 included ruthenium spikes to also address the effects of redox on ruthenium volatility. One of tests addressed the effects of increased flow-sheet nitrate levels using C-106/AY-102 feeds. With high nitrate/nitrite feeds (such as WTP LAW feeds), reductants are required to prevent melt foaming and deleterious effects on glass production rates. Sugar is the baseline WTP reductant for this purpose. WTP HLW feeds typically have relatively low nitrate/nitrite content in comparison to the organic carbon content and, therefore, have typically not required sugar additions. However, HLW feed variability, particularly with respect to nitrate levels, may necessitate the use of sugar in some instances. The tests reported here investigate the effects of variable sugar additions to the melter feed as well as elevated nitrate levels in the waste. Variables held constant to the extent possible included melt temperature, bubbling rate, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW feeds with variable amounts of added sugar and increased nitrate levels; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and

  10. Preliminary low-level waste feed definition guidance - LLW pretreatment interface

    International Nuclear Information System (INIS)

    Shade, J.W.; Connor, J.M.; Hendrickson, D.W.; Powell, W.J.; Watrous, R.A.

    1995-02-01

    The document describes limits for key constituents in the LLW feed, and the bases for these limits. The potential variability in the stream is then estimated and compared to the limits. Approaches for accomodating uncertainty in feed inventory, processing strategies, and process design (melter and disposal system) are discussed. Finally, regulatory constraints are briefly addressed

  11. Compatibility tests of materials for a prototype ceramic melter for defense glass-waste products

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1979-01-01

    Objective is to evaluate the corrosion/erosion resistance of melter materials. Materials tested were Monofrox K3 and E, Serv, Inconel 690, Pt, and SnO. Results show that Inconel 690 is the leading electrode material and Monofrox K3 the leading refractory candidate. Melter lifetime is estimated to be 2 to 5 years for defense waste

  12. Startup of a Joule-heated glass melter with a graphite slurry

    International Nuclear Information System (INIS)

    Allen, T.L.; Porter, M.A.; Routt, K.R.

    1984-01-01

    Startup of a Joule-heated glass melter using a graphite slurry as a conducting medium was demonstrated. This technique can be used for the initial startup and for the restart of a melter used for vitrifying high-level radioactive waste. Theory, physical property data, and a demonstration test are reported

  13. Assessment of combustion and related issues in the DWPF and ITP waste tanks

    International Nuclear Information System (INIS)

    Ginsberg, T.

    1994-04-01

    This report presents a review of the safety analyses described in the DWPF Safety Analysis Report, the combustion analysis of the ITP Tanks 48 and 49, and presents conclusions drawn from interviews staff on issues related to accident analysis, in particular on issues related to combustion phenomena. The major objectives of this report are to clarify the issues related to the modes of combustion and expected loads on process vessels and structures and, in addition, to offer recommendations which would improve the defense-in-depth posture of the DWPF

  14. Material compatibility evaluation for DWPF nitric-glycolic acid-literature review

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Skidmore, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2013-06-01

    Glycolic acid is being evaluated as an alternative for formic and nitric acid in the DWPF flowsheet. Demonstration testing and modeling for this new flowsheet has shown that glycolic acid and glycolate has a potential to remain in certain streams generated during the production of the nuclear waste glass. A literature review was conducted to assess the impact of glycolic acid on the corrosion of the materials of construction for the DWPF facility as well as facilities downstream which may have residual glycolic acid and glycolates present. The literature data was limited to solutions containing principally glycolic acid.

  15. Discrete event simulation of the Defense Waste Processing Facility (DWPF) analytical laboratory

    International Nuclear Information System (INIS)

    Shanahan, K.L.

    1992-02-01

    A discrete event simulation of the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) analytical laboratory has been constructed in the GPSS language. It was used to estimate laboratory analysis times at process analytical hold points and to study the effect of sample number on those times. Typical results are presented for three different simultaneous representing increasing levels of complexity, and for different sampling schemes. Example equipment utilization time plots are also included. SRS DWPF laboratory management and chemists found the simulations very useful for resource and schedule planning

  16. Final Report Melter Tests With AZ-101 HLW Simulant Using A Duramelter 100 Vitrification System VSL-01R10N0-1, Rev. 1, 2/25/02

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Pegg, I.L.

    2011-01-01

    This report provides data, analyses, and conclusions from a series of tests that were conducted at the Vitreous State Laboratory of The Catholic of America (VSL) to determine the processing rates that are achievable with AZ-101 HLW simulants and corresponding melter feeds on a DuraMelter 100 (DM100) vitrification system. One of the most critical pieces of information in determining the required size of the RPP-WTP HLW melter is the specific glass production rate in terms of the mass of glass that can be produced per unit area of melt surface per unit time. The specific glass production rate together with the waste loading (essentially, the ratio of waste-in to glass-out, which is determined from glass formulation activities) determines the melt area that is needed to achieve a given waste processing rate with due allowance for system availability. Tests conducted during Part B1 (VSL-00R2590-2) on the DM1000 vitrification system installed at the Vitreous State Laboratory of The Catholic University of America showed that, without the use of bubblers, glass production rates with AZ-101 and C-106/AY-102 simulants were significantly lower than the Project design basis rate of 0.4 MT/m 2 /d. Conversely, three-fold increases over the design basis rate were demonstrated with the use of bubblers. Furthermore, an un-bubbled control test using a replica of the melter feed used in cold commissioning tests at West Valley reproduced the rates that were observed with that feed on the WVDP production melter. More recent tests conducted on the DM1200 system, which more closely represents the present RPP-WTP design, are in general agreement with these earlier results. Screening tests conducted on the DM10 system have provided good indications of the larger-scale processing rates with bubblers (for both HL W and LAW feeds) but significantly overestimated the DM1000 un-bubbled rate observed for C-106/AY-102 melter feeds. This behavior is believed to be a consequence of the role of

  17. FINAL REPORT MELTER TESTS WITH AZ-101 HLW SIMULANT USING A DURAMELTER 100 VITRIFICATION SYSTEM VSL-01R10N0-1 REV 1 2/25/02

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; KOT WK; PEGG IL

    2011-12-29

    This report provides data, analyses, and conclusions from a series of tests that were conducted at the Vitreous State Laboratory of The Catholic of America (VSL) to determine the processing rates that are achievable with AZ-101 HLW simulants and corresponding melter feeds on a DuraMelter 100 (DM100) vitrification system. One of the most critical pieces of information in determining the required size of the RPP-WTP HLW melter is the specific glass production rate in terms of the mass of glass that can be produced per unit area of melt surface per unit time. The specific glass production rate together with the waste loading (essentially, the ratio of waste-in to glass-out, which is determined from glass formulation activities) determines the melt area that is needed to achieve a given waste processing rate with due allowance for system availability. Tests conducted during Part B1 (VSL-00R2590-2) on the DM1000 vitrification system installed at the Vitreous State Laboratory of The Catholic University of America showed that, without the use of bubblers, glass production rates with AZ-101 and C-106/AY-102 simulants were significantly lower than the Project design basis rate of 0.4 MT/m{sup 2}/d. Conversely, three-fold increases over the design basis rate were demonstrated with the use of bubblers. Furthermore, an un-bubbled control test using a replica of the melter feed used in cold commissioning tests at West Valley reproduced the rates that were observed with that feed on the WVDP production melter. More recent tests conducted on the DM1200 system, which more closely represents the present RPP-WTP design, are in general agreement with these earlier results. Screening tests conducted on the DM10 system have provided good indications of the larger-scale processing rates with bubblers (for both HL W and LAW feeds) but significantly overestimated the DM1000 un-bubbled rate observed for C-106/AY-102 melter feeds. This behavior is believed to be a consequence of the role of

  18. Volatility and entrainment of feed components and product glass characteristics during pilot-scale vitrification of simulated Hanford site low-level waste

    International Nuclear Information System (INIS)

    Shade, J.W.

    1996-01-01

    Commercially available melter technologies were tested for application to vitrification of Hanford site low-level waste (LLW). Testing was conducted at vendor facilities using a non-radioactive LLW simulant. Technologies tested included four Joule-heated melter types, a carbon electrode melter, a cyclone combustion melter, and a plasma torch-fired melter. A variety of samples were collected during the vendor tests and analyzed to provide data to support evaluation of the technologies. This paper describes the evaluation of melter feed component volatility and entrainment losses and product glass samples produced during the vendor tests. All vendors produced glasses that met minimum leach criteria established for the test glass formulations, although in many cases the waste oxide loading was less than intended. Entrainment was much lower in Joule-heated systems than in the combustion or plasma torch-fired systems. Volatility of alkali metals, halogens, B, Mo, and P were severe for non-Joule-heated systems. While losses of sulfur were significant for all systems, the volatility of other components was greatly reduced for some configurations of Joule-heated melters. Data on approaches to reduce NO x generation, resulting from high nitrate and nitrite content in the double-shell slurry feed, are also presented

  19. Hanford low-level vitrification melter testing -- Master list of data submittals

    International Nuclear Information System (INIS)

    Hendrickson, D.W.

    1995-01-01

    The Westinghouse Hanford Company (WHC) is conducting a two-phased effort to evaluate melter system technologies for vitrification of liquid low-level radioactive waste (LLW) streams. The evaluation effort includes demonstration testing of selected glass melter technologies and technical reports regarding the applicability of the glass melter technologies to the vitrification of Hanford LLW tank waste. The scope of this document is to identify and list vendor document submittals in technology demonstration support of the Hanford Low-Level Waste Vitrification melter testing program. The scope of this document is limited to those documents responsive to the Statement of Work, accepted and issued by the LLW Vitrification Program. The purpose of such a list is to maintain configuration control of vendor supplied data and to enable ready access to, and application of, vendor supplied data in the evaluation of melter technologies for the vitrification of Hanford low-level tank wastes

  20. Development of HWVP melter/turntable components for canyon-remote maintenance and replacement

    International Nuclear Information System (INIS)

    Siemens, D.H.; Beary, M.M.; Berger, D.N.; Heath, W.O.; Larson, D.E.

    1985-03-01

    Remote operability and maintainability of vitrification equipment were assessed under shielded-cell conditions. The equipment tested will be applied to immobilize high-level and transuranic liquid waste slurries that resulted from plutonium production for defense weapons. Equipment tested included: (1) a turntable for handling waste canisters under the melter; (2) a removable discharge cone in the melter overflow section; (3) a thermocouple jumper that extends into a shielded cell; (4) remote instrument and electrical connectors; (5) remote, mechanical, and heat transfer aspects of the melter glass overflow section; (6) a reamer to clean out plugged nozzles in the melter top; (7) a closed circuit camera to view the melter interior; and (8) a device to retrieve samples of the glass product. 14 figs

  1. Preparation and Heat-Treatment of DWPF Simulants With and Without Co-Precipitated Noble Metals

    International Nuclear Information System (INIS)

    Koopman, David C.:Eibling, Russel E

    2005-01-01

    The Savannah River National Laboratory is in the process of investigating factors suspected of impacting catalytic hydrogen generation in the Chemical Process Cell of the Defense Waste Processing Facility, DWPF. Noble metal catalyzed hydrogen generation in simulation work constrains the allowable acid addition operating window in DWPF. This constraint potentially impacts washing strategies during sludge batch preparation. It can also influence decisions related to the addition of secondary waste streams to a sludge batch. Noble metals have historically been added as trim chemicals to process simulations. The present study investigated the potential conservatism that might be present from adding the catalytic species as trim chemicals to the final sludge simulant versus co-precipitating the noble metals into the insoluble sludge solids matrix. Parallel preparations of two sludge simulants targeting the composition of Sludge Batch 3 were performed in order to evaluate the impact of the form of noble metals. Identical steps were used except that one simulant had dissolved palladium, rhodium, and ruthenium present during the precipitation of the insoluble solids. Noble metals were trimmed into the other stimulant prior to process tests. Portions of both sludge simulants were held at 97 C for about eight hours to qualitatively simulate the effects of long term storage on particle morphology and speciation. The simulants were used as feeds for Sludge Receipt and Adjustment Tank, SRAT, process simulations. The following conclusions were drawn from the simulant preparation work: (1) The first preparation of a waste slurry simulant with co-precipitated noble metals was successful, based on the data obtained. It appears that 99+% of the noble metals were retained in the simulant. (2) Better control of carbonate, hydroxide, and post-wash trim chemical additions is needed before the new method of simulant preparation will be as reproducible as the old method. (3) The two new

  2. Nuclear waste glass melter: an update of technical progress

    International Nuclear Information System (INIS)

    Brouns, R.A.; Hanson, M.S.

    1984-08-01

    The direct slurry-fed ceramic-lined melter is currently the reference US process for treating defense and civilian high-level liquid waste. Extensive nonradioactive pilot-scale testing at Pacific Northwest Laboratory (PNL) and Savannah River Laboratory has proven the process, defined operating parameters, and identified successful equipment design concepts. Programs at PNL continue to support several of the planned US vitrification plants through preparation of equipment designs and flowsheet testing. Current emphasis is on remotization of equipment, radioactive verification testing, and resolution of remaining technical issues. Development of this technology, technical status, and planned development activities are discussed. 9 references, 4 figures

  3. Metallurgical Evaluation of the Five-Inch Cylindrical Induction Melter

    International Nuclear Information System (INIS)

    Imrich, K.J.

    2000-01-01

    A metallurgical evaluation of the 5-inch cylindrical induction melter (CIM) vessel was performed by the Materials Technology Section to evaluate the metallurgical condition after operating for approximately 375 hours at 1400 to 1500 Degrees Celsius during a 2 year period. Results indicate that wall thinning and significant grain growth occurred in the lower portion of the conical section and the drain tube. No through-wall penetrations were found in the cylindrical and conical sections of the CIM vessel and only one leak site was identified in the drain tube. Failure of the drain tube was associated with a localized over heating and intercrystalline fracture

  4. DM100 AND DM1200 MELTER TESTING WITH HIGH WASTE LOADING FORMULATIONS FOR HANFORD HIGH-ALUMINUM HLW STREAMS, TEST PLAN 09T1690-1

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Pegg, I.L.; Joseph, I.

    2009-01-01

    This Test Plan describes work to support the development and testing of high waste loading glass formulations that achieve high glass melting rates for Hanford high aluminum high level waste (HLW). In particular, the present testing is designed to evaluate the effect of using low activity waste (LAW) waste streams as a source of sodium in place ofchemical additives, sugar or cellulose as a reductant, boehmite as an aluminum source, and further enhancements to waste processing rate while meeting all processing and product quality requirements. The work will include preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM 100) tests designed to examine the effects of enhanced glass formulations, glass processing temperature, incorporation of the LAW waste stream as a sodium source, type of organic reductant, and feed solids content on waste processing rate and product quality. Also included is a confirmatory test on the HLW Pilot Melter (DM1200) with a composition selected from those tested on the DM100. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy's (DOE's) Office of River Protection (ORP) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same waste composition. This Test Plan is prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is about 12,500. This estimate is based upon the inventory ofthe tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat transfer and

  5. Neptunium sorption and co-precipitation of strontium in simulated DWPF salt solution

    International Nuclear Information System (INIS)

    McIntyre, P.F.; Orebaugh, E.G.; King, C.M.

    1988-01-01

    Batch experiments performed using crushed slag saltstone (∼40 mesh) removed >80% of 237 Np from simulated Defense Waste Processing Facility (DWPF) salt solution. The concentration of 237 Np (110 pCi/ml) used was 1000x greater than levels in actual DWPF solutions. Neptunium-239 was used as a tracer and was formed by neutron activation of uranyl nitrate. Results showed that small amounts of crushed saltstone (as little as 0.05 grams), removed >80% of neptunium from 15 ml of simulated DWPF solution after several hours equilibration. The neptunium is sorbed on insoluble carbonates formed in and on the saltstone matrix. Further testing showed that addition of 0.01 and 0.10 ml of 1 molar Ca +2 (ie. Ca (NO 3 ) 2 , CaCl 2 ) into 15 ml of simulated DWPF solution yielded a white carbonate precipitate which also removed >80% of the neptunium after 1 hour equilibration. Further experiments were performed to determine the effectiveness of this procedure to co-precipitate strontium

  6. Noble metal (NM) behavior during simulated HLLW vitrification in induction melter with cold crucible

    International Nuclear Information System (INIS)

    Demin, A.V.; Matyunin, Y.I.; Fedorova, M.I.

    1995-01-01

    The investigation of noble metal (Ru, Rh, Pd) properties in, glass melts are connected with their specific behaviors during HLLW vitrification. Ruthenium, rhodium and palladium volatilities and heterogeneous platinoid phases forming on melts are investigated in reasonable details conformably to Joule's heating ceramic melters. The vitrification conditions in melters with induction heating of melts are differ from the vitrification ones in ceramic melters on some numbers of parameters (the availability of significant temperature gradients and convection flows in melts, short time of molten mass updating in melter and probability of definite interaction between high-frequency field and melt inhomogeneities). The results of simulated HLLW solidification modelling of the vitrification process in induction melter with cold crucible to produce phosphate and boron-silicate materials are presented. The properties of received glasses and behavior of platinoids are shown to have analogies and distinctions in comparison with compounds, synthesized in ceramic melter. The structures of dispersed particles of NM heterogeneous phases forming in glass melts prepared in induction melter with cold crucible are identified. The results of investigations show, that the marked distinctions between two processes can influence (in definite degree) as on property of synthesized materials, as on behavior of platinoid during vitrifications

  7. Compilation of information on modeling of inductively heated cold crucible melters

    International Nuclear Information System (INIS)

    Lessor, D.L.

    1996-03-01

    The objective of this communication, Phase B of a two-part report, is to present information on modeling capabilities for inductively heated cold crucible melters, a concept applicable to waste immobilization. Inductively heated melters are those in which heat is generated using coils around, rather than electrodes within, the material to be heated. Cold crucible or skull melters are those in which the melted material is confined within unmelted material of the same composition. This phase of the report complements and supplements Phase A by Loren Eyler, specifically by giving additional information on modeling capabilities for the inductively heated melter concept. Eyler discussed electrically heated melter modeling capabilities, emphasizing heating by electrodes within the melt or on crucible walls. Eyler also discussed requirements and resources for the computational fluid dynamics, heat flow, radiation effects, and boundary conditions in melter modeling; the reader is referred to Eyler's discussion of these. This report is intended for use in the High Level Waste (HLW) melter program at Hanford. We sought any modeling capabilities useful to the HLW program, whether through contracted research, code license for operation by Department of Energy laboratories, or existing codes and modeling expertise within DOE

  8. Current status of the active test at RRP and development programs for the advanced melter

    International Nuclear Information System (INIS)

    Kanehira, Norio

    2016-01-01

    The vitrification facility in Rokkasho Reprocessing Plant started the active tests to solidify HAW into the glass in 2007 which was the examination of the final stage before the operation, but the active test had to be discontinued due to the trouble of glass melter operation with down of pouring by deposit of noble metals on the melter bottom. After the equipment and operating conditions were improved in response to the result of the mock-up tests, a series of active tests were restarted active tests in May, 2012. These tests were finished with enough confirmation of stability in the state such as glass temperature and controlling the noble metals. JNFL has been developed the advanced melter, Joule heated ceramic melter, and the design of the advanced melter is largely different from the existing one. For the confirmation of the advanced melter performances, the full-scale inactive tests had been performed and successfully finished. This paper describes outline of development for advanced melter in Rokkasho Reprocessing Plant. (author)

  9. Final Report Integrated DM1200 Melter Testing Of Bubbler Configurations Using HLW AZ-101 Simulants VSL-04R4800-4, Rev. 0, 10/5/04

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Gong, W.; Bardakci, T.; D'Angelo, N.A.; Lutze, W.; Callow, R.A.; Brandys, M.; Kot, W.K.; Pegg, I.L.

    2011-01-01

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 HLW simulants. The tests reported herein are a subset of six tests from a larger series of tests described in the Test Plan for the work; results from the other tests have been reported separately. The solids contents of the melter feeds were based on the WTP baseline value for the solids content of the feeds from pretreatment which changed during these tests from 20% to 15% undissolved solids resulting in tests conducted at two feed solids contents. Based on the results of earlier tests with single outlet 'J' bubblers, initial tests were performed with a total bubbling rate of 651 pm. The first set of tests (Tests 1A-1E) addressed the effects of skewing this total air flow rate back and forth between the two installed bubblers in comparison to a fixed equal division of flow between them. The second set of tests (2A-2D) addressed the effects of bubbler depth. Subsequently, as the location, type and number of bubbling outlets were varied, the optimum bubbling rate for each was determined. A third (3A-3C) and fourth (8A-8C) set of tests evaluated the effects of alternative bubbler designs with two gas outlets per bubbler instead of one by placing four bubblers in positions simulating multiple-outlet bubblers. Data from the simulated multiple outlet bubblers were used to design bubblers with two outlets for an additional set of tests (9A-9C). Test 9 was also used to determine the effect of small sugar additions to the feed on ruthenium volatility. Another set of tests (10A-10D) evaluated the effects on production rate of spiking the feed with chloride and sulfate. Variables held constant to the extent possible included melt temperature, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The feed rate was increased to the point that a constant, essentially complete, cold cap was achieved

  10. FINAL REPORT INTEGRATED DM1200 MELTER TESTING OF BUBBLER CONFIGURATIONS USING HLW AZ-101 SIMULANTS VSL-04R4800-4 REV 0 10/5/04

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; LUTZE W; CALLOW RA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 HLW simulants. The tests reported herein are a subset of six tests from a larger series of tests described in the Test Plan for the work; results from the other tests have been reported separately. The solids contents of the melter feeds were based on the WTP baseline value for the solids content of the feeds from pretreatment which changed during these tests from 20% to 15% undissolved solids resulting in tests conducted at two feed solids contents. Based on the results of earlier tests with single outlet 'J' bubblers, initial tests were performed with a total bubbling rate of 651 pm. The first set of tests (Tests 1A-1E) addressed the effects of skewing this total air flow rate back and forth between the two installed bubblers in comparison to a fixed equal division of flow between them. The second set of tests (2A-2D) addressed the effects of bubbler depth. Subsequently, as the location, type and number of bubbling outlets were varied, the optimum bubbling rate for each was determined. A third (3A-3C) and fourth (8A-8C) set of tests evaluated the effects of alternative bubbler designs with two gas outlets per bubbler instead of one by placing four bubblers in positions simulating multiple-outlet bubblers. Data from the simulated multiple outlet bubblers were used to design bubblers with two outlets for an additional set of tests (9A-9C). Test 9 was also used to determine the effect of small sugar additions to the feed on ruthenium volatility. Another set of tests (10A-10D) evaluated the effects on production rate of spiking the feed with chloride and sulfate. Variables held constant to the extent possible included melt temperature, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The feed rate was increased to the point that a constant, essentially complete, cold cap was

  11. Design features of the radioactive Liquid-Fed Ceramic Melter system

    International Nuclear Information System (INIS)

    Holton, L.K. Jr.

    1985-06-01

    During 1983, the Pacific Northwest Laboratory (PNL), at the request of the Department of Energy (DOE), undertook a program with the principal objective of testing the Liquid-Fed Ceramic Melter (LFCM) process in actual radioactive operations. This activity, termed the Radioactive LFCM (RLFCM) Operations is being conducted in existing shielded hot-cell facilities in B-Cell of the 324 Building, 300 Area, located at Hanford, Washington. This report summarizes the design features of the RLFCM system. These features include: a waste preparation and feed system which uses pulse-agitated waste preparation tanks for waste slurry agitation and an air displacement slurry pump for transferring waste slurries to the LFCM; a waste vitrification system (LFCM) - the design features, design approach, and reasoning for the design of the LFCM are described; a canister-handling turntable for positioning canisters underneath the RLFCM discharge port; a gamma source positioning and detection system for monitoring the glass fill level of the product canisters; and a primary off-gas treatment system for removing the majority of the radionuclide contamination from the RLFCM off gas. 8 refs., 48 figs., 6 tabs

  12. An evaluation of foaming potential in the IDMS melter

    International Nuclear Information System (INIS)

    Hutson, N.D.

    1992-01-01

    The present DWPF flowsheet calls for the chemical treatment of waste sludge with 90 wt% formic acid prior to the addition of the Precipitate Hydrolysis Aqueous (PHA) product. An alternative processing methodology, denoted the ''Nitric Acid Flowsheet'', has been proposed. In the application of this flowsheet, nitric acid would be used to neutralize sludge base components (hydroxides and carbonates) prior to the addition of late wash PHA. The late wash PHA will contain sufficient quantities of formic acid to adequately complete necessary reduction-oxidation (REDOX) reactions. The use of this flowsheet may result in a change in the nominal concentrations of two of the major REDOX reaction participants: formate (HCOO minus ) and nitrate (NO 3 minus )

  13. Application of the HWVP measurement error model and feed test algorithms to pilot scale feed testing

    International Nuclear Information System (INIS)

    Adams, T.L.

