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

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

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

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.

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

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

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

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

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

Glass consistency and glass performance

International Nuclear Information System (INIS)

Plodinec, M.J.; Ramsey, W.G.

1994-01-01

Glass produced by the Defense Waste Processing Facility (DWPF) will have to consistently be more durable than a benchmark glass (evaluated using a short-term leach test), with high confidence. The DWPF has developed a Glass Product Control Program to comply with this specification. However, it is not clear what relevance product consistency has on long-term glass performance. In this report, the authors show that DWPF glass, produced in compliance with this specification, can be expected to effectively limit the release of soluble radionuclides to natural environments. However, the release of insoluble radionuclides to the environment will be limited by their solubility, and not glass durability

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

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

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.

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

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

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

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

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.

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

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

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

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

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

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

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

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

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.

SUMMARY OF FY11 SULFATE RETENTION STUDIES FOR DEFENSE WASTE PROCESSING FACILITY GLASS

Energy Technology Data Exchange (ETDEWEB)

Fox, K.; Edwards, T.

2012-05-08

This report describes the results of studies related to the incorporation of sulfate in high level waste (HLW) borosilicate glass produced at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). A group of simulated HLW glasses produced for earlier sulfate retention studies was selected for full chemical composition measurements to determine whether there is any clear link between composition and sulfate retention over the compositional region evaluated. In addition, the viscosity of several glasses was measured to support future efforts in modeling sulfate solubility as a function of predicted viscosity. The intent of these studies was to develop a better understanding of sulfate retention in borosilicate HLW glass to allow for higher loadings of sulfate containing waste. Based on the results of these and other studies, the ability to improve sulfate solubility in DWPF borosilicate glasses lies in reducing the connectivity of the glass network structure. This can be achieved, as an example, by increasing the concentration of alkali species in the glass. However, this must be balanced with other effects of reduced network connectivity, such as reduced viscosity, potentially lower chemical durability, and in the case of higher sodium and aluminum concentrations, the propensity for nepheline crystallization. Future DWPF processing is likely to target higher waste loadings and higher sludge sodium concentrations, meaning that alkali concentrations in the glass will already be relatively high. It is therefore unlikely that there will be the ability to target significantly higher total alkali concentrations in the glass solely to support increased sulfate solubility without the increased alkali concentration causing failure of other Product Composition Control System (PCCS) constraints, such as low viscosity and durability. No individual components were found to provide a significant improvement in sulfate retention (i.e., an increase of the magnitude

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

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

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.

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

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.

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

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

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

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

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

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

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.

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.

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

Decontamination processes for waste glass canisters

International Nuclear Information System (INIS)

Rankin, W.N.

1982-01-01

A Defense Waste Processing Facility (DWPF) is currently being designed to convert Savannah River Plant liquid, high-level radioactive waste into a solid form, such as borosilicate glass. To prevent the spread of radioactivity, the outside of the canisters of waste glass must have very low levels of smearable radioactive contamination before they are removed from the DWPF. Several techniques were considered for canister decontamination: high-pressure water spray, electropolishing, chemical dissolution, and abrasive blasting. An abrasive blasting technique using a glass frit slurry has been selected for use in the DWPF. No additional equipment is needed to process waste generated from decontamination. Frit used as the abrasive will be mixed with the waste and fed to the glass melter. In contrast, chemical and electrochemical techniques require more space in the DWPF, and produce large amounts of contaminated by-products, which are difficult to immobilize by vitrification

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

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

Decontamination of Savannah River Plant waste glass canisters

International Nuclear Information System (INIS)

Rankin, W.N.