    1996-03-01

    The purpose of the feed preparation subsystem in the Hanford Waste Vitrification Plant (HWVP) is to provide, for control of the properties of the slurry that are sent to the melter. The slurry properties are adjusted so that two classes of constraints are satisfied. Processability constraints guarantee that the process conditions required by the melter can be obtained. For example, there are processability constraints associated with electrical conductivity and viscosity. Acceptability constraints guarantee that the processed glass can be safely stored in a repository. An example of an acceptability constraint is the durability of the product glass. The primary control focus for satisfying both processability and acceptability constraints is the composition of the slurry. The primary mechanism for adjusting the composition of the slurry is mixing the waste slurry with frit of known composition. Spent frit from canister decontamination is also recycled by adding it to the melter feed. A number of processes in addition to mixing are used to condition the waste slurry prior to melting, including evaporation and the addition of formic acid. These processes also have an effect on the feed composition

  14. Startup of a Joule-heated glass melter with a graphite slurry

    International Nuclear Information System (INIS)

    Allen, T.L.; Routt, K.R.; Porter, M.A.

    1983-01-01

    This paper discusses the theoretical equations and physical and electrical property data of various graphite slurries for starting up a glass melter. An application test is also included to demonstrate the graphite slurry startup technique

  15. Preliminary evaluation of PSCM and BIPP melter design and operating conditions using physical modeling

    International Nuclear Information System (INIS)

    Skarda, R.J.; Hauser, S.G.; Fort, J.A.

    1985-05-01

    The Glass Melter Physical Modeling investigation was initiated to support Pacific Northwest Laboratory (PNL) Hanford Waste Vitrification Program. Specifically, results discussed herein are those of the modeled B-Plant Immobilization Pilot Plant (BIPP) and Pilot Scale Ceramic Melter (PSCM) designs. The purpose of this study was to evaluate various melter design features using laboratory scale models. Hydrodynamic, thermal, and electrical similarity between the modeling fluid and the molten glass were primary objectives. Stroboscopic velocity measurements (flow visualization), temperature measurements, and electrical potential measurements were used to investigate the molten glass behavior. Results from this effort are to provide input to melter design and proposed operation in addition to providing a data base for verifying numerical models. 13 refs., 48 figs., 24 tabs

  16. Electrical power supply and controls for a remotely operated glass melter for nuclear waste

    International Nuclear Information System (INIS)

    Haideri, A.Q.

    1985-01-01

    An electrical power supply, controls and instruments used for a joule heated glass melter for nuclear waste are discussed. Remotely replaceable interconnection wiring assemblies for power, controls and instruments are also described

  17. Investigation of corrosion experienced in a spray calciner/ceramic melter vitrification system

    International Nuclear Information System (INIS)

    Dierks, R.D.; Mellinger, G.B.; Miller, F.A.; Nelson, T.A.; Bjorklund, W.J.

    1980-08-01

    After periodic testing of a large-scale spray calciner/ceramic melter vitrification system over a 2-yr period, sufficient corrosion was noted on various parts of the vitrification system to warrant its disassembly and inspection. A majority of the 316 SS sintered metal filters on the spray calciner were damaged by chemical corrosion and/or high temperature oxidation. Inconel-601 portions of the melter lid were attacked by chlorides and sulfates which volatilized from the molten glass. The refractory blocks, making up the walls of the melter, were attacked by the waste glass. This attack was occurring when operating temperatures were >1200 0 C. The melter floor was protected by a sludge layer and showed no corrosion. Corrosion to the Inconel-690 electrodes was minimal, and no corrosion was noted in the offgas treatment system downstream of the sintered metal filters. It is believed that most of the melter corrosion occurred during one specific operating period when the melter was operated at high temperatures in an attempt to overcome glass foaming behavior. These high temperatures resulted in a significant release of volatile elements from the molten glass, and also created a situation where the glass was very fluid and convective, which increased the corrosion rate of the refractories. Specific corrosion to the calciner components cannot be proven to have occurred during a specific time period, but the mechanisms of attack were all accelerated under the high-temperature conditions that were experienced with the melter. A review of the materials of construction has been made, and it is concluded that with controlled operating conditions and better protection of some materials of construction corrosion of these systems will not cause problems. Other melter systems operating under similar strenuous conditions have shown a service life of 3 yr

  18. Development of equipments for remote dismantling of joule heated ceramic melter

    International Nuclear Information System (INIS)

    Badgujar, Kiran T.; Usarkar, Sachin G.; Kumar, Binu; Nair, K.N.S.

    2011-01-01

    Joule Heated Ceramic Melter (JHCM) technology has been adopted for industrial scale vitrification of high level liquid waste (HLLW) at Tarapur and Kalpakkam. The melter installed at Advanced Vitrification System (AVS), Tarapur has immobilized 175 m 3 of HLLW in 113 canisters containing 11533Kg of Vitrified Waste Product (VWP). The melter has been in operation for 3 years before shutdown. It is intended to demonstrate the complete procedure of dismantling of Joule Melter in 1:1 scale prior to going in for actual dismantling in the hot cell. The Melter consists of an assembly of Inconel/SS pipes and plates, fuse cast refractories, thermal insulations of various types inside a SS casing and possibly some glass which is left over in the melter. Dismantling of melter involves remote cutting of the outer casing, pipe connections, electrical connections and removal, sizing and packing of internals in a sequential manner to minimise generation of secondary waste. The challenge involves development of remotely operated multi-degrees of freedom fixtures, modification and performance testing of standard industrial cutting and breaking tools and adapting them for remote operations. The work also involves development of equipments for collection of waste generated during the dismantling operation and packaging thus in special packages. Remotely actuated fixtures have been developed for remote top plate and side electrodes cutting. Remotely operated grab has been developed for handling of loose material and grippers have been developed for handling of refractory blocks. Industrial vacuum suction device has been modified into split units to enable for reducing the spread of powder material, while dismantling in progress. The performance test of developed fixtures, equipments, cutting and breaking tools have been carried on 1:1 scale melter model. Various parameters like cutting speed, cutting tool performance, generation of waste volume has been measured and analysed for

  19. Settling of Spinel in A High-Level Waste Glass Melter

    International Nuclear Information System (INIS)

    Pavel Hrma; Pert Schill; Lubomir Nemec

    2002-01-01

    High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors call melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150 degree C (or even higher in advanced melters) to create a melt that becomes glass on cooling. This process is slow and expensive. Moreover, the melters that are currently in use or are going to be used in the U.S. are sensitive to clogging and thus cannot process melt in which solid particles are suspended. These particles settle and gradually accumulate on the melter bottom. Such particles, most often small crystals of spinel ( a mineral containing iron, nickel, chromium, and other minor oxides), inevitably occurred in the melt when the content of the waste in the glass (called waste loading) increases above a certain limit. To avoid the presence of solid particles in the melter, the waste loading is kept rather low, in average 15% lower than in glass formulated for more robust melters

  20. Hazards analyses of hydrogen evolution and ammonium nitrate accumulation in DWPF -- Revision 1

    International Nuclear Information System (INIS)

    Holtzscheiter, E.W.

    1994-01-01

    This revision consists of two reports, the first of which is an analysis of potential ammonium nitrate explosion hazards in the DWPF (Defense Waste Processing Facility). Sections describe the effect of impurities (organic and inorganic (chlorides, chromates, metals and oxides)); the consequences of a hydrogen deflagration or detonation; the role of confinement; the action of heat on ammonium nitrate; the thermal decomposition of ammonium nitrate; the hazard of spontaneous heating; and the explosive decomposition of ammonium nitrate. The second report, Hazard analysis of hydrogen evolution in DWPF: Process vessels and vent system for the late wash/nitric acid flowsheet, contains a description of a revised model for hydrogen generation based on the late wash/nitric acid process. The second part of the report is a sensitivity analysis of the base case conditions and the hydrogen generation model

  1. Material Compatibility Evaluation for DWPF Nitric-Glycolic Acid - Literature Review

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Skidmore, T. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-30

    Glycolic acid is being evaluated as an alternative for formic and nitric acid in the DWPF flowsheet. Demonstration testing and modeling for this new flowsheet has shown that glycolic acid and glycolate has a potential to remain in certain streams generated during the production of the nuclear waste glass. A literature review was conducted to assess the impact of glycolic acid on the corrosion of the materials of construction for the DWPF facility as well as facilities downstream which may have residual glycolic acid and glycolates present. The literature data was limited to solutions containing principally glycolic acid. The reported corrosion rates and degradation characteristics have shown the following for the materials of construction.

  2. Projected radionuclide inventories of DWPF glass from current waste at time of production

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1993-01-01

    The Waste Acceptance Preliminary Specifications (WAPS) require that the DWPF estimate the inventory of long-lived radionuclides present in the waste glass, and report the values in the Waste Form Qualification Report. In this report, conservative (biased high) estimates of the radionuclide inventory of glass produced from waste currently in the Tank Farm are provided. In most cases, these calculated values compare favorably with actual data. In those cases where the agreement is not good, the values reported here are conservative

  3. Nitric-glycolic flowsheet evaluation with the slurry-fed melt rate furnace

    Energy Technology Data Exchange (ETDEWEB)

    Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Miller, D. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fowley, M. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Zamecnik, J. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-03-01

    The Savannah River National Laboratory (SRNL) was tasked to support validation of the Defense Waste Processing Facility (DWPF) melter offgas flammability model for the nitric-glycolic (NG) flowsheet. The work supports Deliverable 4 of the DWPF & Saltstone Facility Engineering Technical Task Request (TTR)1 and is supplemental to the Cold Cap Evaluation Furnace (CEF) testing conducted in 2014.2 The Slurry-fed Melt Rate Furnace (SMRF) was selected for the supplemental testing as it requires significantly less resources than the CEF and could provide a tool for more rapid analysis of melter feeds in the future. The SMRF platform has been used previously to evaluate melt rate behavior of DWPF glasses, but was modified to accommodate analysis of the offgas stream. Additionally, the Melt Rate Furnace (MRF) and Quartz Melt Rate Furnace (QMRF) were utilized for evaluations. MRF data was used exclusively for melt behavior observations and REDuction/OXidation (REDOX) prediction comparisons and will be briefly discussed in conjunction with its support of the SMRF testing. The QMRF was operated similarly to the SMRF for the same TTR task, but will be discussed in a separate future report. The overall objectives of the SMRF testing were to; 1) Evaluate the efficacy of the SMRF as a platform for steady state melter testing with continuous feeding and offgas analysis; and 2) Generate supplemental melter offgas flammability data to support the melter offgas flammability modelling effort for DWPF implementation of the NG flowsheet.

  4. The product composition control system at Savannah River: Statistical process control algorithm

    International Nuclear Information System (INIS)

    Brown, K.G.

    1994-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) will be used to immobilize the approximately 130 million liters of high-level nuclear waste currently stored at the site in 51 carbon steel tanks. Waste handling operations separate this waste into highly radioactive insoluble sludge and precipitate and less radioactive water soluble salts. In DWPF, precipitate (PHA) is blended with insoluble sludge and ground glass frit to produce melter feed slurry which is continuously fed to the DWPF melter. The melter produces a molten borosilicate glass which is poured into stainless steel canisters for cooling and, ultimately, shipment to and storage in an geologic repository. Described here is the Product Composition Control System (PCCS) process control algorithm. The PCCS is the amalgam of computer hardware and software intended to ensure that the melt will be processable and that the glass wasteform produced will be acceptable. Within PCCS, the Statistical Process Control (SPC) Algorithm is the means which guides control of the DWPF process. The SPC Algorithm is necessary to control the multivariate DWPF process in the face of uncertainties arising from the process, its feeds, sampling, modeling, and measurement systems. This article describes the functions performed by the SPC Algorithm, characterization of DWPF prior to making product, accounting for prediction uncertainty, accounting for measurement uncertainty, monitoring a SME batch, incorporating process information, and advantages of the algorithm. 9 refs., 6 figs

  5. Leaching TC-99 from DWPF glass in simulated geologic repository groundwaters

    International Nuclear Information System (INIS)

    Bibler, N.E.; Jurgensen, A.R.

    1986-01-01

    The purpose was to determine if DWPF glass in geologic groundwaters would immobilize Tc-99 as well as it does other elements. A previous study (using a borosilicate glass of a very different composition from DWPF glass) indicated that Tc-99 leached rapidly from the glass suggesting that glass may not be a good matrix for immobilizing Tc-99. It was suggested that the Tc-99 had migrated to vesicles in the glass while the glass was still molten. To determine if borosilicate glass was a good immobilizing matrix for Tc-99, this study was performed using DWPF glass. The leaching of Tc-99 was compared to other elements in the glass. It was shown that rapid leaching will not occur with SRP glass. The leach rate for Tc-99 was nearly identical to that for B, a matrix element in the glass. Another objective was to compare the release of Tc-99 under oxidizing and reducing conditions with other elements in the glass. In the tests described here, even though the glass was dissolving more under reducing conditions as a result of abnormally high pH values, less Tc-99 appeared in solution

  6. BEHAVIOR OF MERCURY DURING DWPF CHEMICAL PROCESS CELL PROCESSING

    Energy Technology Data Exchange (ETDEWEB)

    Zamecnik, J.; Koopman, D.

    2012-04-09

    retention of mercury in the slurry. Both recovery of mercury in the offgas system and removal (segregation + recovery) from the slurry correlate with slurry consistency. Higher slurry consistency results in better retention of Hg in the slurry (less segregation) and better recovery in the offgas system, but the relationships of recovery and retention with consistency are sludge dependent. Some correlation with slurry yield stress and acid stoichiometry was also found. Better retention of mercury in the slurry results in better recovery in the offgas system because the mercury in the slurry is stripped more easily than the segregated mercury at the bottom of the vessel. Although better retention gives better recovery, the time to reach a particular slurry mercury content (wt%) is longer than if the retention is poorer because the segregation is faster. The segregation of mercury is generally a faster process than stripping. The stripping factor (mass of water evaporated per mass of mercury stripped) of mercury at the start of boiling were found to be less than 1000 compared to the assumed design basis value of 750 (the theoretical factor is 250). However, within two hours, this value increased to at least 2000 lb water per lb Hg. For runs with higher mercury recovery in the offgas system, the stripping factor remained around 2000, but runs with low recovery had stripping factors of 4000 to 40,000. DWPF data shows similar trends with the stripping factor value increasing during boiling. These high values correspond to high segregation and low retention of mercury in the sludge. The stripping factor for a pure Hg metal bead in water was found to be about 10,000 lb/lb. About 10-36% of the total Hg evaporated in a SRAT cycle was refluxed back to the SRAT during formic acid addition and boiling. Mercury is dissolved as a result of nitric acid formation from absorption of NO{sub x}. The actual solubility of dissolved mercury in the acidic condensate is about 100 times higher than

  7. Technetium Retention In WTP Law Glass With Recycle Flow-Sheet DM10 Melter Testing VSL-12R2640-1 REV 0

    International Nuclear Information System (INIS)

    Abramowitz, Howard; Callow, Richard A.; Joseph, Innocent

    2012-01-01

    Melter tests were conducted to determine the retention of technetium and other volatiles in glass while processing simulated Low Activity Waste (LAW) streams through a DM10 melter equipped with a prototypical off-gas system that concentrates and recycles fluid effiuents back to the melter feed. To support these tests, an existing DM10 system installed at Vitreous State Laboratory (VSL) was modified to add the required recycle loop. Based on the Hanford Tank Waste Treatment and Immobilization Plant (WTP) LAW off-gas system design, suitably scaled versions of the Submerged Bed Scrubber (SBS), Wet Electrostatic Precipitator (WESP), and TLP vacuum evaporator were designed, built, and installed into the DM10 system. Process modeling was used to support this design effort and to ensure that issues associated with the short half life of the 99m Tc radioisotope that was used in this work were properly addressed and that the system would be capable of meeting the test objectives. In particular, this required that the overall time constant for the system was sufficiently short that a reasonable approach to steady state could be achieved before the 99m Tc activity dropped below the analytical limits of detection. The conceptual design, detailed design, flow sheet development, process model development, Piping and Instrumentation Diagram (P and ID) development, control system design, software design and development, system fabrication, installation, procedure development, operator training, and Test Plan development for the new system were all conducted during this project. The new system was commissioned and subjected to a series of shake-down tests before embarking on the planned test program. Various system performance issues that arose during testing were addressed through a series of modifications in order to improve the performance and reliability of the system. The resulting system provided a robust and reliable platform to address the test objectives

  8. TECHNETIUM RETENTION IN WTP LAW GLASS WITH RECYCLE FLOW-SHEET DM10 MELTER TESTING VSL-12R2640-1 REV 0

    Energy Technology Data Exchange (ETDEWEB)

    Abramowitz, Howard [Catholic Univ. of America, Washington, DC (United States). Vitreous State Lab.; Brandys, Marek [Catholic Univ. of America, Washington, DC (United States). Vitreous State Lab.; Cecil, Richard [Catholic Univ. of America, Washington, DC (United States). Vitreous State Lab.; D& #x27; Angelo, Nicholas [Catholic Univ. of America, Washington, DC (United States). Vitreous State Lab.; Matlack, Keith S. [Catholic Univ. of America, Washington, DC (United States). Vitreous State Lab.; Muller, Isabelle S. [Catholic Univ. of America, Washington, DC (United States). Vitreous State Lab.; Pegg, Ian L. [Energy Solutions, Federal EPC, Inc., Columbia, MD (United States); Callow, Richard A. [Energy Solutions, Federal EPC, Inc., Columbia, MD (United States); Joseph, Innocent

    2012-12-11

    Melter tests were conducted to determine the retention of technetium and other volatiles in glass while processing simulated Low Activity Waste (LAW) streams through a DM10 melter equipped with a prototypical off-gas system that concentrates and recycles fluid effiuents back to the melter feed. To support these tests, an existing DM10 system installed at Vitreous State Laboratory (VSL) was modified to add the required recycle loop. Based on the Hanford Tank Waste Treatment and Immobilization Plant (WTP) LAW off-gas system design, suitably scaled versions of the Submerged Bed Scrubber (SBS), Wet Electrostatic Precipitator (WESP), and TLP vacuum evaporator were designed, built, and installed into the DM10 system. Process modeling was used to support this design effort and to ensure that issues associated with the short half life of the {sup 99m}Tc radioisotope that was used in this work were properly addressed and that the system would be capable of meeting the test objectives. In particular, this required that the overall time constant for the system was sufficiently short that a reasonable approach to steady state could be achieved before the {sup 99m}Tc activity dropped below the analytical limits of detection. The conceptual design, detailed design, flow sheet development, process model development, Piping and Instrumentation Diagram (P&ID) development, control system design, software design and development, system fabrication, installation, procedure development, operator training, and Test Plan development for the new system were all conducted during this project. The new system was commissioned and subjected to a series of shake-down tests before embarking on the planned test program. Various system performance issues that arose during testing were addressed through a series of modifications in order to improve the performance and reliability of the system. The resulting system provided a robust and reliable platform to address the test objectives.

  9. Bench-scale arc melter for R&D in thermal treatment of mixed wastes

    Energy Technology Data Exchange (ETDEWEB)

    Kong, P.C.; Grandy, J.D.; Watkins, A.D.; Eddy, T.L.; Anderson, G.L.

    1993-05-01

    A small dc arc melter was designed and constructed to run bench-scale investigations on various aspects of development for high-temperature (1,500-1,800{degrees}C) processing of simulated transuranic-contaminated waste and soil located at the Radioactive Waste Management Complex (RWMC). Several recent system design and treatment studies have shown that high-temperature melting is the preferred treatment. The small arc melter is needed to establish techniques and procedures (with surrogates) prior to using a similar melter with the transuranic-contaminated wastes in appropriate facilities at the site. This report documents the design and construction, starting and heating procedures, and tests evaluating the melter`s ability to process several waste types stored at the RWMC. It is found that a thin graphite strip provides reliable starting with initial high current capability for partially melting the soil/waste mixture. The heating procedure includes (1) the initial high current-low voltage mode, (2) a low current-high voltage mode that commences after some slag has formed and arcing dominates over the receding graphite conduction path, and (3) a predominantly Joule heating mode during which the current can be increased within the limits to maintain relatively quiescent operation. Several experiments involving the melting of simulated wastes are discussed. Energy balance, slag temperature, and electrode wear measurements are presented. Recommendations for further refinements to enhance its processing capabilities are identified. Future studies anticipated with the arc melter include waste form processing development; dissolution, retention, volatilization, and collection for transuranic and low-level radionuclides, as well as high vapor pressure metals; electrode material development to minimize corrosion and erosion; refractory corrosion and/or skull formation effects; crucible or melter geometry; metal oxidation; and melt reduction/oxidation (redox) conditions.

  10. An Evaluation of Liquidus Temperature as a Function of Waste Loading for a Tank 42 ''Sludge Only''/Frit 200 Flowsheet

    International Nuclear Information System (INIS)

    Peeler, D.

    1999-01-01

    'The waste glass produced in the SRS Defense Waste Processing Faiclity (DWPF) process must comply with Waste Acceptance Product Specifications (WAPS) and process control requirements by demonstrating, to a high degree of confidence, that melter feed will produce glass satisfying all quality and processing requirements.'

  11. Defense Waste Processing Facility (DWPF), Modular CSSX Unit (CSSX), and Waste Transfer Line System of Salt Processing Program (U)

    International Nuclear Information System (INIS)

    CHANG, ROBERT

    2006-01-01

    All of the waste streams from ARP, MCU, and SWPF processes will be sent to DWPF for vitrification. The impact these new waste streams will have on DWPF's ability to meet its canister production goal and its ability to support the Salt Processing Program (ARP, MCU, and SWPF) throughput needed to be evaluated. DWPF Engineering and Operations requested OBU Systems Engineering to evaluate DWPF operations and determine how the process could be optimized. The ultimate goal will be to evaluate all of the Liquid Radioactive Waste (LRW) System by developing process modules to cover all facilities/projects which are relevant to the LRW Program and to link the modules together to: (1) study the interfaces issues, (2) identify bottlenecks, and (3) determine the most cost effective way to eliminate them. The results from the evaluation can be used to assist DWPF in identifying improvement opportunities, to assist CBU in LRW strategic planning/tank space management, and to determine the project completion date for the Salt Processing Program

  12. HLW Melter Control Strategy Without Visual Feedback VSL-12R2500-1 Rev 0

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, A A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Joseph, Innocent [The Catholic University of America, Washington, DC (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States); Callow, Richard A. [The Catholic University of America, Washington, DC (United States); Abramowitz, Howard [The Catholic University of America, Washington, DC (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Brandys, Marek [The Catholic University of America, Washington, DC (United States); Kot, Wing K. [The Catholic University of America, Washington, DC (United States)

    2012-11-13

    Plans for the treatment of high level waste (HL W) at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) are based upon the inventory of the tank wastes, the anticipated performance of the pretreatment processes, and current understanding of the capability of the borosilicate glass waste form [I]. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat and mass transfer and increase glass melting rates. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of ~ 1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HL W waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150°C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage.

  13. Enhancement of the life of refractories through the operational experience of plasma torch melter

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Young Pyo [Technology Institute, Korea Radioactive waste Agency (KORAD), Daejeon (Korea, Republic of); Choi, Jaang Young [Chungnam National University, Daejeon (Korea, Republic of)

    2016-06-15

    The properties of wastes for melting need to be considered to minimize the maintenance of refractory and to discharge the molten slags smoothly from a plasma torch melter. When the nonflammable wastes from nuclear facilities such as concrete debris, glass, sand, etc., are melted, they become acid slags with low basicity since the chemical composition has much more acid oxides than basic oxides. A molten slag does not have good characteristics of discharge and is mainly responsible for the refractory erosion due to its low liquidity. In case of a stationary plasma torch melter with a slant tapping port on the wall, a fixed amount of molten slags remains inside of tapping hole as well as the melter inside after tapping out. Nonmetallic slags keep the temperature higher than melting point of metal because metallic slags located on the bottom of melter by specific gravity difference are simultaneously melted when dual mode plasma torch operates in transferred mode. In order to minimize the refractory erosion, the compatible refractories are selected considering the temperature inside the melter and the melting behavior of slags whether to contact or noncontact with molten slags. An acidic refractory shall not be installed in adjacent to a basic refractory for the resistibility against corrosion.

  14. HLW Melter Control Strategy Without Visual Feedback VSL-12R2500-1 Rev 0

    International Nuclear Information System (INIS)

    Kruger, A A.; Joseph, Innocent; Matlack, Keith S.; Callow, Richard A.; Abramowitz, Howard; Pegg, Ian L.; Brandys, Marek; Kot, Wing K.

    2012-01-01

    Plans for the treatment of high level waste (HL W) at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) are based upon the inventory of the tank wastes, the anticipated performance of the pretreatment processes, and current understanding of the capability of the borosilicate glass waste form [I]. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat and mass transfer and increase glass melting rates. The WTP HLW melter has a glass surface area of 3.75 m 2 and depth of ∼ 1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HL W waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150°C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage

  15. Safety assessment of the liquid-fed ceramic melter process

    International Nuclear Information System (INIS)

    Buelt, J.L.; Partain, W.L.

    1980-08-01

    As part of its development program for the solidification of high-level nuclear waste, Pacific Northwest Laboratory assessed the safety issues for a complete liquid-fed ceramic melter (LFCM) process. The LFCM process, an adaption of commercial glass-making technology, is being developed to convert high-level liquid waste from the nuclear fuel cycle into glass. This safety assessment uncovered no unresolved or significant safety problems with the LFCM process. Although in this assessment the LFCM process was not directly compared with other solidification processes, the safety hazards of the LFCM process are comparable to those of other processes. The high processing temperatures of the glass in the LFCM pose no additional significant safety concerns, and the dispersible inventory of dried waste (calcine) is small. This safety assessment was based on the nuclear power waste flowsheet, since power waste is more radioactive than defense waste at the time of solidification, and all accident conditions for the power waste would have greater radiological consequences than those for defense waste. An exhaustive list of possible off-standard conditions and equipment failures was compiled. These accidents were then classified according to severity of consequence and type of accident. Radionuclide releases to the stack were calculated for each group of accidents using conservative assumptions regarding the retention and decontamination features of the process and facility. Two recommendations that should be considered by process designers are given in the safety assessment

  16. Demonstration test of 'multi-purpose incinerating melter system'

    International Nuclear Information System (INIS)

    Miyazaki, Hitoshi; Tanimoto, Kenichi; Wakui, Hitoshi; Oasada, Kaoru; Ishikawa, Fuyuhiko.

    1994-01-01

    A Multi-Purpose Incinerating Melter System (MIMS) has been developed as a volume reduction technique for a wide variety of radwastes including flame retardants such as spent resin, and non-combustible materials such as concrete, glass and steel. In the MIMS, these wastes are incinerated and/or melted at temperatures between 1,000 and 1,500degC generated by fossil fueled burner to produce obsidian-like ingots with high integrity. A demonstration test program was carried out from 1989 until 1991 using an engineering-scale demonstration unit. In the test program, various simulated wastes with traces of 60 Co, 54 Mn, 59 Fe, 137 Cs, 22 Na and 106 Ru were treated to obtain decontamination factor (DF) data and leach-resistance data of the products. The summarized results drawn from the 13 runs of demonstrative operations are the following: (1) Most involatile radionuclides are transferred into solidified products. (2) Global DF of the system excluding a HEPA filter ranged 1x10 4 thru 1x10 5 for 60 Co, 2x10 2 thru 2x10 3 for 137 Cs and 2x10 2 thru 1x10 4 for 106 Ru. (3) Leaching resistance of the solidified product is a match for that of a typical borosilicate glass waste form. (author)

  17. Demonstration test of 'multi-purpose incinerating melter system'

    Energy Technology Data Exchange (ETDEWEB)

    Miyazaki, Hitoshi; Tanimoto, Kenichi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Wakui, Hitoshi; Oasada, Kaoru; Ishikawa, Fuyuhiko

    1994-03-01

    A Multi-Purpose Incinerating Melter System (MIMS) has been developed as a volume reduction technique for a wide variety of radwastes including flame retardants such as spent resin, and non-combustible materials such as concrete, glass and steel. In the MIMS, these wastes are incinerated and/or melted at temperatures between 1,000 and 1,500degC generated by fossil fueled burner to produce obsidian-like ingots with high integrity. A demonstration test program was carried out from 1989 until 1991 using an engineering-scale demonstration unit. In the test program, various simulated wastes with traces of [sup 60]Co, [sup 54]Mn, [sup 59]Fe, [sup 137]Cs, [sup 22]Na and [sup 106]Ru were treated to obtain decontamination factor (DF) data and leach-resistance data of the products. The summarized results drawn from the 13 runs of demonstrative operations are the following: (1) Most involatile radionuclides are transferred into solidified products. (2) Global DF of the system excluding a HEPA filter ranged 1x10[sup 4] thru 1x10[sup 5] for [sup 60]Co, 2x10[sup 2] thru 2x10[sup 3] for [sup 137]Cs and 2x10[sup 2] thru 1x10[sup 4] for [sup 106]Ru. (3) Leaching resistance of the solidified product is a match for that of a typical borosilicate glass waste form. (author).

  18. The University of Missouri Research Reactor facility can melter system

    International Nuclear Information System (INIS)

    Edwards, C.B. Jr.; Olson, O.L.; Stevens, R.; Brugger, R.M.

    1987-01-01

    At the University of Missouri Research Reactor (MURR), a waste compacting system for reducing the volume of radioactive aluminum cans has been designed, built and put into operation. In MURR's programs of producing radioisotopes and transmutation doping of silicon, a large volume of radioactive aluminum cans is generated. The Can Melter System (CMS) consists of a sorting station, a can masher, an electric furnace and a gas fired furnace. This system reduces the cans and other radioactive metal into barrels of solid metal close to theoretical density. The CMS has been in operation at the MURR now for over two years. Twelve hundred cu ft of cans and other metals have been reduced into 150 cu ft of shipable waste. The construction cost of the CMS was $4950.84 plus 1680 man hours of labor, and the operating cost of the CMS is $18/lb. The radiation exposure to the operator is 8.6 mR/cu ft. The yearly operating savings is $30,000. 20 figs., 10 tabs

  19. Modeling principles applied to the simulation of a joule-heated glass melter

    International Nuclear Information System (INIS)

    Routt, K.R.

    1980-05-01

    Three-dimensional conservation equations applicable to the operation of a joule-heated glass melter were rigorously examined and used to develop scaling relationships for modeling purposes. By rigorous application of the conservation equations governing transfer of mass, momentum, energy, and electrical charge in three-dimensional cylindrical coordinates, scaling relationships were derived between a glass melter and a physical model for the following independent and dependent variables: geometrical size (scale), velocity, temperature, pressure, mass input rate, energy input rate, voltage, electrode current, electrode current flux, total power, and electrical resistance. The scaling relationships were then applied to the design and construction of a physical model of the semiworks glass melter for the Defense Waste Processing Facility. The design and construction of such a model using glycerine plus LiCl as a model fluid in a one-half-scale Plexiglas tank is described

  20. Volatilization and redox testing in a DC arc melter: FY-93 and FY-94

    International Nuclear Information System (INIS)

    Grandy, J.D.; Sears, J.W.; Soelberg, N.R.; Reimann, G.A.; McIlwain, M.E.

    1996-07-01

    The purpose of these experiments was to study the dissolution, retention, volatilization, and trapping of transuranic radionuclide elements (TRUs), mixed fission and activation products, and high vapor pressure metals (HVPMS) during processing in a high temperature arc furnace. In all cases, surrogate elements (lanthanides) were used in place of radioactive ones. The experiments were conducted utilizing a small DC arc melter developed at the Idaho National Engineering Laboratory (INEL) Research Center (IRC). The small arc melter was originally developed in 1992 and has been used previously for waste form studies of iron enriched basalt (IEB) and IEB with zirconium and titanium additions (IEB4). Section 3 contains a description of the small arc melter and its operational capabilities are discussed in Chapter 4. The remainder of the document describes each testing program and then discusses results and findings

  1. Predictive modeling of crystal accumulation in high-level waste glass melters processing radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Matyáš, Josef; Gervasio, Vivianaluxa; Sannoh, Sulaiman E.; Kruger, Albert A.

    2017-11-01

    The effectiveness of HLW vitrification is limited by precipitation/accumulation of spinel crystals [(Fe, Ni, Mn, Zn)(Fe, Cr)2O4] in the glass discharge riser of Joule-heated ceramic melters during idling. These crystals do not affect glass durability; however, if accumulated in thick layer, they can clog the melter and prevent discharge of molten glass into canisters. To address this problem, an empirical model was developed that can predict thicknesses of accumulated layers as a function of glass composition. This model predicts well the accumulation of single crystals and/or small-scale agglomerates, but, excessive agglomeration observed in high-Ni-Fe glass resulted in an under-prediction of accumulated layers, which gradually worsen over time as an increased number of agglomerates formed. Accumulation rate of ~53.8 ± 3.7 µm/h determined for this glass will result in ~26 mm thick layer in 20 days of melter idling.

  2. Erosion/corrosion concerns in feed preparation systems at the Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Gee, J.T.; Chandler, C.T.; Daugherty, W.L.; Imrich, K.J.; Jenkins, C.F.

    1997-01-01

    The Savannah River Site (SRS) has been operating a nuclear fuel cycle since the 1950's to produce nuclear materials in support of the national defense effort. The Department of Energy authorized the construction of the Defense Waste Processing Facility (DWPF) to immobilize the high level radioactive waste resulting from these processes as a durable borosilicate glass. The DWPF, after having undergone extensive testing, has been approved for operations and is currently immobilizing radioactive waste. To ensure reliability of the DWPF remote canyon processing equipment, a materials evaluation program was performed prior to radioactive operations to determine to what extent erosion/corrosion would impact design life of equipment. The program consisted of performing pre-service baseline inspections on critical equipment and follow-up inspections after completion of DWPF cold chemical demonstration runs. Non-destructive examination (NDE) techniques were used to assess erosion/corrosion as well as evaluation of corrosion coupon racks. These results were used to arrive at predicted equipment life for selected feed preparation equipment. It was concluded with the exception of the coil and agitator for the slurry mix evaporator (SME), which are exposed to erosive glass frit particles, all of the equipment should meet its design life

  3. Bench-scale arc melter for R ampersand D in thermal treatment of mixed wastes

    International Nuclear Information System (INIS)

    Kong, P.C.; Grandy, J.D.; Watkins, A.D.; Eddy, T.L.; Anderson, G.L.

    1993-05-01

    A small dc arc melter was designed and constructed to run bench-scale investigations on various aspects of development for high-temperature (1,500-1,800 degrees C) processing of simulated transuranic-contaminated waste and soil located at the Radioactive Waste Management Complex (RWMC). Several recent system design and treatment studies have shown that high-temperature melting is the preferred treatment. The small arc melter is needed to establish techniques and procedures (with surrogates) prior to using a similar melter with the transuranic-contaminated wastes in appropriate facilities at the site. This report documents the design and construction, starting and heating procedures, and tests evaluating the melter's ability to process several waste types stored at the RWMC. It is found that a thin graphite strip provides reliable starting with initial high current capability for partially melting the soil/waste mixture. The heating procedure includes (1) the initial high current-low voltage mode, (2) a low current-high voltage mode that commences after some slag has formed and arcing dominates over the receding graphite conduction path, and (3) a predominantly Joule heating mode during which the current can be increased within the limits to maintain relatively quiescent operation. Several experiments involving the melting of simulated wastes are discussed. Energy balance, slag temperature, and electrode wear measurements are presented. Recommendations for further refinements to enhance its processing capabilities are identified. Future studies anticipated with the arc melter include waste form processing development; dissolution, retention, volatilization, and collection for transuranic and low-level radionuclides, as well as high vapor pressure metals; electrode material development to minimize corrosion and erosion; refractory corrosion and/or skull formation effects; crucible or melter geometry; metal oxidation; and melt reduction/oxidation (redox) conditions

  4. Experimental Plan for Crystal Accumulation Studies in the WTP Melter Riser

    Energy Technology Data Exchange (ETDEWEB)

    Miller, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fowley, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-04-28

    This experimental plan defines crystal settling experiments to be in support of the U.S. Department of Energy – Office of River Protection crystal tolerant glass program. The road map for development of crystal-tolerant high level waste glasses recommends that fluid dynamic modeling be used to better understand the accumulation of crystals in the melter riser and mechanisms of removal. A full-scale version of the Hanford Waste Treatment and Immobilization Plant (WTP) melter riser constructed with transparent material will be used to provide data in support of model development. The system will also provide a platform to demonstrate mitigation or recovery strategies in off-normal events where crystal accumulation impedes melter operation. Test conditions and material properties will be chosen to provide results over a variety of parameters, which can be used to guide validation experiments with the Research Scale Melter at the Pacific Northwest National Laboratory, and that will ultimately lead to the development of a process control strategy for the full scale WTP melter. The experiments described in this plan are divided into two phases. Bench scale tests will be used in Phase 1 (using the appropriate solid and fluid simulants to represent molten glass and spinel crystals) to verify the detection methods and analytical measurements prior to their use in a larger scale system. In Phase 2, a full scale, room temperature mockup of the WTP melter riser will be fabricated. The mockup will provide dynamic measurements of flow conditions, including resistance to pouring, as well as allow visual observation of crystal accumulation behavior.

  5. Selection of melter systems for the DOE/Industrial Center for Waste Vitrification Research

    International Nuclear Information System (INIS)

    Bickford, D.F.

    1993-01-01

    The EPA has designated vitrification as the best developed available technology for immobilization of High-Level Nuclear Waste. In a recent federal facilities compliance agreement between the EPA, the State of Washington, and the DOE, the DOE agreed to vitrify all of the Low Level Radioactive Waste resulting from processing of High Level Radioactive Waste stored at the Hanford Site. This is expected to result in the requirement of 100 ton per day Low Level Radioactive Waste melters. Thus, there is increased need for the rapid adaptation of commercial melter equipment to DOE's needs. DOE has needed a facility where commercial pilot scale equipment could be operated on surrogate (non-radioactive) simulations of typical DOE waste streams. The DOE/Industry Center for Vitrification Research (Center) was established in 1992 at the Clemson University Department of Environmental Systems Engineering, Clemson, SC, to address that need. This report discusses some of the characteristics of the melter types selected for installation of the Center. An overall objective of the Center has been to provide the broadest possible treatment capability with the minimum number of melter units. Thus, units have been sought which have broad potential application, and which had construction characteristics which would allow their adaptation to various waste compositions, and various operating conditions, including extreme variations in throughput, and widely differing radiological control requirements. The report discusses waste types suitable for vitrification; technical requirements for the application of vitrification to low level mixed wastes; available melters and systems; and selection of melter systems. An annotated bibliography is included

  6. Selection of melter systems for the DOE/Industrial Center for Waste Vitrification Research

    Energy Technology Data Exchange (ETDEWEB)

    Bickford, D.F.

    1993-12-31

    The EPA has designated vitrification as the best developed available technology for immobilization of High-Level Nuclear Waste. In a recent federal facilities compliance agreement between the EPA, the State of Washington, and the DOE, the DOE agreed to vitrify all of the Low Level Radioactive Waste resulting from processing of High Level Radioactive Waste stored at the Hanford Site. This is expected to result in the requirement of 100 ton per day Low Level Radioactive Waste melters. Thus, there is increased need for the rapid adaptation of commercial melter equipment to DOE`s needs. DOE has needed a facility where commercial pilot scale equipment could be operated on surrogate (non-radioactive) simulations of typical DOE waste streams. The DOE/Industry Center for Vitrification Research (Center) was established in 1992 at the Clemson University Department of Environmental Systems Engineering, Clemson, SC, to address that need. This report discusses some of the characteristics of the melter types selected for installation of the Center. An overall objective of the Center has been to provide the broadest possible treatment capability with the minimum number of melter units. Thus, units have been sought which have broad potential application, and which had construction characteristics which would allow their adaptation to various waste compositions, and various operating conditions, including extreme variations in throughput, and widely differing radiological control requirements. The report discusses waste types suitable for vitrification; technical requirements for the application of vitrification to low level mixed wastes; available melters and systems; and selection of melter systems. An annotated bibliography is included.

  7. Formation rate of ammonium nitrate in the off-gas line of SRAT and SME in DWPF

    International Nuclear Information System (INIS)

    Lee, L.

    1992-01-01

    A mathematical model for the formation rate of ammonium nitrate in the off-gas line of the Sludge Receipt and Adjustment Tank (SRAT) and the Slurry Mixed Evaporator (SME) in DWPF has been developed. The formation rate of ammonium nitrate in the off-gas line depends on pH, temperature, volume and total concentration of ammonia and ammonium ion. Based on a typical SRAT and SME cycle in DWPF, this model predicts the SRAT contributes about 50 lbs of ammonium nitrate while SME contributes about 60 lbs of ammonium nitrate to the off-gas line

  8. Processing of high-temperature simulated waste glass in a continuous ceramic melter

    International Nuclear Information System (INIS)

    Barnes, S.M.; Brouns, R.A.; Hanson, M.S.

    1980-01-01

    Recent operations have demonstrated that high-melting-point glasses and glass-ceramics can be successfully processed in joule-heated, ceramic-lined melters with minor modifications to the existing technology. Over 500 kg of simulated waste glasses have been processed at temperatures up to 1410 0 C. The processability of the two high-temperature waste forms tested is similar to existing borosilicate waste glasses. High-temperature waste glass formulations produced in the bench-scale melter exhibit quality comparing favorably to standard waste glass formulations

  9. Advanced waste form and Melter development for treatment of troublesome high-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (also with high Al2O3 concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both

  10. Tunable molten oxide pool assisted plasma-melter vitrification systems

    Science.gov (United States)

    Titus, Charles H.; Cohn, Daniel R.; Surma, Jeffrey E.

    1998-01-01

    The present invention provides tunable waste conversion systems and apparatus which have the advantage of highly robust operation and which provide complete or substantially complete conversion of a wide range of waste streams into useful gas and a stable, nonleachable solid product at a single location with greatly reduced air pollution to meet air quality standards. The systems provide the capability for highly efficient conversion of waste into high quality combustible gas and for high efficiency conversion of the gas into electricity by utilizing a high efficiency gas turbine or an internal combustion engine. The solid product can be suitable for various commercial applications. Alternatively, the solid product stream, which is a safe, stable material, may be disposed of without special considerations as hazardous material. In the preferred embodiment, the arc plasma furnace and joule heated melter are formed as a fully integrated unit with a common melt pool having circuit arrangements for the simultaneous independently controllable operation of both the arc plasma and the joule heated portions of the unit without interference with one another. The preferred configuration of this embodiment of the invention utilizes two arc plasma electrodes with an elongated chamber for the molten pool such that the molten pool is capable of providing conducting paths between electrodes. The apparatus may additionally be employed with reduced use or without further use of the gases generated by the conversion process. The apparatus may be employed as a net energy or net electricity producing unit where use of an auxiliary fuel provides the required level of electricity production. Methods and apparatus for converting metals, non-glass forming waste streams and low-ash producing inorganics into a useful gas are also provided. The methods and apparatus for such conversion include the use of a molten oxide pool having predetermined electrical, thermal and physical

  11. Formulation and preparation of Hanford Waste Treatment Plant direct feed low activity waste Effluent Management Facility core simulant

    Energy Technology Data Exchange (ETDEWEB)

    McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, Charles A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL; Adamson, Duane J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL

    2016-05-01

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate, LMOGC) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream is to evaporate it in a new evaporator in the Effluent Management Facility (EMF) and then return it to the LAW melter. It is important to understand the composition of the effluents from the melter and new evaporator so that the disposition of these streams can be accurately planned and accommodated. Furthermore, alternate disposition of the LMOGC stream would eliminate recycling of problematic components, and would enable less integrated operation of the LAW melter and the Pretreatment Facilities. Alternate disposition would also eliminate this stream from recycling within WTP when it begins operations and would decrease the LAW vitrification mission duration and quantity of glass waste, amongst the other problems such a recycle stream present. This LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures and are problematic for the glass waste form, such as halides and sulfate. Because this stream will recycle within WTP, these components accumulate in the Melter Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Diverting the stream reduces the halides and sulfate in the recycled Condensate and is a key outcome of this work. This overall program examines the potential treatment and immobilization of this stream to enable alternative disposal. The objective of this task was to formulate and prepare a simulant of the LAW Melter

  12. NCAW feed chemistry: Effect of starting chemistry on melter offgas and iron redox

    International Nuclear Information System (INIS)

    Smith, P.A.; Vienna, J.D.; Merz, M.D.

    1995-03-01

    The Pacific Northwest Laboratory (PNL) Vitrification Technology Development (PVTD) program has been established to develop technology to support immobilization of selected Hanford wastes. The effort of the PVTD program is directed by the U.S. Department of Energy (DOE). This report is part of the effort and focuses on the effect of starting waste chemistry on the vitrification process. The objective of the investigation was the evaluation of the effect of starting chemistry on the cold cap behavior in the vitrification of simulated neutralized current acid waste (NCAW). In addition this investigation provides an initial laboratory investigation of the cold cap and method for evaluation of alternate reductants

  13. Preliminary evaluation of Am/Cm melter feed preparation process upset recovery flowsheets

    International Nuclear Information System (INIS)

    Stone, M.E.

    2000-01-01

    This document summarizes the results from the development of flowsheets to recover from credible processing errors specified in TTR 99-MNSS/SE-006. The proposed flowsheets were developed in laboratory scale equipment and will be utilized with minor modifications for full scale demonstrations in the Am/Cm Pilot Facility

  14. Effect of melter feed foaming on heat flux to the cold cap

    Czech Academy of Sciences Publication Activity Database

    Lee, S.; Hrma, P.; Pokorný, R.; Kloužek, Jaroslav; VanderVeer, B.J.; Dixon, D.R.; Luksic, S.A.; Rodriguez, C.P.; Chun, J.; Schweiger, M. J.; Kruger, A.A.

    2017-01-01

    Roč. 496, DEC 1 (2017), s. 54-65 ISSN 0022-3115 Institutional support: RVO:67985891 Keywords : cold cap * foam layer * heat flux * heat conductivity * evolved gas Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass OBOR OECD: Ceramics Impact factor: 2.048, year: 2016

  15. FY-97 operations of the pilot-scale glass melter to vitrify simulated ICPP high activity sodium-bearing waste

    International Nuclear Information System (INIS)

    Musick, C.A.

    1997-11-01

    A 3.5 liter refractory-lined joule-heated glass melter was built to test the applicability of electric melting to vitrify simulated high activity waste (HAW). The HAW streams result from dissolution and separation of Idaho Chemical Processing Plant (ICPP) calcines and/or radioactive liquid waste. Pilot scale melter operations will establish selection criteria needed to evaluate the application of joule heating to immobilize ICPP high activity waste streams. The melter was fabricated with K-3 refractory walls and Inconel 690 electrodes. It is designed to be continuously operated at 1,150 C with a maximum glass output rate of 10 lbs/hr. The first set of tests were completed using surrogate HAW-sodium bearing waste (SBW). The melter operated for 57 hours and was shut down due to excessive melt temperatures resulting in low glass viscosity (< 30 Poise). Due to the high melt temperature and low viscosity the molten glass breached the melt chamber. The melter has been dismantled and examined to identify required process improvement areas and successes of the first melter run. The melter has been redesigned and is currently being fabricated for the second run, which is scheduled to begin in December 1997

  16. Calibration and Measurement of the Viscosity of DWPF Start-Up Glass

    International Nuclear Information System (INIS)

    Schumacher, R.F.

    2001-01-01

    The Harrop, High-Temperature Viscometer has been in operation at the Savannah River Technology Center (SRTC) for several years and has proven itself to be reasonably accurate and repeatable. This is particularly notable when taking into consideration the small amount of glass required to make the viscosity determination. The volume of glass required is only 2.60 cc or about 6 to 7 grams of glass depending on the glass density. This may be compared to the more traditional viscosity determinations, which generally require between 100 to 1000 grams of glass. Before starting the present investigation, the unit was re-aligned and the furnace thermal gradients measured. The viscometer was again calibrated with available NIST Standard Reference Material glasses (717a and 710a) and a spindle constant equation was determined. Standard DWPF Waste Compliance Glasses (Purex, HM, and Batch 1) were used to provide additional verification for the determinations at low temperature. The Harrop, High-Temperature Viscometer was then used to determine the viscosity of three random samples of ground and blended DWPF, Black, Start -Up Frit, which were obtained from Pacific Northwest National Laboratory (PNNL). The glasses were in powder form and required melting prior to the viscosity determination. The results from this evaluation will be compared to ''Round Robin'' measurements from other DOE laboratories and a number of commercial laboratories

  17. Impact of Glycolate Anion on Aqueous Corrosion in DWPF and Downstream Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-12

    Glycolic acid is being evaluated as an alternate reductant in the preparation of high level waste for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). During processing, the glycolic acid may not be completely consumed with small quantities of the glycolate anion being carried forward to other high level waste (HLW) facilities. The SRS liquid waste contractor requested an assessment of the impact of the glycolate anion on the corrosion of the materials of construction (MoC) throughout the waste processing system since this impact had not been previously evaluated. A literature review revealed that corrosion data were not available for the MoCs in glycolic-bearing solutions applicable to SRS systems. Data on the material compatibility with only glycolic acid or its derivative products were identified; however, data were limited for solutions containing glycolic acid or the glycolate anion. For the proprietary coating systems applied to the DWPF concrete, glycolic acid was deemed compatible since the coatings were resistant to more aggressive chemistries than glycolic acid. Additionally similar coating resins showed acceptable resistance to glycolic acid.