1982-01-01

A Defense Waste Processing Facility (DWPF) is currently being designed to convert Savannah River Plant (SRP) liquid, high-level radioactive waste into a solid form, such as borosilicate glass. The outside of the canisters of waste glass must have very low levels of smearable radioactive contamination before they are removed from the DWPF to prevent the spread of radioactivity. Several techniques were considered for canister decontamination: high-pressure water spray, electropolishing, chemical dissolution, and abrasive blasting. An abrasive blasting technique using a glass frit slurry has been selected for use in the DWPF. No additional equipment is needed to process waste generated from decontamination. Frit used as the abrasive will be mixed with the waste and fed to the glass melter. In contrast, chemical and electrochemical techniques require more space in the DWPF, and produce large amounts of contaminated byproducts which are difficult to immobilize by vitrification

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

Defining And Characterizing Sample Representativeness For DWPF Melter Feed Samples

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-29

Representative sampling is important throughout the Defense Waste Processing Facility (DWPF) process, and the demonstrated success of the DWPF process to achieve glass product quality over the past two decades is a direct result of the quality of information obtained from the process. The objective of this report was to present sampling methods that the Savannah River Site (SRS) used to qualify waste being dispositioned at the DWPF. The goal was to emphasize the methodology, not a list of outcomes from those studies. This methodology includes proven methods for taking representative samples, the use of controlled analytical methods, and data interpretation and reporting that considers the uncertainty of all error sources. Numerous sampling studies were conducted during the development of the DWPF process and still continue to be performed in order to evaluate options for process improvement. Study designs were based on use of statistical tools applicable to the determination of uncertainties associated with the data needs. Successful designs are apt to be repeated, so this report chose only to include prototypic case studies that typify the characteristics of frequently used designs. Case studies have been presented for studying in-tank homogeneity, evaluating the suitability of sampler systems, determining factors that affect mixing and sampling, comparing the final waste glass product chemical composition and durability to that of the glass pour stream sample and other samples from process vessels, and assessing the uniformity of the chemical composition in the waste glass product. Many of these studies efficiently addressed more than one of these areas of concern associated with demonstrating sample representativeness and provide examples of statistical tools in use for DWPF. The time when many of these designs were implemented was in an age when the sampling ideas of Pierre Gy were not as widespread as they are today. Nonetheless, the engineers and

Defining And Characterizing Sample Representativeness For DWPF Melter Feed Samples

International Nuclear Information System (INIS)

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

2013-01-01

Representative sampling is important throughout the Defense Waste Processing Facility (DWPF) process, and the demonstrated success of the DWPF process to achieve glass product quality over the past two decades is a direct result of the quality of information obtained from the process. The objective of this report was to present sampling methods that the Savannah River Site (SRS) used to qualify waste being dispositioned at the DWPF. The goal was to emphasize the methodology, not a list of outcomes from those studies. This methodology includes proven methods for taking representative samples, the use of controlled analytical methods, and data interpretation and reporting that considers the uncertainty of all error sources. Numerous sampling studies were conducted during the development of the DWPF process and still continue to be performed in order to evaluate options for process improvement. Study designs were based on use of statistical tools applicable to the determination of uncertainties associated with the data needs. Successful designs are apt to be repeated, so this report chose only to include prototypic case studies that typify the characteristics of frequently used designs. Case studies have been presented for studying in-tank homogeneity, evaluating the suitability of sampler systems, determining factors that affect mixing and sampling, comparing the final waste glass product chemical composition and durability to that of the glass pour stream sample and other samples from process vessels, and assessing the uniformity of the chemical composition in the waste glass product. Many of these studies efficiently addressed more than one of these areas of concern associated with demonstrating sample representativeness and provide examples of statistical tools in use for DWPF. The time when many of these designs were implemented was in an age when the sampling ideas of Pierre Gy were not as widespread as they are today. Nonetheless, the engineers and

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

Defense Waste Processing Facility radioactive operations -- Part 2, Glass making

International Nuclear Information System (INIS)

Carter, J.T.; Rueter, K.J.; Ray, J.W.; Hodoh, O.

1996-01-01

The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, SC is the nation's first and world's largest vitrification facility. Following a ten year construction period and nearly 3 year non-radioactive test program, the DWPF began radioactive operations in March, 1996. The results of the first 8 months of radioactive operations are presented. Topics include facility production from waste preparation batching to canister filling

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

Development of an ASTM standard glass durability test, the Product Consistency Test (PCT), for high level radioactive waste glass

International Nuclear Information System (INIS)

Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Ramsey, W.G.