  18. Impact of Glycolate Anion on Aqueous Corrosion in DWPF and Downstream Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-11-20

    Glycolic acid is being evaluated as an alternate reductant in the preparation of high level waste for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). During processing, the glycolic acid may not be completely consumed with small quantities of the glycolate anion being carried forward to other high level waste (HLW) facilities. The SRS liquid waste contractor requested an assessment of the impact of the glycolate anion on the corrosion of the materials of construction (MoC) throughout the waste processing system since this impact had not been previously evaluated. A literature review revealed that corrosion data were not available for the MoCs in glycolic-bearing solutions applicable to SRS systems. Data on the material compatibility with only glycolic acid or its derivative products were identified; however, data were limited for solutions containing glycolic acid or the glycolate anion. For the proprietary coating systems applied to the DWPF concrete, glycolic acid was deemed compatible since the coatings were resistant to more aggressive chemistries than glycolic acid. Additionally, similar coating resins showed acceptable resistance to glycolic acid.

  19. Test plan for evaluation of plasma melter technology for vitrification of high-sodium content low-level radioactive liquid wastes

    International Nuclear Information System (INIS)

    McLaughlin, D.F.; Lahoda, E.J.; Gass, W.R.; D'Amico, N.

    1994-01-01

    This document provides a test plan for the conduct of plasma arc vitrification testing by a vendor in support of the Hanford Tank Waste Remediation System (TWRS) Low-Level Waste (LLW) Vitrification Program. The vendor providing this test plan and conducting the work detailed within it [one of seven selected for glass melter testing under Purchase Order MMI-SVV-384212] is the Westinghouse Science and Technology Center (WSTC) in Pittsburgh, PA. WSTC authors of the test plan are D. F. McLaughlin, E. J. Lahoda, W. R. Gass, and N. D'Amico. The WSTC Program Manager for this test is D. F. McLaughlin. This test plan is for Phase I activities described in the above Purchase Order. Test conduct includes melting of glass frit with Hanford LLW Double-Shell Slurry Feed waste simulant in a plasma arc fired furnace

  20. The Production of Advanced Glass Ceramic HLW Forms using Cold Crucible Induction Melter

    Energy Technology Data Exchange (ETDEWEB)

    Veronica J Rutledge; Vince Maio

    2013-10-01

    Cold Crucible Induction Melters (CCIMs) will favorably change how High-Level radioactive Waste (from nuclear fuel recovery) is treated in the 21st century. Unlike the existing Joule-Heated Melters (JHMs) currently in operation for the glass-based immobilization of High-Level Waste (HLW), CCIMs offer unique material features that will increase melt temperatures, increase throughput, increase mixing, increase loading in the waste form, lower melter foot prints, eliminate melter corrosion and lower costs. These features not only enhance the technology for producing HLW forms, but also provide advantageous attributes to the waste form by allowing more durable alternatives to glass. This paper discusses advantageous features of the CCIM, with emphasis on features that overcome the historical issues with the JHMs presently utilized, as well as the benefits of glass ceramic waste forms over borosilicate glass waste forms. These advantages are then validated based on recent INL testing to demonstrate a first-of-a-kind formulation of a non-radioactive ceramic-based waste form utilizing a CCIM.

  1. The production of advanced glass ceramic HLW forms using cold crucible induction melter

    International Nuclear Information System (INIS)

    Rutledge, V.J.; Maio, V.

    2013-01-01

    Cold Crucible Induction Melters (CCIM) will favorably change how High-Level radioactive Waste (from nuclear fuel recovery) is treated in a near future. Unlike the existing Joule-Heated Melters (JHM) currently in operation for the glass-based immobilization of High-Level Waste (HLW), CCIM offers unique material features that will increase melt temperatures, increase throughput, increase mixing, increase loading in the waste form, lower melter foot prints, eliminate melter corrosion and lower costs. These features not only enhance the technology for producing HLW forms, but also provide advantageous attributes to the waste form by allowing more durable alternatives to glass. It is concluded that glass ceramic waste forms that are tailored to immobilize fission products of HLW can be can be made from the HLW processed with the CCIM. The advantageous higher temperatures reached with the CCIM and unachievable with JHM allows the lanthanides, alkali, alkaline earths, and molybdenum to dissolve into a molten glass. Upon controlled cooling they go into targeted crystalline phases to form a glass ceramic waste form with higher waste loadings than achievable with borosilicate glass waste forms. Natural cooling proves to be too fast for the formation of all targeted crystalline phases

  2. Crystal-Tolerant Glass Approach For Mitigation Of Crystal Accumulation In Continuous Melters Processing Radioactive Waste

    International Nuclear Information System (INIS)

    Kruger, Albert A.; Rodriguez, Carmen P.; Lang, Jesse B.; Huckleberry, Adam R.; Matyas, Josef; Owen, Antoinette T.

    2012-01-01

    High-level radioactive waste melters are projected to operate in an inefficient manner as they are subjected to artificial constraints, such as minimum liquidus temperature (T L ) or maximum equilibrium fraction of crystallinity at a given temperature. These constraints substantially limit waste loading, but were imposed to prevent clogging of the melter with spinel crystals [(Fe, Ni, Mn, Zn)(Fe, Cr) 2 O 4 ]. In the melter, the glass discharge riser is the most likely location for crystal accumulation during idling because of low glass temperatures, stagnant melts, and small diameter. To address this problem, a series of lab-scale crucible tests were performed with specially formulated glasses to simulate accumulation of spinel in the riser. Thicknesses of accumulated layers were incorporated into empirical model of spinel settling. In addition, T L of glasses was measured and impact of particle agglomeration on accumulation rate was evaluated. Empirical model predicted well the accumulation of single crystals and/or smallscale agglomerates, but, excessive agglomeration observed in high-Ni-Fe glass resulted in an under-prediction of accumulated layers, which gradually worsen over time as an increased number of agglomerates formed. Accumulation rate of ∼14.9 +- 1 nm/s determined for this glass will result in ∼26 mm thick layer in 20 days of melter idling

  3. A Joule-Heated Melter Technology For The Treatment And Immobilization Of Low-Activity Waste

    International Nuclear Information System (INIS)

    Kelly, S.E.

    2011-01-01

    This report is one of four reports written to provide background information regarding immobilization technologies remaining under consideration for supplemental immobilization of Hanford's low-activity waste. This paper provides the reader a general understanding of joule-heated ceramic lined melters and their application to Hanford's low-activity waste.

  4. Glass science tutorial: Lecture number-sign 2, Operating electric glass melters. James N. Edmonson, Lecturer

    International Nuclear Information System (INIS)

    Kruger, A.A.

    1994-10-01

    This report contains basic information on electric furnaces used for glass melting and on the properties of glass useful for the stabilization of radioactive wastes. Furnace nomenclature, furnace types, typical silicate glass composition and properties, thermal conductivity information, kinetics of the melting process, glass furnace refractory materials composition and thermal conductivity, and equations required for the operation of glass melters are included

  5. A JOULE-HEATED MELTER TECHNOLOGY FOR THE TREATMENT AND IMMOBILIZATION OF LOW-ACTIVITY WASTE

    Energy Technology Data Exchange (ETDEWEB)

    KELLY SE

    2011-04-07

    This report is one of four reports written to provide background information regarding immobilization technologies remaining under consideration for supplemental immobilization of Hanford's low-activity waste. This paper provides the reader a general understanding of joule-heated ceramic lined melters and their application to Hanford's low-activity waste.

  6. Initial Laboratory-Scale Melter Test Results for Combined Fission Product Waste

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Crum, Jarrod V.; Buchmiller, William C.; Rieck, Bennett T.; Schweiger, Michael J.; Vienna, John D.

    2009-10-01

    This report describes the methods and results used to vitrify a baseline glass, CSLNTM-C-2.5 in support of the AFCI (Advanced Fuel Cycle Initiative) using a Quartz Crucible Scale Melter at the Pacific Northwest National Laboratory. Document number AFCI-WAST-PMO-MI-DV-2009-000184.

  7. Demonstration of an approach to waste form qualification through simulation of liquid-fed ceramic melter process operations

    International Nuclear Information System (INIS)

    Reimus, P.W.; Kuhn, W.L.; Peters, R.D.; Pulsipher, B.A.

    1986-07-01

    During fiscal year 1982, the US Department of Energy (DOE) assigned responsibility for managing civilian nuclear waste treatment programs in the United States to the Nuclear Waste Treatment Program (NWTP) at the Pacific Northwest Laboratory (PNL). One of the principal objectives of this program is to establish relationships between vitrification process control and glass quality. Users of the liquid-fed ceramic melter (LFCM) process will need such relationships in order to establish acceptance of vitrified high-level nuclear waste at a licensed federal repository without resorting to destructive examination of the canisters. The objective is to be able to supply a regulatory agency with an estimate of the composition, durability, and integrity of the glass in each waste glass canister produced from an LFCM process simply by examining the process data collected during the operation of the LFCM. The work described here will continue through FY-1987 and culminate in a final report on the ability to control and monitor an LFCM process through sampling and process control charting of the LFCM feed system

  8. SLUDGE WASHING AND DEMONSTRATION OF THE DWPF FLOWSHEET IN THE SRNL SHIELDED CELLS FOR SLUDGE BATCH 6 QUALIFICATION

    Energy Technology Data Exchange (ETDEWEB)

    Pareizs, J.; Pickenheim, B.; Bannochie, C.; Billings, A.; Bibler, N.; Click, D.

    2010-10-01

    Prior to initiating a new sludge batch in the Defense Waste Processing Facility (DWPF), Savannah River National Laboratory (SRNL) is required to simulate this processing, including Chemical Process Cell (CPC) simulation, waste glass fabrication, and chemical durability testing. This report documents this simulation for the next sludge batch, Sludge Batch 6 (SB6). SB6 consists of Tank 12 material that has been transferred to Tank 51 and subjected to Low Temperature Aluminum Dissolution (LTAD), Tank 4 sludge, and H-Canyon Pu solutions. Following LTAD and the Tank 4 addition, Liquid Waste Operations (LWO) provided SRNL a 3 L sample of Tank 51 sludge for SB6 qualification. Pu solution from H Canyon was also received. SB6 qualification included washing the sample per LWO plans/projections (including the addition of Pu from H Canyon), DWPF CPC simulations, waste glass fabrication (vitrification), and waste glass characterization and chemical durability evaluation. The following are significant observations from this demonstration. Sludge settling improved slightly as the sludge was washed. SRNL recommended (and the Tank Farm implemented) one less wash based on evaluations of Tank 40 heel projections and projections of the glass composition following transfer of Tank 51 to Tank 40. Thorium was detected in significant quantities (>0.1 wt % of total solids) in the sludge. In past sludge batches, thorium has been determined by Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS), seen in small quantities, and reported with the radionuclides. As a result of the high thorium, SRNL-AD has added thorium to their suite of Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES) elements. The acid stoichiometry for the DWPF Sludge Receipt and Adjustment Tank (SRAT) processing of 115%, or 1.3 mol acid per liter of SRAT receipt slurry, was adequate to accomplish some of the goals of SRAT processing: nitrite was destroyed to below 1,000 mg/kg and mercury was removed to

  9. High-temperature vitrification of Hanford residual-liquid waste in a continuous melter

    International Nuclear Information System (INIS)

    Barnes, S.M.

    1980-04-01

    Over 270 kg of high-temperature borosilicate glass have been produced in a series of three short-term tests in the High-Temperature Ceramic Melter vitrification system at PNL. The glass produced was formulated to vitrify simulated Hanford residual-liquid waste. The tests were designed to (1) demonstrate the feasibility of utilizing high-temperature, continuous-vitrification technology for the immobilization of the residual-liquid waste, (2) test the airlift draining technique utilized by the high-temperature melter, (3) compare glass produced in this process to residual-liquid glass produced under laboratory conditions, (4) investigate cesium volatility from the melter during waste processing, and (5) determine the maximum residual-liquid glass production rate in the high-temperature melter. The three tests with the residual-liquid composition confirmed the viability of the continuous-melting vitrification technique for the immobilization of this waste. The airlift draining technique was demonstrated in these tests and the glass produced from the melter was shown to be less porous than the laboratory-produced glass. The final glass produced from the second test was compared to a glass of the same composition produced under laboratory conditions. The comparative tests found the glasses to be indistinguishable, as the small differences in the test results fell within the precision range of the characterization testing equipment. The cesium volatility was examined in the final test. This examination showed that 0.44 wt % of the cesium (assumed to be cesium oxide) was volatilized, which translates to a volatilization rate of 115 mg/cm 2 -h

  10. Determination of halogen content in glass for assessment of melter decontamination factors

    International Nuclear Information System (INIS)

    Goles, R.W.

    1996-03-01

    Melter decontamination factor (DF) values for the halogens (fluorine, chlorine, and iodine) are important to the Hanford Waste Vitrification Plant (HWVP) process because of the potential influence of DF on secondary-waste recycle strategies (fluorine and chlorine) as well as its impact on off-gas emissions (iodine). This study directly establishes the concentrations of halides-in HWVP simulated reference glasses rather than relying on indirect off-gas data. For fluorine and chlorine, pyrohydrolysis coupled with halide (ion chromatographic) detection has proven to be a useful analytical approach suitable for glass matrices, sensitive enough for the range of halogens encountered, and compatible with remote process support applications. Results obtained from pyrohydrolytic analysis of pilot-scale ceramic melter (PSCM) -22 and -23 glasses indicate that the processing behavior of fluorine and chlorine is quite variable even under similar processing conditions. Specifically, PSCM-23 glass exhibited a ∼90% halogen (F and Cl) retention efficiency, while only 20% was incorporated in PSCM-22 glass. These two sets of very dissimilar test results clearly do not form a sufficient basis for establishing design DF values for fluorine and chlorine. Because the present data do not provide any new halogen volatility information, but instead reconfirm the validity of previously obtained offgas derived values, melter DF values of 4, 2, and 1 for fluorine, chlorine, and iodine, respectively, are recommended for adoption; these values were conservatively established by a team of responsible engineers at Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratory (PNL) on the basis of average behavior for many comparable melter tests. In the absence of further HWVP process data, these average melter DFs are the best values currently available

  11. Immobilization of simulated high-level radioactive waste in borosilicate glass: Pilot scale demonstrations

    International Nuclear Information System (INIS)

    Ritter, J.A.; Hutson, N.D.; Zamecnik, J.R.; Carter, J.T.

    1991-01-01

    The Integrated DWPF Melter System (IDMS), operated by the Savannah River Laboratory, is a pilot scale facility used in support of the start-up and operation of the Department of Energy's Defense Waste Processing Facility. The IDMS has successfully demonstrated, on an engineering scale (one-fifth), that simulated high level radioactive waste (HLW) sludge can be chemically treated with formic acid to adjust both its chemical and physical properties, and then blended with simulated precipitate hydrolysis aqueous (PHA) product and borosilicate glass frit to produce a melter feed which can be processed into a durable glass product. The simulated sludge, PHA and frit were blended, based on a product composition program, to optimize the loading of the waste glass as well as to minimize those components which can cause melter processing and/or glass durability problems. During all the IDMS demonstrations completed thus far, the melter feed and the resulting glass that has been produced met all the required specifications, which is very encouraging to future DWPF operations. The IDMS operations also demonstrated that the volatile components of the melter feed (e.g., mercury, nitrogen and carbon, and, to a lesser extent, chlorine, fluorine and sulfur) did not adversely affect the melter performance or the glass product

  12. The dismantling of the one-third-scale Joule ceramic melter and preliminary investigation of electrode corrosion

    International Nuclear Information System (INIS)

    Morris, J.B.; Walmsley, D.; Hollinrake, A.; Horsley, G.

    1986-01-01

    The Harwell one-third scale Joule ceramic melter was dismantled to discover the cause of a fall in electric resistance. The two inconel-690 electrodes were corroded over the lower 40mm sections and were examined by optical and electron microscopy. Sedimentation of Ru species on the floor of the melter may have led to corrosion of the electrodes. Glass withdrawn from the canisters was analyzed for evidence of a segregation mechanism. (UK)

  13. FINAL REPORT REGULATORY OFF GAS EMISSIONS TESTING ON THE DM1200 MELTER SYSTEM USING HLW AND LAW SIMULANTS VSL-05R5830-1 REV 0 10/31/05

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    The operational requirements for the River Protection Project - Waste Treatment Plant (RPP-WTP) Low Activity Waste (LAW) and High Level Waste (HLW) melter systems, together with the feed constituents, impose a number of challenges to the off-gas treatment system. The system must be robust from the standpoints of operational reliability and minimization of maintenance. The system must effectively control and remove a wide range of solid particulate matter, acid mists and gases, and organic constituents (including those arising from products of incomplete combustion of sugar and organics in the feed) to concentration levels below those imposed by regulatory requirements. The baseline design for the RPP-WTP LAW primary off-gas system includes a submerged bed scrubber (SBS), a wet electrostatic precipitator (WESP), and a high efficiency particulate air (HEPA) filter. The secondary off-gas system includes a sulfur-impregnated activated carbon bed (AC-S), a thermal catalytic oxidizer (TCO), a single-stage selective catalytic reduction NOx treatment system (SCR), and a packed-bed caustic scrubber (PBS). The baseline design for the RPP-WTP HLW primary off-gas system includes an SBS, a WESP, a high efficiency mist eliminator (HEME), and a HEPA filter. The HLW secondary off-gas system includes a sulfur-impregnated activated carbon bed, a silver mordenite bed, a TCO, and a single-stage SCR. The one-third scale HLW DM1200 Pilot Melter installed at the Vitreous State Laboratory (VSL) was equipped with a prototypical off-gas train to meet the needs for testing and confirmation of the performance of the baseline off-gas system design. Various modifications have been made to the DM1200 system as the details of the WTP design have evolved, including the installation of a silver mordenite column and an AC-S column for testing on a slipstream of the off-gas flow; the installation of a full-flow AC-S bed for the present tests was completed prior to initiation of testing. The DM1200

  14. Vitrification of HLW produced by uranium/molybdenum fuel reprocessing in cogema's cold crucible melter

    International Nuclear Information System (INIS)

    Quang, R. Do; Petitjean, V.; Hollebeque, F.; Pinet, O.; Flament, T.; Prodhomme, A.; Dalcorso, J. P.

    2003-01-01

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R and D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12% in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed

  15. Vitrification of HLW Produced by Uranium/Molybdenum Fuel Reprocessing in COGEMA's Cold Crucible Melter

    International Nuclear Information System (INIS)

    Do Quang, R.; Petitjean, V.; Hollebecque, F.; Pinet, O.; Flament, T.; Prod'homme, A.

    2003-01-01

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R and D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed

  16. Formulation of special glass frit and its use for decontamination of Joule melter employed for vitrification of high level and radioactive liquid waste

    International Nuclear Information System (INIS)

    Valsala, T.P.; Mishra, P.K.; Thakur, D.A.; Ghongane, D.E.; Jayan, R.V.; Dani, U.; Sonavane, M.S.; Kulkarni, Y.

    2012-01-01

    Advanced vitrification system at TWMP Tarapur was used for successful vitrification of large volume of HLW stored in waste tank farm. After completion of the operational life of the joule melter, dismantling was planned. Prior to the dismantling, the hold up inventory of active glass product from the melter was flushed out using specially formulated inactive glass frit to reduce the air activity buildup in the cell during dismantling operations. The properties of the special glass frit prepared are comparable with that of the regular product glass. More than 94% of holdup activity was flushed out from the joule melter prior to the dismantling of the melter. (author)

  17. Production and remediation of low sludge simulated Purex waste glasses, 2: Effects of sludge oxide additions on glass durability

    International Nuclear Information System (INIS)

    Ramsey, W.G.

    1993-01-01

    Glass produced during the Purex 4 campaigns of the Integrated DWPF Melter System (IDMS) and the 774 Research Melter contained a lower fraction of sludge components than targeted by the Product Composition Control System (PCCS). Purex 4 glass was more durable than the benchmark (EA) glass, but was less durable than most other simulated SRS high-level waste glasses. Further, the measured durability of Purex 4 glass was not as well correlated with the durability predicted from the DWPF process control algorithm, probably because the algorithm was developed to predict the durability of SRS high-level waste glasses with higher sludge content than Purex 4. A melter run, designated Purex 4 Remediation, was performed using the 774 Research Melter to determine if the initial PCCS target composition determined for Purex 4 would produce acceptable glass whose durability could be accurately modeled by the DWPF glass durability algorithm. Reagent grade oxides and carbonates were added to Purex 4 melter feed stock to simulate a higher sludge loading. Each canister of glass produced was sampled and the glass durability was determined by the Product Consistency Test method. This document details the durability data and subsequent analysis

  18. Redox Control For Hanford HLW Feeds VSL-12R2530-1, REV 0

    International Nuclear Information System (INIS)

    Kruger, A. A.; Matlack, Keith S.; Pegg, Ian L.; Kot, Wing K.; Joseph, Innocent

    2012-01-01

    The principal objectives of this work were to investigate the effects of processing simulated Hanford HLW at the estimated maximum concentrations of nitrates and oxalates and to identify strategies to mitigate any processing issues resulting from high concentrations of nitrates and oxalates. This report provides results for a series of tests that were performed on the DM10 melter system with simulated C-106/AY-102 HLW. The tests employed simulated HLW feeds containing variable amounts of nitrates and waste organic compounds corresponding to maximum concentrations proj ected for Hanford HLW streams in order to determine their effects on glass production rate, processing characteristics, glass redox conditions, melt pool foaming, and the tendency to form secondary phases. Such melter tests provide information on key process factors such as feed processing behavior, dynamic effects during processing, processing rates, off-gas amounts and compositions, foaming control, etc., that cannot be reliably obtained from crucible melts

  19. Noble metal behavior during melting of simulated high-level nuclear waste glass feeds

    International Nuclear Information System (INIS)

    Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

    1993-04-01

    Noble metals and their oxides can settle in waste glass melters and cause electrical shorting. Simulated waste feeds from Hanford, Savannah River, and Germany were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600 degrees C--1000 degrees C and examined by electron microscopy to determine shapes, sizes, and distribution of noble metal particles as a function of temperature. Individual noble metal particles and agglomerates of rhodium (Rh), ruthenium (RuO 2 ), and palladium (Pd), as well as their alloys, were seen. the majority of particles and agglomerates were generally less than 10 microns; however, large agglomerations (up to 1 mm) were found in the German feed. Detailed particle distribution and characterization was performed for a Hanford waste to provide input to computer modeling of particle settling in the melter

  20. Redox Control For Hanford HLW Feeds VSL-12R2530-1, REV 0

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Kot, Wing K. [The Catholic University of America, Washington, DC (United States); Joseph, Innocent [The Catholic University of America, Washington, DC (United States)

    2012-12-13

    The principal objectives of this work were to investigate the effects of processing simulated Hanford HLW at the estimated maximum concentrations of nitrates and oxalates and to identify strategies to mitigate any processing issues resulting from high concentrations of nitrates and oxalates. This report provides results for a series of tests that were performed on the DM10 melter system with simulated C-106/AY-102 HLW. The tests employed simulated HLW feeds containing variable amounts of nitrates and waste organic compounds corresponding to maximum concentrations proj ected for Hanford HLW streams in order to determine their effects on glass production rate, processing characteristics, glass redox conditions, melt pool foaming, and the tendency to form secondary phases. Such melter tests provide information on key process factors such as feed processing behavior, dynamic effects during processing, processing rates, off-gas amounts and compositions, foaming control, etc., that cannot be reliably obtained from crucible melts.

  1. Report - Melter Testing of New High Bismuth HLW Formulations VSL-13R2770-1

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Pegg, I. L.; Kot, W. K.; Gan, H.; Matlack, K. S.

    2013-11-13

    The primary objective of the work described was to test two glasses formulated for a high bismuth waste stream on the DM100 melter system. Testing was designed to determine processing characteristics and production rates, assess the tendency for foaming, and confirm glass properties. The glass compositions tested were previously developed to maintain high waste loadings and processing rates while suppressing the foaming observed in previous tests

  2. SETTLING OF SPINEL IN A HIGH-LEVEL WASTE GLASS MELTER

    International Nuclear Information System (INIS)

    Pavel Hrma; Pert Schill; Lubomir Nemec

    2002-01-01

    High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors called melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150 decrees C to create a melt that becomes glass on cooling

  3. Testing of the melter lid refractory for the West Valley Demonstration Project (WVDP)

    International Nuclear Information System (INIS)

    Gupta, A.; Jain, V.; Mahoney, J.L.; Holman, T.M.

    1991-01-01

    Monofrax H and Mulfrax 202 refractory were tested for potential application as the melter lid refractory for the WVDP. Resistance to spalling and corrosion by the slurry and offgas salts were primary criteria for selection. Test specimens were subjected to thermal cycling between 450 and 1,100C for five weeks. Visual examination indicated some corrosion but no spalling. SEM/EDS analysis was performed to determine the glass/refractory interface corrosion mechanism. The refractory selection basis will be discussed

  4. The integrated melter off-gas treatment systems at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Vance, R.F.

    1991-12-01

    The West Valley Demonstration project was established by an act of Congress in 1980 to solidify the high level radioactive liquid wastes produced from operation of the Western New York Nuclear Services Center from 1966 to 1972. The waste will be solidified as borosilicate glass. This report describes the functions, the controlling design criteria, and the resulting design of the melter off-gas treatment systems

  5. Phase II of a Six sigma Initiative to Study DWPF SME Analytical Turnaround Times: SRNL's Evaluation of Carbonate-Based Dissolution Methods

    International Nuclear Information System (INIS)

    Edwards, Thomas

    2005-01-01

    The Analytical Development Section (ADS) and the Statistical Consulting Section (SCS) of the Savannah River National Laboratory (SRNL) are participating in a Six Sigma initiative to improve the Defense Waste Processing Facility (DWPF) Laboratory. The Six Sigma initiative has focused on reducing the analytical turnaround time of samples from the Slurry Mix Evaporator (SME) by developing streamlined sampling and analytical methods [1]. The objective of Phase I was to evaluate the sub-sampling of a larger sample bottle and the performance of a cesium carbonate (Cs 2 CO 3 ) digestion method. Successful implementation of the Cs 2 CO 3 fusion method in the DWPF would have important time savings and convenience benefits because this single digestion would replace the dual digestion scheme now used. A single digestion scheme would result in more efficient operations in both the DWPF shielded cells and the inductively coupled plasma--atomic emission spectroscopy (ICP-AES) laboratory. By taking a small aliquot of SME slurry from a large sample bottle and dissolving the vitrified SME sample with carbonate fusion methods, an analytical turnaround time reduction from 27 hours to 9 hours could be realized in the DWPF. This analytical scheme has the potential for not only dramatically reducing turnaround times, but also streamlining operations to minimize wear and tear on critical shielded cell components that are prone to fail, including the Hydragard(trademark) sampling valves and manipulators. Favorable results from the Phase I tests [2] led to the recommendation for a Phase II effort as outlined in the DWPF Technical Task Request (TTR) [3]. There were three major tasks outlined in the TTR, and SRNL issued a Task Technical and QA Plan [4] with a corresponding set of three major task activities: (1) Compare weight percent (wt%) total solids measurements of large volume samples versus peanut vial samples. (2) Evaluate Cs 2 CO 3 and K 2 CO 3 fusion methods using DWPF simulated

  6. Predictive modeling of crystal accumulation in high-level waste glass melters processing radioactive waste

    Science.gov (United States)

    Matyáš, Josef; Gervasio, Vivianaluxa; Sannoh, Sulaiman E.; Kruger, Albert A.

    2017-11-01

    The effectiveness of high-level waste vitrification at Hanford's Waste Treatment and Immobilization Plant may be limited by precipitation/accumulation of spinel crystals [(Fe, Ni, Mn, Zn)(Fe, Cr)2O4] in the glass discharge riser of Joule-heated ceramic melters during idling. These crystals do not affect glass durability; however, if accumulated in thick layers, they can clog the melter and prevent discharge of molten glass into canisters. To address this problem, an empirical model was developed that can predict thicknesses of accumulated layers as a function of glass composition. This model predicts well the accumulation of single crystals and/or small-scale agglomerates, but excessive agglomeration observed in high-Ni-Fe glass resulted in an underprediction of accumulated layers, which gradually worsened over time as an increased number of agglomerates formed. The accumulation rate of ∼53.8 ± 3.7 μm/h determined for this glass will result in a ∼26 mm-thick layer after 20 days of melter idling.

  7. Pilot-scale ceramic melter 1985-1986 rebuild: Nuclear Waste Treatment Program

    International Nuclear Information System (INIS)

    Koegler, S.S.

    1987-07-01

    The pilot-scale ceramic melter (PSCM) was subsequently dismantled, and the damaged and corroded components were repaired or replaced. The PSCM rebuild ensures that the melter will be available for an additional three to five years of planned testing. An analysis of the corrosion products and the failed electrodes indicated that the electrode bus connection welds may have failed due to a combination of chemical and mechanical effects. The electrodes were replaced with a design similar to the original electrodes, but with improved electrical bus connections. The implications of the PSCM electrode corrosion evaluation are that, although Inconel 690 has excellent corrosion resistance to molten glass, corrosion at the melt line in stagnant regions is a significant concern. Functional changes made during the rebuild included increases in wall and floor insulation to better simulate well-insulated melters, a decrease in the lid height for more prototypical plenum and off-gas conditions, and installation of an Inconel 690 trough and dam to improve glass pouring and prevent glass seepage. 9 refs., 33 figs., 5 tabs

  8. Thermal stress analysis of an Am/Cm stabilization bushing melter

    International Nuclear Information System (INIS)

    Gong, C.; Hardy, B.J.

    1996-01-01

    Decades of nuclear material production at the Savannah River Site (SRS) has resulted in the generation of large quantities of the isotopes Am 243 and Cm 244 . Currently, the Am and Cm isotopes are stored as a nitric acid solution in a tank. The Am and Cm isotopes have great commercial value but must be transferred to the Oak Ridge National Laboratory (ORNL) for processing. The nitric acid solution contains other isotopes and is intensely radioactive, which makes storage a problem and precludes shipment in the liquid form. In order to stabilize the material for onsite storage and to permit transport the material from SRS to ORNL, it has been proposed that the Am and Cm be separated from other isotopes in the solution and vitrified. The vitrification process in the Platinum-Rhodium alloy vessel generates a wide spectrum of temperature distributions. The melter is partially supported by a suspension system and confined by the flexible insulation. The combination of the fluctuation of temperature distribution and variable boundary conditions, induces stresses and strains in the melter. The thermal stress analysis is carried out with the finite element code ABAQUS. This analysis is closely associated with the design, manufacture and testing of the melter. The results were compared with the test data

  9. SLUDGE WASHING AND DEMONSTRATION OF THE DWPF FLOWSHEET IN THE SRNL SHIELDED CELLS FOR SLUDGE BATCH 5 QUALIFICATION

    Energy Technology Data Exchange (ETDEWEB)

    Pareizs, J; Cj Bannochie, C; Damon Click, D; Dan Lambert, D; Michael Stone, M; Bradley Pickenheim, B; Amanda Billings, A; Ned Bibler, N

    2008-11-10

    Sludge Batch 5 (SB5) is predominantly a combination of H-modified (HM) sludge from Tank 11 that underwent aluminum dissolution in late 2007 to reduce the total mass of sludge solids and aluminum being fed to the Defense Waste Processing Facility (DWPF) and Purex sludge transferred from Tank 7. Following aluminum dissolution, the addition of Tank 7 sludge and excess Pu to Tank 51, Liquid Waste Operations (LWO) provided the Savannah River National Laboratory (SRNL) a 3-L sample of Tank 51 sludge for SB5 qualification. SB5 qualification included washing the sample per LWO plans/projections (including the addition of a Pu/Be stream from H Canyon), DWPF Chemical Process Cell (CPC) simulations, waste glass fabrication (vitrification), and waste glass chemical durability evaluation. This report documents: (1) The washing (addition of water to dilute the sludge supernatant) and concentration (decanting of supernatant) of the Tank 51 qualification sample to adjust sodium content and weight percent insoluble solids to Tank Farm projections. (2) The performance of a DWPF CPC simulation using the washed Tank 51 sample. This includes a Sludge Receipt and Adjustment Tank (SRAT) cycle, where acid is added to the sludge to destroy nitrite and remove mercury, and a Slurry Mix Evaporator (SME) cycle, where glass frit is added to the sludge in preparation for vitrification. The SME cycle also included replication of five canister decontamination additions and concentrations. Processing parameters for the CPC processing were based on work with a non radioactive simulant. (3) Vitrification of a portion of the SME product and Product Consistency Test (PCT) evaluation of the resulting glass. (4) Rheology measurements of the initial slurry samples and samples after each phase of CPC processing. This work is controlled by a Task Technical and Quality Assurance Plan (TTQAP) , and analyses are guided by an Analytical Study Plan. This work is Technical Baseline Research and Development (R

  10. DETERMINATION OF REPORTABLE RADIONUCLIDES FOR DWPF SLUDGE BATCH 4 MACROBATCH 5

    International Nuclear Information System (INIS)

    Bannochie, C; Ned Bibler, N; David Diprete, D

    2008-01-01

    The Waste Acceptance Product Specifications (WAPS)1 1.2 require that 'The Producer shall report the inventory of radionuclides (in Curies) that have half-lives longer than 10 years and that are, or will be, present in concentrations greater than 0.05 percent of the total inventory for each waste type indexed to the years 2015 and 3115'. As part of the strategy to meet WAPS 1.2, the Defense Waste Processing Facility (DWPF) will report for each waste type, all radionuclides (with half-lives greater than 10 years) that have concentrations greater than 0.01 percent of the total inventory from time of production through the 1100 year period from 2015 through 3115. The initial listing of radionuclides to be included is based on the design-basis glass as identified in the Waste Form Compliance Plan (WCP)2 and Waste Form Qualification Report (WQR)3. However, it is required that this list be expanded if other radionuclides with half-lives greater than 10 years are identified that may meet the greater than 0.01% criterion for Curie content. Specification 1.6 of the WAPS, International Atomic Energy Agency (IAEA) Safeguards Reporting for High Level Waste (HLW), requires that the ratio by weights of the following uranium and plutonium isotopes be reported: U-233, U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, and Pu-242. Therefore, the complete set of reportable radionuclides must also include this set of U and Pu isotopes. The DWPF is receiving radioactive sludge slurry from HLW Tank 40. The radioactive sludge slurry in Tank 40 is a blend of the previous contents of Tank 40 (Sludge Batch 3) and the sludge that was transferred to Tank 40 from Tank 51. The blend of sludge from Tank 51 and Tank 40 defines Sludge Batch 4 (also referred to as Macrobatch 5 (MB5)). This report develops the list of reportable radionuclides and associated activities and determines the radionuclide activities as a function of time. The DWPF will use this list and the activities as one of

  11. Iron Phosphate Glass for Vitrifying Hanford AZ102 LAW in Joule Heated and Cold Crucible Induction Melters - 12240

    Energy Technology Data Exchange (ETDEWEB)

    Day, Delbert E.; Brow, Richard K.; Ray, Chandra S.; Reis, Signo T. [Missouri University of Science and Technology, 1870 Miner Circle, Rolla, MO 65409 (United States); Kim, Cheol-Woon [MO-SCI Corporation, 4040 HyPoint North, Rolla, MO 65401 (United States); Vienna, John D.; Sevigny, Gary [Pacific North West National Laboratory, Battelle Blvd., Richland, WA 99352 (United States); Peeler, David; Johnson, Fabienne C.; Hansen, Eric K. [Savannah River National Laboratory, Savannah River Site, 999-W, Aiken, SC 29803 (United States); Soelberg, Nick [Idaho National Laboratory, 2525 Fremont Avenue, Idaho Falls, ID 83415 (United States); Pegg, Ian L.; Gan, Hao [Catholic University of America, 620 Michigan Avenue, N.E., Washington, DC 20064 (United States)

    2012-07-01

    composition produces a homogeneous glass with a density of 2.80 ± 0.04 g/cm{sup 3} after melting between 1000 and 1050 deg. C for 3 to 5 h. (2) This is the first time that an iron phosphate glass was melted in the JHM continuously for 10 days achieving a specific melting rate of 1010 kg/m2/day and in the CCIM for 70 hrs with a melting rate of 664 kg/m{sup 2}/day. (3) The analyzed (ICP-AES) compositions for all the glasses prepared under several melting conditions are in excellent agreement with the target composition. The variation of melting conditions include: (i) use of small scale/short melting time to large scale/long melting time (300 g to 80 kg, 4 h to 10 days) operations including melting in the JHM and CCIM, (ii) use of dry or wet (slurry) melter feed, (iii) addition of reductant (sugar) in the batch, and (iv) bubbling the melt with air. (4) The chemical durability as measured by PCT and VHT for the quenched and CCC-treated waste forms prepared from laboratory, JHM or CCIM melting exceeds the DOE requirements for LAW. (5) Depending upon the melting time (4 h to 10 days), the average concentration of SO{sub 3} in the MS26AZ102F-2 iron phosphate glass waste form varies from about 1.78 (RSM) to 3.74 (laboratory melt) wt% which corresponds to an SO{sub 3} retention of 41 to 86% when melted between 1030 at 1050 deg. C in air. (6) The retention of SO{sub 3} in the glass was reduced when a reductant (sugar) was added to the slurry. (7) The retention of other problem components like Cesium and Rhenium/Tc-99 is also high; from 92 to 100% for Cs{sub 2}O under different melting conditions, and from 66 to 33% for Re{sub 2}O{sub 7} for melts processed from 3 to 5 hours and up to 10 days. (8) Corrosion tests on Inconel 693 and K-3 refractory at temperatures between 1000 and 1050 deg. C indicate that both materials should be suitable for melting iron phosphate glasses. (9) The viscosity and electrical conductivity of the MS26AZ102F-2 melt are within the acceptable limits for RSM

  12. Final Report - Testing of Optimized Bubbler Configuration for HLW Melter VSL-13R2950-1, Rev. 0, dated 6/12/2013

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Pegg, I. L.; Callow, R. A.; Joseph, I.; Matlack, K. S.; Kot, W. K.

    2013-11-13

    The principal objective of this work was to determine the glass production rate increase and ancillary effects of adding more bubbler outlets to the current WTP HLW melter baseline. This was accomplished through testing on the HLW Pilot Melter (DM1200) at VSL. The DM1200 unit was selected for these tests since it was used previously with several HLW waste streams including the four tank wastes proposed for initial processing at Hanford. This melter system was also used for the development and optimization of the present baseline WTP HLW bubbler configuration for the WTP HLW melter, as well as for MACT testing for both HLW and LAW. Specific objectives of these tests were to: Conduct DM1200 melter testing with the baseline WTP bubbling configuration and as augmented with additional bubblers. Conduct DM1200 melter testing to differentiate the effects of total bubbler air flow and bubbler distribution on glass production rate and cold cap formation. Collect melter operating data including processing rate, temperatures at a variety of locations within the melter plenum space, melt pool temperature, glass melt density, and melter pressure with the baseline WTP bubbling configuration and as augmented with additional bubblers. Collect melter exhaust samples to compare particulate carryover for different bubbler configurations. Analyze all collected data to determine the effects of adding more bubblers to the WTP HLW melter to inform decisions regarding future lid re-designs. The work used a high aluminum HLW stream composition defined by ORP, for which an appropriate simulant and high waste loading glass formulation were developed and have been previously processed on the DM1200.

  13. 1/6TH SCALE STRIP EFFLUENT FEED TANK-MIXING RESULTS USING MCU SOLVENT

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, E

    2006-02-01

    The purpose of this task was to determine if mixing was an issue for the entrainment and dispersion of the Modular Caustic Side Solvent Extraction (CSSX) Unit (MCU) solvent in the Defense Waste Processing Facility (DWPF) Strip Effluent Feed Tank (SEFT). The MCU strip effluent stream containing the Cs removed during salt processing will be transferred to the DWPF for immobilization in HLW glass. In lab-scale DWPF chemical process cell testing, mixing of the solvent in the dilute nitric acid solution proved problematic, and the Savannah River National Laboratory (SRNL) was requested to perform scaled SEFT mixing tests to evaluate whether the problem was symptomatic of the lab-scale set-up or of the solvent. The solvent levels tested were 228 and 235 ppm, which represented levels near the estimated DWPF solvent limit of 239 ppm in 0.001M HNO{sub 3} solution. The 239 ppm limit was calculated by Norato in X-CLC-S-00141. The general approach for the mixing investigation was to: (1) Investigate the use of fluorescent dyes to aid in observing the mixing behavior. Evaluate and compare the physical properties of the fluorescent dyed MCU solvents to the baseline Oak Ridge CSSX solvent. Based on the data, use the dyed MCU solvent that best approximates the physical properties. (2) Use approximately a 1/6th linear scale of the SEFT to replicate the internal configuration for DWPF mixing. (3) Determine agitator speed(s) for scaled testing based on the DWPF SEFT mixing speed. (4) Perform mixing tests using the 1/6th SEFT and determine any mixing issues (entrainment/dispersion, accumulation, adhesion) through visual observations and by pulling samples to assess uniformity. The mixing tests used MCU solvent fabricated at SRNL blended with Risk Reactor DFSB-K43 fluorescent dye. This dyed SRNL MCU solvent had equivalent physical properties important to mixing as compared to the Oak Ridge baseline solvent, blended easily with the MCU solvent, and provided an excellent visual aid.

  14. IMPACT OF IRRADIATION AND THERMAL AGING ON DWPF SIMULATED SLUDGE PROPERTIES

    International Nuclear Information System (INIS)

    Eibling, R; Michael Stone, M

    2006-01-01

    The research and development programs in support of the Defense Waste Processing Facility (DWPF) and other high-level waste vitrification processes require the use of both nonradioactive waste simulants and actual waste samples. While actual waste samples are the ideal materials to study, acquiring large quantities of actual waste is difficult and expensive. Tests utilizing actual high-level waste require the use of expensive shielded cells facilities to provide sufficient shielding for the researchers. Nonradioactive waste simulants have been used for laboratory testing, pilot-scale testing and full-scale integrated facility testing. These waste simulants were designed to reproduce the chemical and, if possible, the physical properties of the actual high-level waste. This technical report documents a study on the impact of irradiating a Sludge Batch 3 (SB3) simulant and of additional tests on aging a SB3 simulant by additional thermal processing. Prior simulant development studies examined methods of producing sludge and supernate simulants and processes that could be used to alter the physical properties of the simulant to more accurately mimic the properties of actual waste. Development of a precipitated sludge simulant for the River Protection Project (RPP) demonstrated that the application of heat for a period of time could significantly alter the rheology of the sludge simulant. The RPP precipitated simulant used distillation to concentrate the sludge solids and produced a reduction in sludge yield stress of up to 80% compared to the initial sludge properties. Observations at that time suggested that a substantial fraction of the iron hydroxide had converted to the oxide during the distillation. DWPF sludge simulant studies showed a much smaller reduction in yield stress (∼10%), demonstrated the impact of shear on particle size, and showed that smaller particle sizes yielded higher yield stress products. The current study documented in this report focuses

  15. Design and performance of feed-delivery systems for simulated radioactive waste slurries

    International Nuclear Information System (INIS)

    Perez, J.M. Jr.

    1983-02-01

    Processes for vitrifying simulated high-level radioactive waste have been developed at the Pacific Northwest Laboratory (PNL) over the last several years. Paralleling this effort, several feed systems used to deliver the simulated waste slurry to the melter have been tested. Because there had been little industrial experience in delivering abrasive slurries at feed rates of less than 10 L/min, early experience helped direct the design of more-dependable systems. Also, as feed delivery requirements changed, the feed system was modified to meet these new requirements. The various feed systems discussed in this document are part of this evolutionary process, so they have not been ranked against each other. The four slurry feed systems discussed are: (1) vertical-cantilevered centrifugal pump system; (2) airlift feed systems; (3) pressurized-loop systems; and (4) positive-displacement pump system. 20 figures, 11 tables

  16. US bureau of mines small-scale arc melter tests

    International Nuclear Information System (INIS)

    O'Connor, W.K.; Oden, L.L.; Turner, P.C.; Davis, D.L.

    1993-01-01

    The US Bureau of Mines, in cooperation with the Idaho National Engineering Laboratory (INEL), conducted over 30 hours of melting tests to vitrify simulated low-level radioactive wastes from the INEL. Radioactive Waste Management Complex (RWMC). Five separate waste compositions were investigated, each consisting of noncontaminated soil from the RWMC and surrogate materials used to simulate the actual buried wastes. The RWMC soil and five waste compositions were melted in a 50-lb, single-phase electric arc furnace with a water-cooled shell. These tests were conducted to determine melting parameters in preparation for a large-scale melting campaign to be conducted in the Bureau's 1-metric ton (mt), water-cooled-wall, 3-phase electric arc furnace. Bulk chemical composition was determined for each of the feed materials and for the slag, metal, fume solids, and offgas furnace products, and distributions were calculated for the key elements. The material balance for the furnace operation indicates that from 63 to 84 pct of the feed reported to the slag. Cerium, used as the surrogate for the radionuclides in the wastes, demonstrated an extremely strong affinity for the slag product. Although slag temperatures as low as 1,250 C were recorded when melting the RWMC soil, temperatures in excess of 1,600 C were necessary to achieve the fluidity required for a successful slag tap

  17. Determination of heat conductivity of waste glass feed and its applicability for modeling the batch-to-glass conversion

    Energy Technology Data Exchange (ETDEWEB)

    Hujova, Miroslava [Laboratory of Inorganic Materials, Joint Workplace of the University of Chemistry and Technology Prague and the Institute, Institute of Rock Structure and Mechanics of the ASCR, Prague Czech Republic; Pokorny, Richard [Laboratory of Inorganic Materials, Joint Workplace of the University of Chemistry and Technology Prague and the Institute, Institute of Rock Structure and Mechanics of the ASCR, Prague Czech Republic; Klouzek, Jaroslav [Laboratory of Inorganic Materials, Joint Workplace of the University of Chemistry and Technology Prague and the Institute, Institute of Rock Structure and Mechanics of the ASCR, Prague Czech Republic; Dixon, Derek R. [Radiological Materials & Detection Group, Pacific Northwest National Laboratory, Richland Washington; Cutforth, Derek A. [Radiological Materials & Detection Group, Pacific Northwest National Laboratory, Richland Washington; Lee, Seungmin [Radiological Materials & Detection Group, Pacific Northwest National Laboratory, Richland Washington; McCarthy, Benjamin P. [Radiological Materials & Detection Group, Pacific Northwest National Laboratory, Richland Washington; Schweiger, Michael J. [Radiological Materials & Detection Group, Pacific Northwest National Laboratory, Richland Washington; Kruger, Albert A. [U.S. Department of Energy, Office of River Protection, Richland Washington; Hrma, Pavel [Radiological Materials & Detection Group, Pacific Northwest National Laboratory, Richland Washington

    2017-07-10

    The heat conductivity of reacting melter feed affects the heat transfer and conversion process in the cold cap (the reacting feed floating on molten glass). To investigate it, we simulated the feed conditions and morphology in the cold-cap by preparing “fast-dried slurry blocks”, formed by rapidly evaporating water from feed slurry poured onto a 200°C surface. A heat conductivity meter was used to measure heat conductivity of samples cut from the fast-dried slurry blocks, samples of a cold cap retrieved from a laboratory-scale melter, and loose dry powder feed samples. Our study indicates that the heat conductivity of the feed in the cold cap is significantly higher than that of loose dry powder feed, resulting from the feed solidification during the water evaporation from the feed slurry. To assess the heat transfer at higher temperatures when feed turns into foam, we developed a theoretical model that predicts the foam heat conductivity based on morphology data from in-situ X-ray computed tomography. The implications for the mathematical modeling of the cold cap are discussed.

  18. Design and performance of a 100-kg/h, direct calcine-fed electric-melter system for nuclear-waste vitrification

    International Nuclear Information System (INIS)

    Dierks, R.D.

    1980-11-01

    This report describes the physical characteristics of a ceramic-lined, joule-heated glass melter that is directly connected to the discharge of a spray calciner and is currently being used to study the vitrification of simulated nuclear-waste slurries. Melter performance characteristics and subsequent design improvements are described. The melter contains 0.24 m 3 of glass with a glass surface area of 0.76 m 2 , and is heated by the flow of an alternating current (ranging from 600 to 1200 amps) between two Inconel-690 slab-type electrodes immersed in the glass at either end of the melter tank. The melter was maintained at operating temperature (900 to 1260 0 C) for 15 months, and produced 62,000 kg of glass. The maximum sustained operating period was 122 h, during which glass was produced at the rate of 70 kg/h

  19. Analyses and Comparison of Bulk and Coil Surface Samples from the DWPF Slurry Mix Evaporator

    International Nuclear Information System (INIS)

    Hay, M.; Nash, C.; Stone, M.