1994-01-01

The nation's first, and the world's largest, facility to immobilize high-level nuclear waste in durable borosilicate glass has started operation at the Savannah River Site (SRS) in Aiken, South Carolina. The product specifications on the glass wasteform produced in the Defense Waste Processing Facility (DWPF) required extensive characterization of the glass product before actual production began and for continued characterization during production. To aid in this characterization, a glass durability (leach) test was needed that was easily reproducible, could be performed remotely on highly radioactive samples, and could yield results rapidly. Several standard leach tests were examined with a variety of test configurations. Using existing tests as a starting point, the DWPF Product Consistency Test (PCT was developed in which crushed glass samples are exposed to 90 ± 2 degree C deionized water for seven days. Based on extensive testing, including a seven-laboratory round robin and confirmatory testing with radioactive samples, the PCT is very reproducible, yields reliable results rapidly, and can be performed in shielded cell facilities with radioactive samples

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.

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.

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.

Effects of waste content of glass waste forms on Savannah River high-level waste disposal costs

International Nuclear Information System (INIS)

McDonell, W.R.; Jantzen, C.M.

1985-01-01

Effects of the waste content of glass waste forms of Savannah River high-level waste disposal costs are evaluated by their impact on the number of waste canisters produced. Changes in waste content affect onsite Defense Waste Processing Facility (DWPF) costs as well as offsite shipping and repository emplacement charges. A nominal 1% increase over the 28 wt % waste loading of DWPF glass would reduce disposal costs by about $50 million for Savannah River wastes generated to the year 2000. Waste form modifications under current study include adjustments of glass frit content to compensate for added salt decontamination residues and increased sludge loadings in the DWPF glass. Projected cost reductions demonstrate significant incentives for continued optimization of the glass waste loadings. 13 refs., 3 figs., 3 tabs

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

High level radioactive waste management facility design criteria

International Nuclear Information System (INIS)

Sheikh, N.A.; Salaymeh, S.R.

1993-01-01

This paper discusses the engineering systems for the structural design of the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). At the DWPF, high level radioactive liquids will be mixed with glass particles and heated in a melter. This molten glass will then be poured into stainless steel canisters where it will harden. This process will transform the high level waste into a more stable, manageable substance. This paper discuss the structural design requirements for this unique one of a kind facility. A special emphasis will be concentrated on the design criteria pertaining to earthquake, wind and tornado, and flooding

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

DWPF simulant CPC studies for SB8

Energy Technology Data Exchange (ETDEWEB)

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

2013-06-25

The Savannah River National Laboratory (SRNL) accepted a technical task request (TTR) from Waste Solidification Engineering to perform simulant tests to support the qualification of Sludge Batch 8 (SB8) and to develop the flowsheet for SB8 in the Defense Waste Processing Facility (DWPF). These efforts pertained to the DWPF Chemical Process Cell (CPC). Separate studies were conducted for frit development and glass properties (including REDOX). The SRNL CPC effort had two primary phases divided by the decision to drop Tank 12 from the SB8 constituents. This report focuses on the second phase with SB8 compositions that do not contain the Tank 12 piece. A separate report will document the initial phase of SB8 testing that included Tank 12. The second phase of SB8 studies consisted of two sets of CPC studies. The first study involved CPC testing of an SB8 simulant for Tank 51 to support the CPC demonstration of the washed Tank 51 qualification sample in the SRNL Shielded Cells facility. SB8-Tank 51 was a high iron-low aluminum waste with fairly high mercury and moderate noble metal concentrations. Tank 51 was ultimately washed to about 1.5 M sodium which is the highest wash endpoint since SB3-Tank 51. This study included three simulations of the DWPF Sludge Receipt and Adjustment Tank (SRAT) cycle and Slurry Mix Evaporator (SME) cycle with the sludge-only flowsheet at nominal DWPF processing conditions and three different acid stoichiometries. These runs produced a set of recommendations that were used to guide the successful SRNL qualification SRAT/SME demonstration with actual Tank 51 washed waste. The second study involved five SRAT/SME runs with SB8-Tank 40 simulant. Four of the runs were designed to define the acid requirements for sludge-only processing in DWPF with respect to nitrite destruction and hydrogen generation. The fifth run was an intermediate acid stoichiometry demonstration of the coupled flowsheet for SB8. These runs produced a set of processing