    2012-01-01

    Sludge samples from the DWPF Slurry Mix Evaporator (SME) heating coil frame and coil surface were characterized to identify differences that might help identify heat transfer fouling materials. The SME steam coils have seen increased fouling leading to lower boil-up rates. Samples of the sludge were taken from the coil frame somewhat distant from the coil (bulk tank material) and from the coil surface (coil surface sample). The results of the analysis indicate the composition of the two SME samples are very similar with the exception that the coil surface sample shows ∼5-10X higher mercury concentration than the bulk tank sample. Elemental analyses and x-ray diffraction results did not indicate notable differences between the two samples. The ICP-MS and Cs-137 data indicate no significant differences in the radionuclide composition of the two SME samples. Semi-volatile organic analysis revealed numerous organic molecules, these likely result from antifoaming additives. The compositions of the two SME samples also match well with the analyzed composition of the SME batch with the exception of significantly higher silicon, lithium, and boron content in the batch sample indicating the coil samples are deficient in frit relative to the SME batch composition.

  20. Determination of Reportable Radionuclides for DWPF Sludge Batch 2 (Macro Batch 3)

    International Nuclear Information System (INIS)

    Bibler, N.E.

    2002-01-01

    The Waste Acceptance Product Specifications (WAPS) 1.2 require that ''The Producer shall report the inventory of radionuclides (in Curies) that have half-lives longer than 10 years and that are, or will be, present in concentrations greater than 0.05 percent of the total inventory for each waste type indexed to the years 2015 and 3115''. As part of the strategy to meet WAPS 1.2, the Defense Waste Processing Facility (DWPF) will report for each waste type, all radionuclides (with half-lives greater than 10 years) that have concentrations greater than 0.01 percent of the total inventory from time of production through the 1100 year period from 2015 through 3115. The initial listing of radionuclides to be included is based on the design-basis glass as identified in the Waste Form Compliance Plan (WCP) and Waste Form Qualification Report (WQR). However, it is required that this list be expanded if other radionuclides with half-lives greater than 10 years are identified that meet the greater than 0.01 percent criterion for Curie content

  1. Integration of SWPF into the DWPF Flowsheet: Gap Analysis and Test Matrix Development

    Energy Technology Data Exchange (ETDEWEB)

    Peeler, D. K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-12-10

    Based on Revision 19 of the High Level Waste (HLW) System Plan, it is anticipated that the Salt Waste Processing Facility (SWPF) will be integrated into the Defense Waste Processing Facility (DWPF) flowsheet in October 2018 (or with Sludge Batch 11 (SB11)). Given that, Savannah River Remediation (SRR) has requested a technical basis be developed that validates the current Product Composition Control System (PCCS) models for use during the processing of the SWPF-based coupled flowsheet or that leads to the refinements of or modifications to the models that are needed so that the models may be used during the processing of the SWPF-based coupled flowsheet. To support this objective, Savannah River National Laboratory (SRNL) has completed three key interim activities prior to validation of the current or development of refined PCCS models over the anticipated glass composition region for SWPF processing. These three key activities include: (1) defining the glass compositional region over which SWPF is anticipated to be processed, (2) comparing the current PCCS model validation ranges to the SWPF glass compositional region from which compositional gaps can be identified, and (3) developing a test matrix to cover the compositional gaps.

  2. Rheology of Savannah River Site Tank 51 HLW radioactive sludge

    International Nuclear Information System (INIS)

    Ha, B.C.

    1993-01-01

    Savannah River Site (SRS) Tank 51 HLW radioactive sludge represents a major portion of the first batch of sludge to be vitrified in the Defense Waste Processing Facility (DWPF) at SRS. The rheological properties of Tank 51 sludge will determine if the waste sludge can be pumped by the current DWPF process cell pump design and the homogeneity of melter feed slurries. The rheological properties of Tank 51 sludge and sludge/frit slurries at various solids concentrations were measured remotely in the Shielded Cells Operations (SCO) at the Savannah River Technology Center (SRTC) using a modified Haake Rotovisco viscometer system. Rheological properties of Tank 51 radioactive sludge/Frit 202 slurries increased drastically when the solids content was above 41 wt %. The yield stresses of Tank 51 sludge and sludge/frit slurries fall within the limits of the DWPF equipment design basis. The apparent viscosities also fall within the DWPF design basis for sludge consistency. All the results indicate that Tank 51 waste sludge and sludge/frit slurries are pumpable throughout the DWPF processes based on the current process cell pump design, and should produce homogeneous melter feed slurries

  3. ALTERNATIVE ANALYTICAL DIGESTION SCHEME FOR THE DEFENSE WASTE PROCESSING FACILITY (DWPF) SLURRY RECEIPT AND ADJUSTMENT TANK (SRAT) ANALYSES

    International Nuclear Information System (INIS)

    Click, D; Charles02 Coleman, C; Frank Pennebaker, F; Kristine Zeigler, K; Tommy Edwards, T

    2007-01-01

    As part of the radioactive sludge batch qualification, Savannah River National Laboratory (SRNL) performs a verification of the digestion methods to be used by the Defense Waste Processing Facility (DWPF) Lab for elemental analysis of Sludge Receipt and Adjustment Tank (SRAT) receipt process control samples and SRAT product process control samples. Verification of these methods on Sludge Batch 4 (SB4) radioactive sludge slurry indicated SB4 contains a higher concentration of aluminum (Al) than previous sludge batches. Aluminum plays a direct role in vitrification chemistry. At moderate levels, Al assists in glass forming, but at elevated levels Al can increase the viscosity of the molten glass which can adversely impact glass production rate and the volume of glass produced via limiting waste loading.3 Most of the Al present in SB4 is in the form of Al hydroxide as a mixture of gibbsite [α-aluminum trihydroxide, α-Al(OH) 3 ] and boehmite (α-aluminum oxyhydroxide, α-AlOOH) in an unknown ratio. Testing done at SRNL indicates Gibbsite is soluble at low pH but boehmite has limited solubility in the acid mixture (DWPF Cold Chem Method (CC), 25 mL nitric acid (HNO 3 ) and 25 mL hydrofluoric acid (HF)) used by DWPF to digest process control samples. Because Al plays such an important part in vitrification chemistry, it is necessary to have a robust digestion method that will dissolve all forms of Al present in the radioactive sludge while not increasing the analytical lab turnaround time. SRNL initially suggested that the DWPF lab use the sodium peroxide/hydroxide fusion (PF) digestion method4 to digest SRAT receipt and SRAT product radioactive sludge as an alternative to the acid digestion method to ensure complete digestion based on results obtained from digesting a SB4 radioactive sample.2 However, this change may have a significant impact on the DWPF lab analytical turnaround time due to the inefficiency in drying the radioactive sludge contained in a peanut

  4. Oxygen enriched combustion system performance study. Phase 2: 100 percent oxygen enriched combustion in regenerative glass melters, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Tuson, G.B.; Kobayashi, H.; Campbell, M.J.

    1994-08-01

    The field test project described in this report was conducted to evaluate the energy and environmental performance of 100% oxygen enriched combustion (100% OEC) in regenerative glass melters. Additional objectives were to determine other impacts of 100% OEC on melter operation and glass quality, and to verify on a commercial scale that an on-site Pressure Swing Adsorption oxygen plant can reliably supply oxygen for glass melting with low electrical power consumption. The tests constituted Phase 2 of a cooperative project between the United States Department of Energy, and Praxair, Inc. Phase 1 of the project involved market and technical feasibility assessments of oxygen enriched combustion for a range of high temperature industrial heating applications. An assessment of oxygen supply options for these applications was also performed during Phase 1, which included performance evaluation of a pilot scale 1 ton per day PSA oxygen plant. Two regenerative container glass melters were converted to 100% OEC operation and served as host sites for Phase 2. A 75 ton per day end-fired melter at Carr-Lowrey Glass Company in Baltimore, Maryland, was temporarily converted to 100% OEC in mid- 1990. A 350 tpd cross-fired melter at Gallo Glass Company in Modesto, California was rebuilt for permanent commercial operation with 100% OEC in mid-1991. Initially, both of these melters were supplied with oxygen from liquid storage. Subsequently, in late 1992, a Pressure Swing Adsorption oxygen plant was installed at Gallo to supply oxygen for 100% OEC glass melting. The particular PSA plant design used at Gallo achieves maximum efficiency by cycling the adsorbent beds between pressurized and evacuated states, and is therefore referred to as a Vacuum/Pressure Swing Adsorption (VPSA) plant.

  5. Improvement of melter off-gas design for commercial HALW vitrification facility

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, A.; Kitamura, M.; Yamanaka, T. [Ishikawajima-Harima Heavy Industries Co., Ltd., Yokohama (Japan); Yoshioka, M.; Endo, N.; Asano, N. [Japan Nuclear Cycle Development Institute, Ibaraki (Japan)

    2001-07-01

    The Japan commercial reprocessing plant is now under construction, and it will commence the operation in 2005. The High Active Liquid Waste (HALW) generated at the plant is treated into glass product at the vitrification facility using the Liquid Fed Joule-Heated Ceramic Melter (LFCM). The characteristic of the LFCM is that the HALW is fed directly onto the molten glass surface with the glass forming material. This process was developed by the Japan Nuclear Cycle Development Institute (JNC). The JNC process was first applied to the Tokai Vitrification Facility (TVF), which is a pilot scale plant having about 1/6 capacity of the commercial facility. The TVF has been in operation since 1995. During the operation, the rapid increase of the differential pressure between the melter plenum and the dust scrubber was observed. This phenomenon is harmful to the long-term continuous operation of TVF. And, it is also anticipated that the same phenomenon will occur in commercial vitrification facility. In order to solve this problem, the countermeasures were studied and developed. Through the study on the deposit growing mechanism, it was probable that the rapid increased differential pressure was attributed to the condensation of meta-boric acid at the outlet of the air-film cooler slits. And, the heating and the humidification of purge air were judged to be effective as the countermeasures to suppress the condensation. On the other hand, the water injection into melter off-gas pipe was found to be very effective to reduce the differential pressure as the results of the various tests. The deposit adhered on the inner surface of the off-gas pipe was almost washed out. And, it was also demonstrated that the system was superior to other systems by virtue of its simplicity and stability. In order to apply the system to the commercial scale plant, the scale-up tests were conducted at JNC mock-up facility using the acrylic model. (author)

  6. Improvement of melter off-gas design for commercial HALW vitrification facility

    International Nuclear Information System (INIS)

    Ohno, A.; Kitamura, M.; Yamanaka, T.; Yoshioka, M.; Endo, N.; Asano, N.

    2001-01-01

    The Japan commercial reprocessing plant is now under construction, and it will commence the operation in 2005. The High Active Liquid Waste (HALW) generated at the plant is treated into glass product at the vitrification facility using the Liquid Fed Joule-Heated Ceramic Melter (LFCM). The characteristic of the LFCM is that the HALW is fed directly onto the molten glass surface with the glass forming material. This process was developed by the Japan Nuclear Cycle Development Institute (JNC). The JNC process was first applied to the Tokai Vitrification Facility (TVF), which is a pilot scale plant having about 1/6 capacity of the commercial facility. The TVF has been in operation since 1995. During the operation, the rapid increase of the differential pressure between the melter plenum and the dust scrubber was observed. This phenomenon is harmful to the long-term continuous operation of TVF. And, it is also anticipated that the same phenomenon will occur in commercial vitrification facility. In order to solve this problem, the countermeasures were studied and developed. Through the study on the deposit growing mechanism, it was probable that the rapid increased differential pressure was attributed to the condensation of meta-boric acid at the outlet of the air-film cooler slits. And, the heating and the humidification of purge air were judged to be effective as the countermeasures to suppress the condensation. On the other hand, the water injection into melter off-gas pipe was found to be very effective to reduce the differential pressure as the results of the various tests. The deposit adhered on the inner surface of the off-gas pipe was almost washed out. And, it was also demonstrated that the system was superior to other systems by virtue of its simplicity and stability. In order to apply the system to the commercial scale plant, the scale-up tests were conducted at JNC mock-up facility using the acrylic model. (author)

  7. DC Graphite Arc Melter for vitrification of low-level waste

    International Nuclear Information System (INIS)

    Desrosiers, A.E.; Wilver, P.J.; Wittle, J.K.

    1996-01-01

    The volume of mixed waste continues to increase with few options for its permanent disposal other than storage on site. This mixed waste is being generated by not only the Department of Energy at government sites but by the private sector in hospitals and at electrical utility sites. Bartlett Services, Inc. proposes to offer a service to treat these materials to both reduce the volume and stabilize the radionuclides in a vitrified material. This product will be formed in the DC Graphite Arc Melters developed by Electro-Pyrolysis, Inc. and being offered for commercial design, sale and installation by Svedala Industries, Pyro Division. The process is a high temperature procedure which pyrolytically decomposes the organic portion of the waste to form clean hydrogen and carbon monoxide and solid carbon. The inorganic portion, containing the radioactive components, melts to produce a stable glass which is resistant to environmental leaching and will remain stable until the radioactivity has decreased to a safe level. Glasses produced with surrogate materials such as cesium and cerium have been shown to pass the Product Compatibility Test (PCT). The process being proposed for this treatment utilizes a sealed melter system having the capability of melting wastes containing both metallic and inorganic materials. This process, unlike joule heated melters, is capable of operating to temperatures of 1600 degrees C or higher. Since the system is heated electrically, oxidation is not required to create the heat. Since the system is pyrolytic, relatively small quantities of gas are produced. These gases may have beneficial uses in producing chemicals or may be used as a clean fuel

  8. Crystal accumulation in the Hanford Waste Treatment Plant high level waste melter: Summary of 2017 experiments

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fowley, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2018-01-11

    A full-scale, transparent mock-up of the Hanford Tank Waste Treatment and Immobilization Project High Level Waste glass melter riser and pour spout has been constructed to allow for testing with visual feedback of particle settling, accumulation, and resuspension when operating with a controlled fraction of crystals in the glass melt. Room temperature operation with silicone oil and magnetite particles simulating molten glass and spinel crystals, respectively, allows for direct observation of flow patterns and settling patterns. The fluid and particle mixture is recycled within the system for each test.

  9. DEMONSTRATION AND EVALUATION OF POTENTIAL HIGH LEVEL WASTE MELTER DECONTAMINATION TECHNOLOGIES FOR SAVANNAH RIVER SITE

    International Nuclear Information System (INIS)

    Weger, Hans; Kodanda, Raja Tilek Meruva; Mazumdar, Anindra; Srivastava, Rajiv Ph.D.; Ebadian, M.A. Ph.D.

    2003-01-01

    Four hand-held tools were tested for failed high-level waste melter decontamination and decommissioning (D and D). The forces felt by the tools during operation were measured using a tri-axial accelerometer since they will be operated by a remote manipulator. The efficiency of the tools was also recorded. Melter D and D consists of three parts: (1) glass fracturing: removing from the furnace the melted glass that can not be poured out through normal means, (2) glass cleaning: removing the thin layer of glass that has formed over the surface of the refractory material, and (3) K-3 refractory breakup: removing the K-3 refractory material. Surrogate glass, from a formula provided by the Savannah River Site, was melted in a furnace and poured into steel containers. K-3 refractory material, the same material used in the Defense Waste Processing Facility, was utilized for the demonstrations. Four K-3 blocks were heated at 1150 C for two weeks with a glass layer on top to simulate the hardened glass layer on the refractory surface in the melter. Tools chosen for the demonstrations were commonly used D and D tools, which have not been tested specifically for the different aspects of melter D and D. A jackhammer and a needle gun were tested for glass fracturing; a needle gun and a rotary grinder with a diamond face wheel (diamond grinder) were tested for glass cleaning; and a jackhammer, diamond grinder, and a circular saw with a diamond blade were tested for refractory breakup. The needle gun was not capable of removing or fracturing the surrogate glass. The diamond grinder only had a removal rate of 3.0 x 10-4 kg/s for K-3 refractory breakup and needed to be held firmly against the material. However, the diamond grinder was effective for glass cleaning, with a removal rate of 3.9 cm2/s. The jackhammer was successful in fracturing glass and breaking up the K-3 refractory block. The jackhammer had a glass-fracturing rate of 0.40 kg/s. The jackhammer split the K-3 refractory

  10. Alternate Reductant Cold Cap Evaluation Furnace Phase II Testing

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, F. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Stone, M. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Miller, D. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-09-03

    Savannah River Remediation (SRR) conducted a Systems Engineering Evaluation (SEE) to determine the optimum alternate reductant flowsheet for the Defense Waste Processing Facility (DWPF). Specifically, two proposed flowsheets (nitric–formic–glycolic and nitric–formic–sugar) were evaluated based upon results from preliminary testing. Comparison of the two flowsheets among evaluation criteria indicated a preference towards the nitric–formic–glycolic flowsheet. Further research and development of this flowsheet eliminated the formic acid, and as a result, the nitric–glycolic flowsheet was recommended for further testing. Based on the development of a roadmap for the nitric–glycolic acid flowsheet, Waste Solidification Engineering (WS-E) issued a Technical Task Request (TTR) to address flammability issues that may impact the implementation of this flowsheet. Melter testing was requested in order to define the DWPF flammability envelope for the nitric-glycolic acid flowsheet. The Savannah River National Laboratory (SRNL) Cold Cap Evaluation Furnace (CEF), a 1/12th scale DWPF melter, was selected by the SRR Alternate Reductant project team as the melter platform for this testing. The overall scope was divided into the following sub-tasks as discussed in the Task Technical and Quality Assurance Plan (TTQAP): Phase I - A nitric–formic acid flowsheet melter test (unbubbled) to baseline the CEF cold cap and vapor space data to the benchmark melter flammability models; Phase II - A nitric–glycolic acid flowsheet melter test (unbubbled and bubbled) to: Define new cold cap reactions and global kinetic parameters in support of the melter flammability model development; Quantify off-gas surging potential of the feed; Characterize off-gas condensate for complete organic and inorganic carbon species. After charging the CEF with cullet from Phase I CEF testing, the melter was slurry-fed with glycolic flowsheet based SB6-Frit 418 melter feed at 36% waste

  11. Miscibility Evaluation Of The Next Generation Solvent With Polymers Currently Used At DWPF, MCU, And Saltstone

    Energy Technology Data Exchange (ETDEWEB)

    Fondeur, F. F.

    2013-04-17

    The Office of Waste Processing, within the Office of Technology Innovation and Development, funded the development of an enhanced Caustic-Side Solvent Extraction (CSSX) solvent for deployment at the Savannah River Site for removal of cesium from High Level Waste. This effort lead to the development of the Next Generation Solvent (NGS) with Tris (3,7-dimethyl octyl) guanidine (TiDG). The first deployment target for the NGS solvent is within the Modular CSSX Unit (MCU). Deployment of a new chemical within an existing facility requires verification that the new chemical components are compatible with the installed equipment. In the instance of a new organic solvent, the primary focus is on compatibility of the solvent with organic polymers used in the affected facility. This report provides the calculated data from exposing these polymers to the Next Generation Solvent. An assessment of the dimensional stability of polymers known to be used or present in the MCU, Defense Waste Processing Facility (DWPF), and Saltstone facilities that will be exposed to the NGS showed that TiDG could selectively affect the elastomers and some thermoplastics to varying extents, but the typical use of these polymers in a confined geometry will likely prevent the NGS from impacting component performance. The polymers identified as of primary concern include Grafoil® (flexible graphite), Tefzel®, Isolast®, ethylene-propylene-diene monomer (EPDM) rubber, nitrile-butadiene rubber (NBR), styrene-butadiene rubber (SBR), ultra high molecular weight polyethylene (UHMWPE), and fluorocarbon rubber (FKM). Certain polymers like NBR and EPDM were found to interact mildly with NGS but their calculated swelling and the confined geometry will impede interaction with NGS. In addition, it was found that Vellumoid (cellulose fibers-reinforced glycerin and protein) may leach protein and Polyvinyl Chloride (PVC) may leach plasticizer (such as Bis-Ethylhexyl-Phthalates) into the NGS solvent. Either case

  12. An Assessment of Using Vibrational Compaction of Calcined HLW and LLW in DWPF Canisters

    International Nuclear Information System (INIS)

    Yi, Yun-Bo; Amme, Robert C.; Shayer, Zeev

    2008-01-01

    Since 1963, the INEL has calcined almost 8 million gallons of liquid mixed waste and liquid high-level waste, converting it to some 1.1 million gallons of dry calcine (about 4275.0 m3), which consists of alumina-and zirconia-based calcine and zirconia-sodium blend calcine. In addition, if all existing and projected future liquid wastes are solidified, approximately 2,000 m3 of additional calcine will be produced primarily from sodium-bearing waste. Calcine is a more desirable material to store than liquid radioactive waste because it reduces volume, is much less corrosive, less chemically reactive, less mobile under most conditions, easier to monitor and more protective of human health and the environment. This paper describes the technical issue involved in the development of a feasible solution for further volume reduction of calcined nuclear waste for transportation and long term storage, using a standard DWPF canister. This will be accomplished by developing a process wherein the canisters are transported into a vibrational machine, for further volume reduction by about 35%. The random compaction experiments show that this volume reduction is achievable. The main goal of this paper is to demonstrate through computer modeling that it is feasible to use volume reduction vibrational machine without developing stress/strain forces that will weaken the canister integrity. Specifically, the paper presents preliminary results of the stress/strain analysis of the DWPF canister as a function of granular calcined height during the compaction and verifying that the integrity of the canister is not compromised. This preliminary study will lead to the development of better technology for safe compactions of nuclear waste that will have significant economical impact on nuclear waste storage and treatment. The preliminary results will guide us to find better solutions to the following questions: 1) What are the optimum locations and directions (vertical versus horizontal or

  13. Hanford Waste Vitrification Plant full-scale feed preparation testing with water and process simulant slurries

    International Nuclear Information System (INIS)

    Gaskill, J.R.; Larson, D.E.; Abrigo, G.P.

    1996-03-01

    The Hanford Waste Vitrification Plant was intended to convert selected, pretreated defense high-level waste and transuranic waste from the Hanford Site into a borosilicate glass. A full-scale testing program was conducted with nonradioactive waste simulants to develop information for process and equipment design of the feed-preparation system. The equipment systems tested included the Slurry Receipt and Adjustment Tank, Slurry Mix Evaporator, and Melter-Feed Tank. The areas of data generation included heat transfer (boiling, heating, and cooling), slurry mixing, slurry pumping and transport, slurry sampling, and process chemistry. 13 refs., 129 figs., 68 tabs

  14. ALTERNATE REDUCTANT COLD CAP EVALUATION FURNACE PHASE I TESTING

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, F.; Miller, D.; Zamecnik, J.; Lambert, D.

    2014-04-22

    Savannah River Remediation (SRR) conducted a Systems Engineering Evaluation (SEE) to determine the optimum alternate reductant flowsheet for the Defense Waste Processing Facility (DWPF). Specifically, two proposed flowsheets (nitric–formic–glycolic and nitric–formic–sugar) were evaluated based upon results from preliminary testing. Comparison of the two flowsheets among evaluation criteria indicated a preference towards the nitric–formic–glycolic flowsheet. Further evaluation of this flowsheet eliminated the formic acid1, and as a result, the nitric–glycolic flowsheet was recommended for further testing. Based on the development of a roadmap for the nitric–glycolic acid flowsheet, Waste Solidification Engineering (WS-E) issued a Technical Task Request (TTR) to address flammability issues that may impact the implementation of this flowsheet. Melter testing was requested in order to define the DWPF flammability envelope for the nitric glycolic acid flowsheet. The Savannah River National Laboratory (SRNL) Cold Cap Evaluation Furnace (CEF), a 1/12th scale DWPF melter, was selected by the SRR Alternate Reductant project team as the melter platform for this testing. The overall scope was divided into the following sub-tasks as discussed in the Task Technical and Quality Assurance Plan (TTQAP): Phase I - A nitric–formic acid flowsheet melter test (unbubbled) to baseline the Cold Cap Evaluation Furnace (CEF) cold cap and vapor space data to the benchmark melter flammability models Phase II - A nitric–glycolic acid flowsheet melter test (unbubbled and bubbled) to: o Define new cold cap reactions and global kinetic parameters for the melter flammability models o Quantify off-gas surging potential of the feed o Characterize off-gas condensate for complete organic and inorganic carbon species Prior to startup, a number of improvements and modifications were made to the CEF, including addition of cameras, vessel support temperature measurement, and a heating

  15. Multiphase, multi-electrode Joule heat computations for glass melter and in situ vitrification simulations

    International Nuclear Information System (INIS)

    Lowery, P.S.; Lessor, D.L.

    1991-02-01

    Waste glass melter and in situ vitrification (ISV) processes represent the combination of electrical thermal, and fluid flow phenomena to produce a stable waste-from product. Computational modeling of the thermal and fluid flow aspects of these processes provides a useful tool for assessing the potential performance of proposed system designs. These computations can be performed at a fraction of the cost of experiment. Consequently, computational modeling of vitrification systems can also provide and economical means for assessing the suitability of a proposed process application. The computational model described in this paper employs finite difference representations of the basic continuum conservation laws governing the thermal, fluid flow, and electrical aspects of the vitrification process -- i.e., conservation of mass, momentum, energy, and electrical charge. The resulting code is a member of the TEMPEST family of codes developed at the Pacific Northwest Laboratory (operated by Battelle for the US Department of Energy). This paper provides an overview of the numerical approach employed in TEMPEST. In addition, results from several TEMPEST simulations of sample waste glass melter and ISV processes are provided to illustrate the insights to be gained from computational modeling of these processes. 3 refs., 13 figs

  16. Effects of feed process variables on Hanford Vitrification Plant performance

    International Nuclear Information System (INIS)

    Farnsworth, R.K.; Peterson, M.E.; Wagner, R.N.

    1987-01-01

    As a result of nuclear defense activities, high-level liquid radioactive wastes have been generated at the Hanford Site for over 40 yr. The Hanford Waste Vitrification Plant (HWVP) is being proposed to immobilize these wastes in a waste form suitable for disposal in a geologic repository. Prior to vitrification, the waste will undergo several conditioning steps before being fed to the melter. The effect of certain process variables on the resultant waste slurry properties must be known to assure processability of the waste slurry during feed preparation. Of particular interest are the rheological properties, which include the yield stress and apparent viscosity. Identification of the rheological properties of the slurry is required to adequately design the process equipment used for feed preparation (agitators, mixing tanks, concentrators, etc.). Knowledge of the slurry rheological properties is also necessary to establish processing conditions and operational limits for maximum plant efficiency and reliability. A multivariable study was performed on simulated HWVP feed to identify the feed process variables that have a significant impact on rheology during processing. Two process variables were evaluated in this study: (a) the amount of formic acid added to the feed and (b) the degree of shear encountered by the feed during processing. The feed was physically and rheologically characterized at various stages during feed processing

  17. Corrosion study for a radioactive waste vitrification facility

    International Nuclear Information System (INIS)

    Imrich, K.J.; Jenkins, C.F.

    1993-01-01

    A corrosion monitoring program was setup in a scale demonstration melter system to evaluate the performance of materials selected for use in the Defense Waste Processing Facility (DWPF) at the DOE's Savannah River Site. The system is a 1/10 scale prototypic version of the DWPF. In DWPF, high activity radioactive waste will be vitrified and encapsulated for long term storage. During this study twenty-six different alloys, including DWPF reference materials of construction and alternate higher alloy materials, were subjected to process conditions and environments characteristic of the DWPF except for radioactivity. The materials were exposed to low pH, elevated temperature (to 1200 degree C) environments containing abrasive slurries, molten glass, mercury, halides and sulfides. General corrosion rates, pitting susceptibility and stress corrosion cracking of the materials were investigated. Extensive data were obtained for many of the reference materials. Performance in the Feed Preparation System was very good, whereas coupons from the Quencher Inlet region of the Melter Off-Gas System experienced localized attack

  18. Dew point, internal gas pressure, and chemical composition of the gas within the free volume of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    Harbour, J.R.; Herman, D.T.; Crump, S.; Miller, T.J.; McIntosh, J.

    1996-01-01

    The Defense Waste Processing Facility (DWPF) produced 55 canistered waste forms containing simulated waste glass during the four Waste Qualification campaigns of the DWPF Startup Test Program. Testing of the gas within the free volume of these canisters for dew point, internal gas pressure, and chemical composition was performed as part of a continuing effort to demonstrate compliance with the Waste Acceptance Product Specifications. Results are presented for six glass-filled canisters. The dew points within the canisters met the acceptance criterion of < 20 degrees C for all six canisters. Factors influencing the magnitude of the dew point are presented. The chemical composition of the free volume gas was indistinguishable from air for all six canisters. Hence, no foreign materials were present in the gas phase of these canisters. The internal gas pressures within the sealed canisters were < 1 atm at 25 degrees C for all six canisters which readily met the acceptance criterion of an internal gas pressure of less than 1.5 atm at 25 degrees C. These results provided the evidence required to demonstrate compliance with the Waste Acceptance Product Specifications

  19. Initial results from the canistered waste forms produced during the first campaign of the DWPF Startup Test Program

    International Nuclear Information System (INIS)

    Harbour, J.R.

    1995-01-01

    As part of the Defense Waste Processing Facility (DWPF) Startup Test Program, approximately 90 canisters will be filled with glass containing simulated radioactive waste during five separate campaigns. The first campaign is a facility acceptance test to demonstrate the operability of the facility and to collect initial data on the glass and the canistered waste forms. During the next four campaigns (the waste qualification campaigns) data will be obtained which will be used to demonstrate that the DWPF product meets DOE's Waste Acceptance Product Specifications (WAPS). Currently 12 of the 16 canisters have been filled with glass during the first campaign (FA-13). This paper describes the tests that have been carried out on these 12 glass-filled canisters and presents the data with reference to the acceptance criteria of the WAPS. These tests include measurement of canister dimensions prior to and after glass filling. dew point, composition, and pressure of the gas within the free volume of the canister, fill height, free volume, weight, leak rates of welds and temporary seals, and weld parameters

  20. Characterization of the Defense Waste Processing Facility (DWPF) Environmental Assessment (EA) glass Standard Reference Material. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Crawford, C.L.; Pickett, M.A.

    1993-06-01

    Liquid high-level nuclear waste at the Savannah River Site (SRS) will be immobilized by vitrification in borosilicate glass. The glass will be produced and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). Other waste form producers, such as West Valley Nuclear Services (WVNS) and the Hanford Waste Vitrification Project (HWVP), will also immobilize high-level radioactive waste in borosilicate glass. The canistered waste will be stored temporarily at each facility for eventual permanent disposal in a geologic repository. The Department of Energy has defined a set of requirements for the canistered waste forms, the Waste Acceptance Product Specifications (WAPS). The current Waste Acceptance Primary Specification (WAPS) 1.3, the product consistency specification, requires the waste form producers to demonstrate control of the consistency of the final waste form using a crushed glass durability test, the Product Consistency Test (PCI). In order to be acceptable, a waste glass must be more durable during PCT analysis than the waste glass identified in the DWPF Environmental Assessment (EA). In order to supply all the waste form producers with the same standard benchmark glass, 1000 pounds of the EA glass was fabricated. The chemical analyses and characterization of the benchmark EA glass are reported. This material is now available to act as a durability and/or redox Standard Reference Material (SRM) for all waste form producers.

  1. FY13 GLYCOLIC-NITRIC ACID FLOWSHEET DEMONSTRATIONS OF THE DWPF CHEMICAL PROCESS CELL WITH SIMULANTS

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, D.; Zamecnik, J.; Best, D.

    2014-03-13

    Savannah River Remediation is evaluating changes to its current Defense Waste Processing Facility flowsheet to replace formic acid with glycolic acid in order to improve processing cycle times and decrease by approximately 100x the production of hydrogen, a potentially flammable gas. Higher throughput is needed in the Chemical Processing Cell since the installation of the bubblers into the melter has increased melt rate. Due to the significant maintenance required for the safety significant gas chromatographs and the potential for production of flammable quantities of hydrogen, eliminating the use of formic acid is highly desirable. Previous testing at the Savannah River National Laboratory has shown that replacing formic acid with glycolic acid allows the reduction and removal of mercury without significant catalytic hydrogen generation. Five back-to-back Sludge Receipt and Adjustment Tank (SRAT) cycles and four back-to-back Slurry Mix Evaporator (SME) cycles were successful in demonstrating the viability of the nitric/glycolic acid flowsheet. The testing was completed in FY13 to determine the impact of process heels (approximately 25% of the material is left behind after transfers). In addition, back-to-back experiments might identify longer-term processing problems. The testing was designed to be prototypic by including sludge simulant, Actinide Removal Product simulant, nitric acid, glycolic acid, and Strip Effluent simulant containing Next Generation Solvent in the SRAT processing and SRAT product simulant, decontamination frit slurry, and process frit slurry in the SME processing. A heel was produced in the first cycle and each subsequent cycle utilized the remaining heel from the previous cycle. Lower SRAT purges were utilized due to the low hydrogen generation. Design basis addition rates and boilup rates were used so the processing time was shorter than current processing rates.

  2. DWPF Flowsheet Studies with Simulants to Determine Modular Caustic Side Solvent Extraction Unit Solvent Partitioning and Verify Actinide Removal Process Incorporation Strategy

    International Nuclear Information System (INIS)

    Herman, C

    2006-01-01

    The Actinide Removal Process (ARP) facility and the Modular Caustic Side Solvent Extraction Unit (MCU) are scheduled to begin processing salt waste in fiscal year 2007. A portion of the streams generated in the salt processing facilities will be transferred to the Defense Waste Processing Facility (DWPF) to be incorporated in the glass matrix. Before the streams are introduced, a combination of impact analyses and research and development studies must be performed to quantify the impacts on DWPF processing. The Process Science and Engineering (PS and E) section of the Savannah River National Laboratory (SRNL) was requested via Technical Task Request (TTR) HLW/DWPF/TTR-2004-0031 to evaluate the impacts on DWPF processing. Simulant Chemical Process Cell (CPC) flowsheet studies have been performed using previous composition and projected volume estimates for the ARP sludge/monosodium titanate (MST) stream. Due to changes in the flammability control strategy for DWPF for salt processing, the incorporation strategy for ARP has changed and additional ARP flowsheet tests were necessary to validate the new processing strategy. The last round of ARP testing included the incorporation of the MCU stream and identified potential processing issues with the MCU solvent. The identified issues included the potential carry-over and accumulation of the MCU solvent components in the CPC condensers and in the recycle stream to the Tank Farm. Therefore, DWPF requested SRNL to perform additional MCU flowsheet studies to better quantify the organic distribution in the CPC vessels. The previous MCU testing used a Sludge Batch 4 (SB4) simulant since it was anticipated that both of these facilities would begin salt processing during SB4 processing. The same sludge simulant recipe was used in this round of ARP and MCU testing to minimize the number of changes between the two phases of testing so a better comparison could be made. ARP and MCU stream simulants were made for this phase of

  3. Feeding Tubes

    Science.gov (United States)

    ... feeding therapies have been exhausted. Please review product brand and method of placement carefully with your physician ... Total Parenteral Nutrition. Resources: Oley Foundation Feeding Tube Awareness Foundation Children’s Medical Nutrition Alliance APFED’s Educational Webinar ...

  4. Hydrogen generation in SRAT with nitric acid and late washing flowsheets

    International Nuclear Information System (INIS)

    Hsu, C.W.

    1992-01-01

    Melter feed preparation processes, incorporating a final wash of the precipitate slurry feed to Defense Waste Processing Facility (DWPF) and a partial substitution of the SRAT formic acid requirement with nitric acid, should not produce peak hydrogen generation rates during Cold Chemical Runs (CCR's) and radioactive operation greater than their current, respective hydrogen design bases of 0.024 lb/hr and 1.5 lb/hr. A single SRAT bench-scale process simulation for CCR-s produced a DWPF equivalent peak hydrogen generation rate of 0.004 lb/hr. During radioactive operation, the peak hydrogen generation rate will be dependent on the extent DWPF deviates from the nominal precipitate hydrolysis and melter feed preparation process operating parameters. Two actual radioactive sludges were treated according to the new flowsheets. The peak hydrogen evolution rates were equivalent to 0.038 and 0.20 lb/hr (DWPF scale) respectively. Compared to the formic acid -- HAN hydrolysis flowsheets, these peak rates were reduced by a factor of 2.5 and 3.4 for Tank 15 and Tank 11 sludges, respectively

  5. Technical Exchange on Improved Design and Performance of High Level Waste Melters - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    SK Sundaram; ML Elliott; D Bickford

    1999-11-19

    SIA Radon is responsible for management of low- and intermediate-level radioactive waste (LILW) produced in Central Russia. In cooperation with Minatom organizations Radon carries out R and D programs on treatment of simulated high level waste (HLW) as well. Radon scientists deal with a study of materials for LILW, HLW, and Nuclear Power Plants (NPP) wastes immobilization, and development and testing of processes and technologies for waste treatment and disposal. Radon is mostly experienced in LILW vitrification. This experience can be carried over to HLW vitrification especially in field of melting systems. The melter chosen as a basic unit for the vitrification plant is a cold crucible. Later on Radon experience in LILW vitrification as well as our results on simulated HLW vitrification are briefly described.

  6. Cold crucible induction melter test for crystalline ceramic waste form fabrication: A feasibility assessment

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, Jake W., E-mail: jake.amoroso@srnl.doe.gov [Savannah River National Laboratory, Aiken, SC 29808 (United States); Marra, James; Dandeneau, Christopher S. [Savannah River National Laboratory, Aiken, SC 29808 (United States); Brinkman, Kyle; Xu, Yun [Clemson University, Clemson, SC 29634 (United States); Tang, Ming [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Maio, Vince [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Webb, Samuel M. [Stanford Synchrotron Radiation Lightsource, SLAC National Accelerator Laboratory, Menlo Park, CA 94086 (United States); Chiu, Wilson K.S. [University of Connecticut, Storrs, Connecticut 06269-3139 (United States)

    2017-04-01

    The first scaled proof-of-principle cold crucible induction melter (CCIM) test to process a multiphase ceramic waste form from a simulated combined (Cs/Sr, lanthanide and transition metal fission products) commercial used nuclear fuel waste stream was recently conducted in the United States. X-ray diffraction, 2-D X-ray absorption near edge structure (XANES), 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 fabricated CCIM material. Characterization analyses confirmed that a crystalline ceramic with a desirable phase assemblage was produced from a melt using a CCIM. Primary hollandite, pyrochlore/zirconolite, and perovskite phases were identified in addition to minor phases rich in Fe, Al, or Cs. The material produced in the CCIM was chemically homogeneous and displayed a uniform phase assemblage with acceptable aqueous chemical durability.

  7. Technical Exchange on Improved Design and Performance of High Level Waste Melters - Final Report

    International Nuclear Information System (INIS)

    Sundaram, S.K.; Elliott, M.L.; Bickford, D.

    1999-01-01

    SIA Radon is responsible for management of low- and intermediate-level radioactive waste (LILW) produced in Central Russia. In cooperation with Minatom organizations Radon carries out R and D programs on treatment of simulated high level waste (HLW) as well. Radon scientists deal with a study of materials for LILW, HLW, and Nuclear Power Plants (NPP) wastes immobilization, and development and testing of processes and technologies for waste treatment and disposal. Radon is mostly experienced in LILW vitrification. This experience can be carried over to HLW vitrification especially in field of melting systems. The melter chosen as a basic unit for the vitrification plant is a cold crucible. Later on Radon experience in LILW vitrification as well as our results on simulated HLW vitrification are briefly described

  8. Vitrification of HLLW Surrogate Solutions Containing Sulfate in a Direct-Induction Cold Crucible Melter

    International Nuclear Information System (INIS)

    Tronche, E.; Lacombe, J.; Ledoux, A.; Boen, R.; Ladirat, C.H.

    2009-01-01

    Efforts were made in the People's Republic of China to solidify legacy high level liquid waste (HLLW) by the Liquid-Fed Ceramic Melter process (LFCM) in the 1990's. This process was to be a continuous process with high throughput as in the French Marcoule Vitrification Plant (AVM) or the LFCM. In this context, the CEA (Commissariat a l'Energie Atomique is a French government-funded technological research organization) suggests the Cold Crucible Induction Melter (CCIM) technology that has been developed by the CEA since the 1980's to improve the performance of the vitrification process. In this context a series of vitrification tests has been carried out in a CCIM. CEA and AREVA have designed an integrated platform based on the CCIM technology on a sufficient scale to be used for demonstration programs of the one-step process. In 2003 a test was carried out at Marcoule in southern France on simulated HLLW with high sulfur content. In order to ensure the tests performed at Marcoule were consistent with the Chinese waste-forms, the glass frit was supplied by a Chinese Industry. The CCIM facility is described in detail, including process instrumentation. The test run is also described, including how the solution was directly fed on the surface of the molten glass. A maximum capacity was determined according to the applied process parameters including the high operating temperature. The electrical power supply characteristics are detailed and a glass mass balance is also presented covering more than seven hundred kilograms of glass produced in a sixty-hour test run. (authors)

  9. Computer Modeling Of High-Level Waste Glass Temperatures Within DWPF Canisters During Pouring And Cool Down

    International Nuclear Information System (INIS)

    Amoroso, J.

    2011-01-01

    This report describes the results of a computer simulation study to predict the temperature of the glass at any location inside a DWPF canister during pouring and subsequent cooling. These simulations are an integral part of a larger research focus aimed at developing methods to predict, evaluate, and ultimately suppress nepheline formation in HLW glasses. That larger research focus is centered on holistically understanding nepheline formation in HLW glass by exploring the fundamental thermal and chemical driving forces for nepheline crystallization with respect to realistic processing conditions. Through experimental work, the goal is to integrate nepheline crystallization potential in HLW glass with processing capability to ultimately optimize waste loading and throughput while maintaining an acceptable product with respect to durability. The results of this study indicated severe temperature gradients and prolonged temperature dwell times exist throughout different locations in the canister and that the time and temperatures that HLW glass is subjected to during processing is a function of pour rate. The simulations indicate that crystallization driving forces are not uniform throughout the glass volume in a DWPF (or DWPF-like) canister and illustrate the importance of considering overall kinetics (chemical and thermal driving forces) of nepheline formation when developing methods to predict and suppress its formation in HLW glasses. The intended path forward is to use the simulation data both as a driver for future experimental work and, as an investigative tool for evaluating the impact of experimental results. Simulation data will be used to develop laboratory experiments to more acutely evaluate nepheline formation in HLW glass by incorporating the simulated temperatures throughout the canister into the laboratory experiments. Concurrently, laboratory experiments will be performed to identify nepheline crystallization potential in HLW glass as a function of

  10. Effects of alumina sources (gibbsite, boehmite, and corundum) on melting behaviour of high-level radioactive waste melter feed

    Czech Academy of Sciences Publication Activity Database

    Lee, S.; Hrma, P.; Pokorný, R.; Kloužek, Jaroslav; VanderVeer, B.J.; Rodriguez, C.P.; Chun, J.; Schweiger, M. J.; Kruger, A.A.

    2017-01-01

    Roč. 2, č. 11 (2017), s. 603-608 ISSN 2059-8521 Institutional support: RVO:67985891 Keywords : foam * specific heat * porosity Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass OBOR OECD: Ceramics

  11. A technical basis to relax the dew point specification for the environment in the vapor space in DWPF canisters

    International Nuclear Information System (INIS)

    Louthan, M.R. Jr.

    1995-05-01

    This memorandum establishes the technical basis to conclude that relaxing, from 0 C to 20 C, the dew point specification for the atmosphere in the vapor space (free volume) of a DWPF canister will not provide an environment that will cause significant amounts of corrosion induced degradation of the canister wall. The conclusion is based on engineering analysis, experience and review of the corrosion literature. The basic assumptions underlying the conclusion are: (1) the canister was fabricated from Type 304L stainless steel; (2) the corrosion behavior of the canister material, including base metal, fusion zones and heat effected zones, is typified by literature data for, and industrial experience with, 300 series austenitic stainless steels; and (3) the glass-metal crevices created during the pouring operation will not alter the basic corrosion resistance of the steel although such crevices might serve as sites for the initiation of minor amounts of corrosion on the canister wall

  12. Verification Of The Defense Waste Processing Facility's (DWPF) Process Digestion Methods For The Sludge Batch 8 Qualification Sample

    International Nuclear Information System (INIS)

    Click, D. R.; Edwards, T. B.; Wiedenman, B. J.; Brown, L. W.

    2013-01-01

    This report contains the results and comparison of data generated from inductively coupled plasma atomic emission spectroscopy (ICP-AES) analysis of Aqua Regia (AR), Sodium Peroxide/Sodium Hydroxide Fusion Dissolution (PF) and Cold Chem (CC) method digestions and Cold Vapor Atomic Absorption analysis of Hg digestions from the DWPF Hg digestion method of Sludge Batch 8 (SB8) Sludge Receipt and Adjustment Tank (SRAT) Receipt and SB8 SRAT Product samples. The SB8 SRAT Receipt and SB8 SRAT Product samples were prepared in the SRNL Shielded Cells, and the SRAT Receipt material is representative of the sludge that constitutes the SB8 Batch or qualification composition. This is the sludge in Tank 51 that is to be transferred into Tank 40, which will contain the heel of Sludge Batch 7b (SB7b), to form the SB8 Blend composition

  13. Yield Stress Reduction of Radioactive Waste Slurries by Addition of Surfactants

    International Nuclear Information System (INIS)

    MICHAEL, STONE

    2005-01-01

    The Savannah River Site (SRS) and Hanford site are in the process of stabilizing millions of gallons of radioactive waste slurries remaining from production of nuclear materials for the Department of Energy (DOE). The Defense Waste Processing Facility (DWPF) at SRS is currently vitrifying the waste in borosilicate glass while the facilities at the Hanford site are in the design/construction phase. Both processes utilize slurry-fed joule heated melters to vitrify the waste slurries. The rheological properties of the waste slurries limit the total solids content that can be processed by the remote equipment during the pretreatment and melter feed processes. The use of a surface active agent, or surfactant, to increase the solids loading that can be fed to the melters would increase melt rate by reducing the heat load on the melter required to evaporate the water in the feed. The waste slurries are non-Newtonian fluids with rheological properties that were modeled using the Bingham Plastic mod el (this model is typically used by SRNL when studying the DWPF process1).The results illustrate that altering the surface chemistry of the particulates in the waste slurries can lead to a reduction in the yield stress. Dolapix CE64 is an effective surfactant over a wide range of pH values and was effective for all simulants tested. The effectiveness of the additive increased in DWPF simulants as the concentration of the additive was increased. No maxi main effectiveness was observed. Particle size measurements indicate that the additive acted as a flocculant in the DWPF samples and as a dispersant in the RPP samples

  14. Silicate Based Glass Formulations for Immobilization of U.S. Defense Wastes Using Cold Crucible Induction Melters

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Gary L.; Kim, Dong-Sang; Schweiger, Michael J.; Marra, James C.; Lang, Jesse B.; Crum, Jarrod V.; Crawford, Charles L.; Vienna, John D.

    2014-05-22

    The cold crucible induction melter (CCIM) is an alternative technology to the currently deployed liquid-fed, ceramic-lined, Joule-heated melter for immobilizing of U.S. tank waste generated from defense related reprocessing. In order to accurately evaluate the potential benefits of deploying a CCIM, glasses must be developed specifically for that melting technology. Related glass formulation efforts have been conducted since the 1990s including a recent study that is first documented in this report. The purpose of this report is to summarize the silicate base glass formulation efforts for CCIM testing of U.S. tank wastes. Summaries of phosphate based glass formulation and phosphate and silicate based CCIM demonstration tests are reported separately (Day and Ray 2013 and Marra 2013, respectively). Combined these three reports summarize the current state of knowledge related to waste form development and process testing of CCIM technology for U.S. tank wastes.

  15. Computational Fluid Dynamics Modeling of Bubbling in a Viscous Fluid for Validation of Waste Glass Melter Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Abboud, Alexander William [Idaho National Laboratory; Guillen, Donna Post [Idaho National Laboratory

    2016-01-01

    At the Hanford site, radioactive waste stored in underground tanks is slated for vitrification for final disposal. A comprehensive knowledge of the glass batch melting process will be useful in optimizing the process, which could potentially reduce the cost and duration of this multi-billion dollar cleanup effort. We are developing a high-fidelity heat transfer model of a Joule-heated ceramic lined melter to improve the understanding of the complex, inter-related processes occurring with the melter. The glass conversion rates in the cold cap layer are dependent on promoting efficient heat transfer. In practice, heat transfer is augmented by inserting air bubblers into the molten glass. However, the computational simulations must be validated to provide confidence in the solutions. As part of a larger validation procedure, it is beneficial to split the physics of the melter into smaller systems to validate individually. The substitution of molten glass for a simulant liquid with similar density and viscosity at room temperature provides a way to study mixing through bubbling as an isolated effect without considering the heat transfer dynamics. The simulation results are compared to experimental data obtained by the Vitreous State Laboratory at the Catholic University of America using bubblers placed within a large acrylic tank that is similar in scale to a pilot glass waste melter. Comparisons are made for surface area of the rising air bubbles between experiments and CFD simulations for a variety of air flow rates and bubble injection depths. Also, computed bubble rise velocity is compared to a well-accepted expression for bubble terminal velocity.