DWPF simulant CPC studies for SB8

International Nuclear Information System (INIS)

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

2013-01-01

The Savannah River National Laboratory (SRNL) accepted a technical task request (TTR) from Waste Solidification Engineering to perform simulant tests to support the qualification of Sludge Batch 8 (SB8) and to develop the flowsheet for SB8 in the Defense Waste Processing Facility (DWPF). These efforts pertained to the DWPF Chemical Process Cell (CPC). Separate studies were conducted for frit development and glass properties (including REDOX). The SRNL CPC effort had two primary phases divided by the decision to drop Tank 12 from the SB8 constituents. This report focuses on the second phase with SB8 compositions that do not contain the Tank 12 piece. A separate report will document the initial phase of SB8 testing that included Tank 12. The second phase of SB8 studies consisted of two sets of CPC studies. The first study involved CPC testing of an SB8 simulant for Tank 51 to support the CPC demonstration of the washed Tank 51 qualification sample in the SRNL Shielded Cells facility. SB8-Tank 51 was a high iron-low aluminum waste with fairly high mercury and moderate noble metal concentrations. Tank 51 was ultimately washed to about 1.5 M sodium which is the highest wash endpoint since SB3-Tank 51. This study included three simulations of the DWPF Sludge Receipt and Adjustment Tank (SRAT) cycle and Slurry Mix Evaporator (SME) cycle with the sludge-only flowsheet at nominal DWPF processing conditions and three different acid stoichiometries. These runs produced a set of recommendations that were used to guide the successful SRNL qualification SRAT/SME demonstration with actual Tank 51 washed waste. The second study involved five SRAT/SME runs with SB8-Tank 40 simulant. Four of the runs were designed to define the acid requirements for sludge-only processing in DWPF with respect to nitrite destruction and hydrogen generation. The fifth run was an intermediate acid stoichiometry demonstration of the coupled flowsheet for SB8. These runs produced a set of processing

Nuclear waste glass product consistency test (PCT), Version 5.0

International Nuclear Information System (INIS)

Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Ramsey, W.G.; Waters, B.J.

1992-06-01

Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced in the Defense Waste Processing Facility (DWPF), poured into stainless steel canisters, and eventually disposed of in a geologic repository. In order to comply with the Waste Acceptance Preliminary Specifications (WAPS), the durability of the glass needs to be measured during production to assure its long term stability and radionuclide release properties. A durability test, designated the Produce Consistency Test (PCT), was developed for DWPF glass in order to meet the WAPS requirements. The response of the PCT procedure was based on extensive testing with glasses of widely different compositions. The PCT was determined to be very reproducible, to yield reliable results rapidly, and to be easily performed in shielded cell facilities with radioactive samples. Version 5.0 of the PCT procedure is attached

Defense Waste Processing Facility -- Radioactive operations -- Part 3 -- Remote operations

International Nuclear Information System (INIS)

Barnes, W.M.; Kerley, W.D.; Hughes, P.D.

1997-01-01

The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, South Carolina is the nation's first and world's largest vitrification facility. Following a ten year construction period and nearly three years of non-radioactive testing, the DWPF began radioactive operations in March 1996. Radioactive glass is poured from the joule heated melter into the stainless steel canisters. The canisters are then temporarily sealed, decontaminated, resistance welded for final closure, and transported to an interim storage facility. All of these operations are conducted remotely with equipment specially designed for these processes. This paper reviews canister processing during the first nine months of radioactive operations at DWPF. The fundamental design consideration for DWPF remote canister processing and handling equipment are discussed as well as interim canister storage

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.

Efficient handling of high-level radioactive cell waste in a vitrification facility analytical laboratory

International Nuclear Information System (INIS)

Roberts, D.W.; Collins, K.J.