  16. Evaporation Of Hanford Waste Treatment Plant Direct Feed Low Activity Waste Effluent Management Facility Core Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Mcclane, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-01

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate, LMOGC) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation, and recycled to the LAW vitrification facility. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream is to evaporate it in a new evaporator, in the Effluent Management Facility (EMF), and then return it to the LAW melter. It is important to understand the composition of the effluents from the melter and new evaporator, so that the disposition of these streams can be accurately planned and accommodated. Furthermore, alternate disposition of the LMOGC stream would eliminate recycling of problematic components, and would reduce the need for closely integrated operation of the LAW melter and the Pretreatment Facilities. Long-term implementation of this option after WTP start-up would decrease the LAW vitrification mission duration and quantity of glass waste, amongst the other operational complexities such a recycle stream presents. In order to accurately plan for the disposition path, it is key to experimentally determine the fate of contaminants. To do this, testing is needed to accurately account for the buffering chemistry of the components, determine the achievable evaporation end point, identify insoluble solids that form, and determine the distribution of key regulatory-impacting constituents. The LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures, have limited solubility in the glass waste form, and represent a materials corrosion concern, such as halides and sulfate. Because this stream will recycle within WTP, these components will accumulate in the Melter Condensate

  17. IMPACTS OF SMALL COLUMN ION EXCHANGE STREAMS ON DWPF GLASS FORMULATION: KT01, KT02, KT03, AND KT04-SERIES GLASS COMPOSITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K.; Edwards, T.

    2010-11-01

    Four series of glass compositions were selected, fabricated, and characterized as part of a study to determine the impacts of the addition of Crystalline Silicotitanate (CST) and Monosodium Titanate (MST) from the Small Column Ion Exchange (SCIX) process on the Defense Waste Processing Facility (DWPF) glass waste form and the applicability of the DWPF process control models. The KT01 and KT02-series of glasses were chosen to allow for the identification of the influence of the concentrations of major components of the glass on the retention of TiO{sub 2}. The KT03 series of glasses was chosen to allow for the identification of these influences when higher Nb{sub 2}O{sub 5} and ZrO{sub 2} concentrations are included along with TiO2. The KT04 series of glasses was chosen to investigate the properties and performance of glasses based on the best available projections of actual compositions to be processed at the DWPF (i.e., future sludge batches including the SCIX streams).

  18. The Product Composition Control System at Savannah River: The statistical process control algorithm

    International Nuclear Information System (INIS)

    Brown, K.G.

    1993-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) in Aiken, South Carolina, will be used to immobilize the approximately 130 million liters of high-level nuclear waste currently stored at the site in 51 carbon steel tanks. Waste handling operations separate this waste into highly radioactive insoluble sludge and precipitate and less radioactive water soluble salts. (In a separate facility, the soluble salts are disposed of as low-level waste in a mixture of cement, slag, and flyash.) In DWPF, precipitate (PHA) is blended with insoluble sludge and ground glass tit to produce melter feed slurry which is continuously fed to the DWPF melter. The melter produces a molten borosilicate glass which is poured into stainless steel canisters for cooling and, ultimately, shipment to and storage in a geologic repository. The repository requires that the glass wasteform be resistant to leaching by underground water that might contact it. In addition, there are processing constraints on melt viscosity, liquidus temperature, and waste solubility

  19. Material interactions between system components and glass product melts in a ceramic melter

    International Nuclear Information System (INIS)

    Knitter, R.

    1989-07-01

    The interactions of the ceramic and metallic components of a ceramic melter for the vitrification of High Active Waste were investigated with simulated glass product melts in static crucible tests at 1000 0 C and 1150 0 C. Corrosion of the fusion-cast Al 2 O 3 -ZrO 2 -SiO 2 - and Al 2 O 3 -ZrO 2 -SiO 2 -Cr 2 O 3 -refractories (ER 1711 and ER 2161) is characterized by homogeneous chemical dissolution and diffusion through the glass matrix of the refractory. The resulting boundary compositions lead to characteristic modification and formation of phases, not only inside the refractory but also in the glass melt. The attack of the electrode material, a Ni-Cr-Fe-alloy Inconel 690, by the glass melt takes place via grain boundaries and leads to the oxidation of Cr and growth of Cr 2 O 3 -crystals at the boundary layer. Noble metals, added to the glass melt can form solid solutions with the alloy with varying compositions. (orig.) [de

  20. Connecting section and associated systems concept for the spray calciner/in-can melter process

    International Nuclear Information System (INIS)

    Petkus, L.L.; Gorton, P.S.; Blair, H.T.

    1981-06-01

    For a number of years, researchers at the Pacific Northwest Laboratory have been developing processes and equipment for converting high-level liquid wastes to solid forms. One of these processes is the Spray Calciner/In-Can Melter system. To immobilize high-level liquid wastes, this system must be operated remotely, and the calcine must be reliably conveyed from the calciner to the melting furnace. A concept for such a remote conveyance system was developed at the Pacific Northwest Laboratory, and equipment was tested under full-scale, nonradioactive conditions. This concept and the design of demonstration equipment are described, and the results of equipment operation during experimental runs of 7 d are presented. The design includes a connecting section and its associated systems - a canister sypport and alignment concept and a weight-monitoring system for the melting furnace. Overall, the runs demonstrated that the concept design is an acceptable method of connecting the two pieces of process equipment together. Although the connecting section has not been optimized in all areas of concern, it provides a first-generation design of a production-oriented system

  1. Impact Of Particle Agglomeration On Accumulation Rates In The Glass Discharge Riser Of HLW Melter

    International Nuclear Information System (INIS)

    Kruger, A. A.; Rodriguez, C. A.; Matyas, J.; Owen, A. T.; Jansik, D. P.; Lang, J. B.

    2012-01-01

    The major factor limiting waste loading in continuous high-level radioactive waste (HLW) melters is an accumulation of particles in the glass discharge riser during a frequent and periodic idling of more than 20 days. An excessive accumulation can produce robust layers a few centimeters thick, which may clog the riser, preventing molten glass from being poured into canisters. Since the accumulation rate is driven by the size of particles we investigated with x-ray microtomography, scanning electron microscopy, and image analysis the impact of spinel forming components, noble metals, and alumina on the size, concentration, and spatial distribution of particles, and on the accumulation rate. Increased concentrations of Fe and Ni in the baseline glass resulted in the formation of large agglomerates that grew over the time to an average size of ∼185+-155 μm, and produced >3 mm thick layer after 120 h at 850 deg C. The noble metals decreased the particle size, and therefore significantly slowed down the accumulation rate. Addition of alumina resulted in the formation of a network of spinel dendrites which prevented accumulation of particles into compact layers

  2. Preliminary experiments to simulate glass/electrode interactions within a Joule Ceramic Melter

    International Nuclear Information System (INIS)

    Dalton, J.T.; Paige, E.L.; Sutcliffe, P.W.

    1986-01-01

    Preliminary isothermal corrosion tests have been made on Inconel 690 coupon samples immersed in Harvest II M9 glass with and without excess additions of Li 2 O (1.5%) and RuO 2 (20%) together with TeO 2 (2%) at 1200 0 C for periods up to 100 hours. Inconel 690 corrosion and the products and ruthenium redox conditions within the glass approximate to those observed in the 1/3rd scale Joule Ceramic Melter operations. Corrosion takes place by an oxidation mechanism to form a chromium-rich surface oxide, and dissolution of this surface oxide by the surrounding glass. Additions of excess Li 2 O increase the corrosion rate of Inconel 690, whereas RuO 2 + TeO 2 are neutral. The latter however have a marked effect in lowering the room temperature resistivity by at least 5 orders of magnitude even though relatively small fraction of the RuO 2 precipitates were reduced to ruthenium metal. (author)

  3. The integrated melter off-gas treatment systems at the West Valley Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Vance, R.F. [West Valley Nuclear Services Co., Inc., NY (United States)

    1995-02-01

    The West Valley Demonstration Project was established by Public Law 96-368, the {open_quotes}West Valley Demonstration Project Act, {close_quotes} on October 1, l980. Under this act, Congress directed the Department of Energy to carry out a high level radioactive waste management demonstration project at the Western New York Nuclear Service Center in West Valley, New York. The purpose of this project is to demonstrate solidification techniques which can be used for preparing high level radioactive waste for disposal. In addition to developing this technology, the West Valley Demonstration Project Act directs the Department of Energy to: (1) develop containers suitable for permanent disposal of the high level waste; (2) transport the solidified high level waste to a Federal repository; (3) dispose of low level and transuranic waste produced under the project; and (4) decontaminate and decommission the facilities and materials associated with project activities and the storage tanks originally used to store the liquid high level radioactive waste. The process of vitrification will be used to solidify the high level radioactive liquid wastes into borosilicate glass. This report describes the functions, the controlling design criteria, and the resulting design of the melter off-gas treatment systems which are used in the vitrification process.

  4. Demonstration of the Defense Waste Processing Facility vitrification process for Tank 42 radioactive sludge -- Glass preparation and characterization

    International Nuclear Information System (INIS)

    Bibler, N.E.; Fellinger, T.L.; Marshall, K.M.; Crawford, C.L.; Cozzi, A.D.; Edwards, T.B.

    1999-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) is currently processing and immobilizing the radioactive high level waste sludge at SRS into a durable borosilicate glass for final geological disposal. The DWPF has recently finished processing the first radioactive sludge batch, and is ready for the second batch of radioactive sludge. The second batch is primarily sludge from Tank 42. Before processing this batch in the DWPF, the DWPF process flowsheet has to be demonstrated with a sample of Tank 42 sludge to ensure that an acceptable melter feed and glass can be made. This demonstration was recently completed in the Shielded Cells Facility at SRS. An earlier paper in these proceedings described the sludge composition and processes necessary for producing an acceptable melter fee. This paper describes the preparation and characterization of the glass from that demonstration. Results substantiate that Tank 42 sludge after mixing with the proper amount of glass forming frit (Frit 200) can be processed to make an acceptable glass

  5. IMPACTS OF ANTIFOAM ADDITIONS AND ARGON BUBBLING ON DEFENSE WASTE PROCESSING FACILITY REDUCTION/OXIDATION

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C.; Johnson, F.

    2012-06-05

    During melting of HLW glass, the REDOX of the melt pool cannot be measured. Therefore, the Fe{sup +2}/{Sigma}Fe ratio in the glass poured from the melter must be related to melter feed organic and oxidant concentrations to ensure production of a high quality glass without impacting production rate (e.g., foaming) or melter life (e.g., metal formation and accumulation). A production facility such as the Defense Waste Processing Facility (DWPF) cannot wait until the melt or waste glass has been made to assess its acceptability, since by then no further changes to the glass composition and acceptability are possible. therefore, the acceptability decision is made on the upstream process, rather than on the downstream melt or glass product. That is, it is based on 'feed foward' statistical process control (SPC) rather than statistical quality control (SQC). In SPC, the feed composition to the melter is controlled prior to vitrification. Use of the DWPF REDOX model has controlled the balanjce of feed reductants and oxidants in the Sludge Receipt and Adjustment Tank (SRAT). Once the alkali/alkaline earth salts (both reduced and oxidized) are formed during reflux in the SRAT, the REDOX can only change if (1) additional reductants or oxidants are added to the SRAT, the Slurry Mix Evaporator (SME), or the Melter Feed Tank (MFT) or (2) if the melt pool is bubble dwith an oxidizing gas or sparging gas that imposes a different REDOX target than the chemical balance set during reflux in the SRAT.

  6. Recycle Waste Collection Tank (RWCT) simulant testing in the PVTD feed preparation system

    International Nuclear Information System (INIS)

    Abrigo, G.P.; Daume, J.T.; Halstead, S.D.; Myers, R.L.; Beckette, M.R.; Freeman, C.J.; Hatchell, B.K.

    1996-03-01

    (This is part of the radwaste vitrification program at Hanford.) RWCT was to routinely receive final canister decontamination sand blast frit and rinse water, Decontamination Waste Treatment Tank bottoms, and melter off-gas Submerged Bed Scrubber filter cake. In order to address the design needs of the RWCT system to meet performance levels, the PNL Vitrification Technology (PVTD) program used the Feed Preparation Test System (FPTS) to evaluate its equipment and performance for a simulant of RWCT slurry. (FPTS is an adaptation of the Defense Waste Processing Facility feed preparation system and represents the initially proposed Hanford Waste Vitrification Plant feed preparation system designed by Fluor-Daniel, Inc.) The following were determined: mixing performance, pump priming, pump performance, simulant flow characterization, evaporator and condenser performance, and ammonia dispersion. The RWCT test had two runs, one with and one without tank baffles

  7. 2013 CEF RUN - PHASE 1 DATA ANALYSIS AND MODEL VALIDATION

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A.

    2014-05-08

    Phase 1 of the 2013 Cold cap Evaluation Furnace (CEF) test was completed on June 3, 2013 after a 5-day round-the-clock feeding and pouring operation. The main goal of the test was to characterize the CEF off-gas produced from a nitric-formic acid flowsheet feed and confirm whether the CEF platform is capable of producing scalable off-gas data necessary for the revision of the DWPF melter off-gas flammability model; the revised model will be used to define new safety controls on the key operating parameters for the nitric-glycolic acid flowsheet feeds including total organic carbon (TOC). Whether the CEF off-gas data were scalable for the purpose of predicting the potential flammability of the DWPF melter exhaust was determined by comparing the predicted H{sub 2} and CO concentrations using the current DWPF melter off-gas flammability model to those measured during Phase 1; data were deemed scalable if the calculated fractional conversions of TOC-to-H{sub 2} and TOC-to-CO at varying melter vapor space temperatures were found to trend and further bound the respective measured data with some margin of safety. Being scalable thus means that for a given feed chemistry the instantaneous flow rates of H{sub 2} and CO in the DWPF melter exhaust can be estimated with some degree of conservatism by multiplying those of the respective gases from a pilot-scale melter by the feed rate ratio. This report documents the results of the Phase 1 data analysis and the necessary calculations performed to determine the scalability of the CEF off-gas data. A total of six steady state runs were made during Phase 1 under non-bubbled conditions by varying the CEF vapor space temperature from near 700 to below 300°C, as measured in a thermowell (T{sub tw}). At each steady state temperature, the off-gas composition was monitored continuously for two hours using MS, GC, and FTIR in order to track mainly H{sub 2}, CO, CO{sub 2}, NO{sub x}, and organic gases such as CH{sub 4}. The standard

  8. Preparation and evaporation of Hanford Waste treatment plant direct feed low activity waste effluent management facility simulant

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Howe, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-09-07

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate, LMOGC) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation, and recycled to the LAW vitrification facility. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream involves concentrating the condensate in a new evaporator at the Effluent Management Facility (EMF) and returning it to the LAW melter. The LMOGC stream will contain components, e.g. halides and sulfates, that are volatile at melter temperatures, have limited solubility in glass waste forms, and present a material corrosion concern. Because this stream will recycle within WTP, these components are expected to accumulate in the LMOGC stream, exacerbating their impact on the number of LAW glass containers that must be produced. Diverting the stream reduces the halides and sulfates in the glass and is a key objective of this program. In order to determine the disposition path, it is key to experimentally determine the fate of contaminants. To do this, testing is needed to account for the buffering chemistry of the components, determine the achievable evaporation end point, identify insoluble solids that form, determine the formation and distribution of key regulatoryimpacting constituents, and generate an aqueous stream that can be used in testing of the subsequent immobilization step. This overall program examines the potential treatment and immobilization of the LMOGC stream to enable alternative disposal. The objective of this task was to (1) prepare a simulant of the LAW Melter Off-gas Condensate expected during DFLAW operations, (2) demonstrate evaporation in order to predict the final composition of the effluents from the EMF

  9. Evaluation of alternative chemical additives for high-level waste vitrification feed preparation processing

    International Nuclear Information System (INIS)

    Seymour, R.G.

    1995-01-01

    During the development of the feed processing flowsheet for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), research had shown that use of formic acid (HCOOH) could accomplish several processing objectives with one chemical addition. These objectives included the decomposition of tetraphenylborate, chemical reduction of mercury, production of acceptable rheological properties in the feed slurry, and controlling the oxidation state of the glass melt pool. However, the DEPF research had not shown that some vitrification slurry feeds had a tendency to evolve hydrogen (H 2 ) and ammonia (NH 3 ) as the result of catalytic decomposition of CHOOH with noble metals (rhodium, ruthenium, palladium) in the feed. Testing conducted at Pacific Northwest Laboratory and later at the Savannah River Technical Center showed that the H 2 and NH 3 could evolve at appreciable rates and quantities. The explosive nature of H 2 and NH 3 (as ammonium nitrate) warranted significant mitigation control and redesign of both facilities. At the time the explosive gas evolution was discovered, the DWPF was already under construction and an immediate hardware fix in tandem with flowsheet changes was necessary. However, the Hanford Waste Vitrification Plant (HWVP) was in the design phase and could afford to take time to investigate flowsheet manipulations that could solve the problem, rather than a hardware fix. Thus, the HWVP began to investigate alternatives to using HCOOH in the vitrification process. This document describes the selection, evaluation criteria, and strategy used to evaluate the performance of the alternative chemical additives to CHOOH. The status of the evaluation is also discussed

  10. Late Wash/Nitric Acid flowsheet hydrogen generation bases for simulation of a deflagration/detonation in the DWPF CPC

    International Nuclear Information System (INIS)

    Ritter, J.A.

    1993-01-01

    Hydrogen generation data obtained from IDMS runs PX4 and PX5 will be used to determine a bases for a deflagration/detonation simulation in the DWPF CPC. This simulation is necessary due to the new chemistry associated with the Late Wash/ Nitric Acid flowsheet and process modifications associated with the presence of H 2 in the offgas. The simulation will be performed by Professor Van Brunt from the University of South Carolina. The scenario which leads up to the deflagration/detonation simulation will be chosen such that the following conditions apply. The SRAT is filled to its maximum operating level with 9,600 gal of sludge, which corresponds to the minimum vapor space above the sludge. The SRAT is at the boiling point, producing H 2 at a very low rate (about 10 % of the peak) and 15 scfm of air inleakage is entering the SRAT. Then, the H 2 generation rate will be allowed to increase exponentially (catalyst activation) until it readies the peak H 2 generation rate of the IDMS run, after which the H 2 generation rate will be allowed to decay exponentially (catalyst deactivation) until the total amount of H2 produced is between 85 and 100% of that produced during the IDMS run

  11. REMOTE IN-CELL SAMPLING IMPROVEMENTS PROGRAM AT THESAVANNAH RIVER SITE (SRS) DEFENSE WASTE PROCESSING FACILITY (DWPF)

    International Nuclear Information System (INIS)

    Marzolf, A

    2007-01-01

    Remote Systems Engineering (RSE) of the Savannah River National Lab (SRNL) in combination with the Defense Waste Processing Facility(DWPF) Engineering and Operations has evaluated the existing equipment and processes used in the facility sample cells for 'pulling' samples from the radioactive waste stream and performing equipment in-cell repairs/replacements. RSE has designed and tested equipment for improving remote in-cell sampling evolutions and reducing the time required for in-cell maintenance of existing equipment. The equipment within the present process tank sampling system has been in constant use since the facility start-up over 17 years ago. At present, the method for taking samples within the sample cells produces excessive maintenance and downtime due to frequent failures relative to the sampling station equipment and manipulator. Location and orientation of many sampling stations within the sample cells is not conducive to manipulator operation. The overextension of manipulators required to perform many in-cell operations is a major cause of manipulator failures. To improve sampling operations and reduce downtime due to equipment maintenance, a Portable Sampling Station (PSS), wireless in-cell cameras, and new commercially available sampling technology has been designed, developed and/or adapted and tested. The uniqueness of the design(s), the results of the scoping tests, and the benefits relative to in-cell operation and reduction of waste are presented

  12. Environmental Assessment for the Operation of the Glass Melter Thermal Treatment Unit at the US Department of Energy's Mound Plant, Miamisburg, Ohio

    International Nuclear Information System (INIS)

    1995-06-01

    The glass melter would thermally treat mixed waste (hazardous waste contaminated with radioactive constituents largely tritium, Pu-238, and/or Th-230) that was generated at the Mound Plant and is now in storage, by stabilizing the waste in glass blocks. Depending on the radiation level of the waste, the glass melter may operate for 1 to 6 years. Two onsite alternatives and seven offsite alternatives were considered. This environmental assessment indicates that the proposed action does not constitute a major Federal action significantly affecting the human environment according to NEPA, and therefore the finding of no significant impact is made, obviating the need for an environmental impact statement

  13. Environmental Assessment for the Operation of the Glass Melter Thermal Treatment Unit at the US Department of Energy`s Mound Plant, Miamisburg, Ohio

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    The glass melter would thermally treat mixed waste (hazardous waste contaminated with radioactive constituents largely tritium, Pu-238, and/or Th-230) that was generated at the Mound Plant and is now in storage, by stabilizing the waste in glass blocks. Depending on the radiation level of the waste, the glass melter may operate for 1 to 6 years. Two onsite alternatives and seven offsite alternatives were considered. This environmental assessment indicates that the proposed action does not constitute a major Federal action significantly affecting the human environment according to NEPA, and therefore the finding of no significant impact is made, obviating the need for an environmental impact statement.

  14. Laboratory optimization tests of technetium decontamination of Hanford Waste Treatment Plant low activity waste melter off-gas condensate simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, Kathryn M.L. [Savannah River Site (SRS), Aiken, SC (United States); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States)

    2015-11-01

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable simplified operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste.

  15. Volatilization of heavy metals and radionuclides from soil heated in an induction ''cold'' crucible melter

    International Nuclear Information System (INIS)

    Aloy, A.S.; Belov, V.Z.; Trofimenko, A.S.; Dmitriev, S.A.; Stefanovsky, S.V.; Gombert, D.; Knecht, D.A.

    1997-01-01

    The behavior of heavy metals and radionuclides during high-temperature treatment is very important for the design and operational capabilities of the off-gas treatment system, as well as for a better understanding of the nature and forms of the secondary waste. In Russia, a process for high-temperature melting in an induction heated cold crucible system is being studied for vitrification of Low Level Waste (LLW) flyash and SYNROC production with simulated high level waste (HLW). This work was done as part of a Department of Energy (DOE) funded research project for thermal treatment of mixed low level waste (LLW). Soil spiked with heavy metals (Cd, Pb) and radionuclides (Cs-137, U-239, Pu-239) was used as a waste surrogate. The soil was melted in an experimental lab-scale system that consisted of a high-frequency generator (1.76 MHz, 60 kW), a cold crucible melter (300 mm high and 90 mm in diameter), a shield box, and an off-gas system. The process temperature was 1,350--1,400 C. Graphite and silicon carbide were used as sacrificial conductive materials to start heating and initial melting of the soil batch. The off-gas system was designed in such a manner that after each experiment, it can be disconnected to collect and analyze all deposits to determine the mass balance. The off-gases were also sampled during an experiment to analyze for hydrogen, NO x , carbon dioxide, carbon monoxide and chlorine formation. This paper describes distribution and mass balance of metals and radionuclides in various parts of the off-gas system. The leach rate of the solidified blocks identified by the PCT method is also reported

  16. PLUTONIUM SOLUBILITY IN HIGH-LEVEL WASTE ALKALI BOROSILICATE GLASS

    Energy Technology Data Exchange (ETDEWEB)

    Marra, J.; Crawford, C.; Fox, K.; Bibler, N.

    2011-01-04

    The solubility of plutonium in a Sludge Batch 6 (SB6) reference glass and the effect of incorporation of Pu in the glass on specific glass properties were evaluated. A Pu loading of 1 wt % in glass was studied. Prior to actual plutonium glass testing, surrogate testing (using Hf as a surrogate for Pu) was conducted to evaluate the homogeneity of significant quantities of Hf (Pu) in the glass, determine the most appropriate methods to evaluate homogeneity for Pu glass testing, and to evaluate the impact of Hf loading in the glass on select glass properties. Surrogate testing was conducted using Hf to represent between 0 and 1 wt % Pu in glass on an equivalent molar basis. A Pu loading of 1 wt % in glass translated to {approx}18 kg Pu per Defense Waste Processing Facility (DWPF) canister, or about 10X the current allowed limit per the Waste Acceptance Product Specifications (2500 g/m{sup 3} of glass or about 1700 g/canister) and about 30X the current allowable concentration based on the fissile material concentration limit referenced in the Yucca Mountain Project License Application (897 g/m{sup 3}3 of glass or about 600 g Pu/canister). Based on historical process throughput data, this level was considered to represent a reasonable upper bound for Pu loading based on the ability to provide Pu containing feed to the DWPF. The task elements included evaluating the distribution of Pu in the glass (e.g. homogeneity), evaluating crystallization within the glass, evaluating select glass properties (with surrogates), and evaluating durability using the Product Consistency Test -- Me