1998-01-01

The Savannah River Site''s (SRS) Defense Waste Processing Facility (DWPF) near Aiken, South Carolina, is the world''s largest and the United State''s first high level waste vitrification facility. For the past 1.5 years, DWPF has been vitrifying high level radioactive liquid waste left over from the Cold War. The vitrification process involves the stabilization of high level radioactive liquid waste into borosilicate glass. The glass is contained in stainless steel canisters. DWPF has filled more than 200 canisters 3.05 meters (10 feet) long and 0.61 meters (2 foot) diameter. Since operations began at DWPF in March of 1996, high level radioactive solid waste continues to be generated due to operating the facility''s analytical laboratory. The waste is referred to as cell waste and is routinely removed from the analytical laboratories. Through facility design, engineering controls, and administrative controls, DWPF has established efficient methods of handling the high level waste generated in its laboratory facility. These methods have resulted in the prevention of undue radiation exposure, wasted man-hours, expenses due to waste disposal, and the spread of contamination. This level of efficiency was not reached overnight, but it involved the collaboration of Radiological Control Operations and Laboratory personnel working together to devise methods that best benefited the facility. This paper discusses the methods that have been incorporated at DWPF for the handling of cell waste. The objective of this paper is to provide insight to good radiological and safety practices that were incorporated to handle high level radioactive waste in a laboratory setting

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

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

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

Preliminary technical data summary defense waste processing facility stage 2

International Nuclear Information System (INIS)

1980-12-01

This Preliminary Technical Data Summary presents the technical basis for design of Stage 2 of the Staged Defense Waste Processing Facility (DWPF). Process changes incorporated in the staged DWPF relative to the Alternative DWPF described in PTDS No. 3 (DPSTD-77-13-3) are the result of ongoing research and development and are aimed at reducing initial capital investment and developing a process to efficiently immobilize the radionuclides in Savannah River Plant (SRP) high-level liquid waste. The radionuclides in SRP waste are present in sludge that has settled to the bottom of waste storage tanks and in crystallized salt and salt solution (supernate). Stage 1 of the DWPF receives washed, aluminum dissolved sludge from the waste tank farms and immobilizes it in a borosilicate glass matrix. The supernate is retained in the waste tank farms until completion of Stage 2 of the DWPF at which time it is filtered and decontaminated by ion exchange in the Stage 2 facility. The decontaminated supernate is concentrated by evaporation and mixed with cement for burial. The radioactivity removed from the supernate is fixed in borosilicate glass along with the sludge. This document gives flowsheets, material and curie balances, material and curie balance bases, and other technical data for design of Stage 2 of the DWPF. Stage 1 technical data are presented in DPSTD-80-38

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

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

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

Environmental information document defense waste processing facility

International Nuclear Information System (INIS)

1981-07-01

This report documents the impact analysis of a proposed Defense Waste Processing Facility (DWPF) for immobilizing high-level waste currently being stored on an interim basis at the Savannah River Plant (SRP). The DWPF will process the waste into a form suitable for shipment to and disposal in a federal repository. The DWPF will convert the high-level waste into: a leach-resistant form containing above 99.9% of all the radioactivity, and a residue of slightly contaminated salt. The document describes the SRP site and environs, including population, land and water uses; surface and subsurface soils and waters; meteorology; and ecology. A conceptual integrated facility for concurrently producing glass waste and saltcrete is described, and the environmental effects of constructing and operating the facility are presented. Alternative sites and waste disposal options are addressed. Also environmental consultations and permits are discussed

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.

Preliminary technical data summary for the Defense Waste Processing Facility, Stage 1

International Nuclear Information System (INIS)

1980-09-01

This Preliminary Technical Data Summary presents the technical basis for design of Stage 1 of the Staged Defense Waste Processing Facility (DWPF), a process to efficiently immobilize the radionuclides in Savannah River Plant (SRP) high-level liquid waste. The radionuclides in SRP waste are present in sludge that has settled to the bottom of waste storage tanks and in crystallized salt and salt solution (supernate). Stage 1 of the DWPF receives washed, aluminum dissolved sludge from the waste tank farms and immobilizes it in a borosilicate glass matrix. The supernate is retained in the waste tank farms until completion of Stage 2 of the DWPF at which time it filtered and decontaminated by ion exchange in the Stage 2 facility. The decontaminated supernate is concentrated by evaporation and mixed with cement for burial. The radioactivity removed from the supernate is fixed in borosilicate glass along with the sludge. This document gives flowsheets, material, and curie balances, material and curie balance bases, and other technical data for design of the Stage 1 DWPF

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

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.

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.

Glass as a waste form for the immobilization of plutonium

International Nuclear Information System (INIS)

Bates, J.K.; Ellison, A.J.G.; Emery, J.W.; Hoh, J.C.

1995-01-01

Several alternatives for disposal of surplus plutonium are being considered. One method is incorporating Pu into glass and in this paper we discuss the development and corrosion behavior of an alkali-tin-silicate glass and update results in testing Pu doped Defense Waste Processing Facility (DWPF) reference glasses. The alkali-tin-silicate glass was engineered to accommodate a high Pu loading and to be durable under conditions likely to accelerate glass reaction. The glass dissolves about 7 wt% Pu together with the neutron absorber Gd, and under test conditions expected to accelerate the glass reaction with water, is resistant to corrosion. The Pu and the Gd are released from the glass at nearly the same rate in static corrosion tests in water, and are not segregated into surface alteration phases when the glass is reacted in water vapor. Similar results for the behavior of Pu and Gd are found for the DWPF reference glasses, although the long-term rate of reaction for the reference glasses is more rapid than for the alkali-tin-silicate glass

Analytical methods and laboratory facility for the Defense Waste Processing Facility

International Nuclear Information System (INIS)

Coleman, C.J.; Dewberry, R.A.; Lethco, A.J.; Denard, C.D.

1985-01-01

This paper describes the analytical methods, instruments, and laboratory that will support vitrification of defense waste. The Defense Waste Processing Facility (DWPF) is now being constructed at Savannah River Plant (SRP). Beginning in 1989, SRP high-level defense waste will be immobilized in borosilicate glass for disposal in a federal repository. The DWPF will contain an analytical laboratory for performing process control analyses. Additional analyses will be performed for process history and process diagnostics. The DWPF analytical facility will consist of a large shielded sampling cell, three shielded analytical cells, a laboratory for instrumental analysis and chemical separations, and a counting room. Special instrumentation is being designed for use in the analytical cells, including microwave drying/dissolution apparatus, and remote pipetting devices. The instrumentation laboratory will contain inductively coupled plasma, atomic absorption, Moessbauer spectrometers, a carbon analyzer, and ion chromatography equipment. Counting equipment will include intrinsic germanium detectors, scintillation counters, Phoswich alpha, beta, gamma detectors, and a low-energy photon detector

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

Nuclear waste glass product consistency test (PCT): Version 7.0. Revision 3

International Nuclear Information System (INIS)

Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Ramsey, W.G.

1994-06-01

Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced in the Defense Waste Processing Facility (DWPF), poured into stainless steel canisters, and eventually disposed of in a geologic repository. In order to comply with the Waste Acceptance Product Specifications (WAPS), the durability of the glass needs to be measured during production to assure its long term stability and radionuclide release properties. A durability test, designated the Product Consistency Test (PCT), was developed for DWPF glass in order to meet the WAPS requirements. The response of the PCT procedure was based on extensive testing with glasses of widely different compositions. The PCT was determined to be very reproducible, to yield reliable results rapidly, and to be easily performed in shielded cell facilities with radioactive samples. Version 7.0 of the PCT procedure is attached. This draft version has been submitted to ASTM for full committee (C26, Nuclear Fuel Cycle) ballot after being balloted successfully through subcommittee C26.13 on Repository Waste Package Materials Testing

Defense Waste Processing Facility, Savannah River Plant

International Nuclear Information System (INIS)

After 10 years of research, development, and testing, the US Department of Energy is building a new facility which will prepare high-level radioactive waste for permanent disposal. The Defense Waste Processing Facility, known as the DWPF, will be the first production-scale facility of its kind in the United States. In the DWPF, high-level waste produced by defense activities at the Savannah River Plant will be processed into a solid form, borosilicate glass, suitable for permanent off-site geologic disposal. With construction beginning in the fall of 1983, the DWPT is scheduled to be operational in 1989. By 2005, the DWPF will have immobilized the backlog of high-level waste which has been accumulating in storage tanks at the Savannah River Plant since 1954. Canisters of the immobilized waste will then be ready for permanent disposal deep under the ground, safely isolated from the environment

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

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

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

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 -- Method A (PCT-A). The behavior of Pu in the melter was evaluated using paper studies and corresponding analyses of DWPF melter pour samples.The results of the testing indicated that at 1 wt % Pu in the glass, the Pu was homogeneously distributed and did not result in any formation of plutonium-containing crystalline phases as long as the glass was prepared under 'well-mixed' conditions

Plutonium Solubility In High-Level Waste Alkali Borosilicate Glass

International Nuclear Information System (INIS)

Marra, J.; Crawford, C.; Fox, K.; Bibler, N.

2011-01-01

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 ∼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 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 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 -- Method A (PCT-A). The behavior of Pu in the melter was evaluated using paper studies and corresponding analyses of DWPF melter pour samples.The results of the testing indicated that at 1 wt % Pu in the glass, the Pu was homogeneously distributed and did not result in any formation of plutonium-containing crystalline phases as long as the glass was prepared under 'well-mixed' conditions. The incorporation of 1 wt

Design and construction of the defense waste processing facility project at the Savannah River Plant

International Nuclear Information System (INIS)

Baxter, R.G.

1986-01-01

The Du Pont Company is building for the Department of Energy a facility to vitrify high-level radioactive waste at the Savannah River Plant (SRP) near Aiken, South Carolina. The Defense Waste Processing Facility (DWPF) will solidify existing and future radioactive wastes by immobilizing the waste in Processing Facility (DWPF) will solidify existing and future radioactives wastes by immobilizing the waste in borosilicate glass contained in stainless steel canisters. The canisters will be sealed, decontaminated and stored, prior to emplacement in a federal repository. At the present time, engineering and design is 90% complete, construction is 25% complete, and radioactive processing in the $870 million facility is expected to begin by late 1989. This paper describes the SRP waste characteristics, the DWPF processing, building and equipment features, and construction progress of the facility

Comparative risk assessments for the production and interim storage of glass and ceramic waste forms: defense waste processing facility

International Nuclear Information System (INIS)

Huang, J.C.; Wright, W.V.

1982-04-01

The Defense Waste Processing Facility (DWPF) for immobilizing nuclear high level waste (HLW) is scheduled to be built at the Savannah River Plant (SRP). High level waste is produced when SRP reactor components are subjected to chemical separation operations. Two candidates for immobilizing this HLW are borosilicate glass and crystalline ceramic, either being contained in weld-sealed stainless steel canisters. A number of technical analyses are being conducted to support a selection between these two waste forms. The present document compares the risks associated with the manufacture and interim storage of these two forms in the DWPF. Process information used in the risk analysis was taken primarily from a DWPF processibility analysis. The DWPF environmental analysis provided much of the necessary environmental information. To perform the comparative risk assessments, consequences of the postulated accidents are calculated in terms of: (1) the maximum dose to an off-site individual; and (2) the dose to off-site population within 80 kilometers of the DWPF, both taken in terms of the 50-year inhalation dose commitment. The consequences are then multiplied by the estimated accident probabilities to obtain the risks. The analyses indicate that the maximum exposure risk to an individual resulting from the accidents postulated for both the production and interim storage of either waste form represents only an insignificant fraction of the natural background radiation of about 90 mrem per year per person in the local area. They also show that there is no disaster potential to the off-site population. Therefore, the risks from abnormal events in the production and the interim storage of the DWPF waste forms should not be considered as a dominant factor in the selection of the final waste form

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

SPEEDUP modeling of the defense waste processing facility at the SRS

International Nuclear Information System (INIS)

Smith, F.G. III.

1997-01-01

A computer model has been developed for the dynamic simulation of batch process operations within the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). The DWPF chemically treats high level waste materials from the site tank farm and vitrifies the resulting slurry into a borosilicate glass for permanent disposal. The DWPF consists of three major processing areas: Salt Processing Cell (SPC), Chemical Processing Cell (CPC) and the Melt Cell. A fully integrated model of these process units has been developed using the SPEEDUP trademark software from Aspen Technology. Except for glass production in the Melt Cell, all of the chemical operations within DWPF are batch processes. Since SPEEDUP is designed for dynamic modeling of continuous processes, considerable effort was required to device batch process algorithms. This effort was successful and the model is able to simulate batch operations and the dynamic behavior of the process. The model also includes an optimization calculation that maximizes the waste content in the final glass product. In this paper, we will describe the process model in some detail and present preliminary results from a few simulation studies

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.

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

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.

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

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

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)

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.

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

Supplemental environmental impact statement - defense waste processing facility

International Nuclear Information System (INIS)

1994-11-01

This document supplements the Final Environmental Impact Statement (EIS) DOE Issued in 1982 (DOE/EIS-0082) to construct and operate the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), a major DOE installation in southwestern South Carolina. That EIS supported the decision to construct and operate the DWPF to immobilize high-level waste generated as a result of nuclear materials processing at SRS. The DWPF would use a vitrification process to incorporate the radioactive waste into borosilicate glass and seal it in stainless steel canisters for eventual disposal at a permanent geologic repository. The DWPF is now mostly constructed and nearly ready for full operation. However, DOE has made design changes to the DWPF since the 1982 EIS to improve efficiency and safety of the facility. Each of these modifications was subjected to appropriate NEPA review. The purpose of this Supplemental EIS is to assist DOE in deciding whether and how to proceed with operation of the DWPF as modified since 1982 while ensuring appropriate consideration of potential environmental effects. In this document, DOE assesses the potential environmental impacts of completing and operating the DWPF in light of these design changes, examines the impact of alternatives, and identifies potential actions to be taken to reduce adverse impacts. Evaluations of impacts on water quality, air quality, ecological systems, land use, geologic resources, cultural resources, socioeconomics, and health and safety of onsite workers and the public are included in the assessment

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

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)

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

Waste glass melting stages

International Nuclear Information System (INIS)

Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

1993-04-01

Three different simulated nuclear waste glass feeds, consisting of dried waste and glass frit, were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600 degrees C--1000 degrees C. Simulated melter feeds from the Hanford Waste Vitrification Plant (HWVP), the Defense Waste Processing Facility (DWPF), and Kernforschungszentrum Karlsruhe (KfK) in Germany were used. The samples were thin-sectioned and examined by optical microscopy to investigate the stages of the conversion from feed to glass. Various phenomena were seen, such as frit softening, bubble formation, foaming, bubble motion and removal, convective mixing, and homogenization. Behavior of different feeds was similar, although the degree of gas generation and melt homogenization varied

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.

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

The remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

International Nuclear Information System (INIS)

Callan, J.E.

1986-01-01

The Defense Waste Processing Facility (DWPF) is a complete production area being constructed at the Savannah River Plant for the immobilization of nuclear waste in glass. The remote handling of canisters filled with nuclear waste in glass is an essential part of the process of the DWPF at the Savannah River Plant. The canisters are filled with nuclear waste containing up to 235,000 curies of radioactivity. Handling and movement of these canisters must be accomplished remotely since they radiate up to 5000 R/h. Within the Vitrification Building during filling, cleaning, and sealing, canisters are moved using standard cranes and trolleys and a specially designed grapple. During transportation to the Glass Waste Storage Building, a one-of-a-kind, specially designed Shielded Canister Transporter (SCT) is used. 8 figs

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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. The reported corrosion rates and degradation characteristics have shown the following for the materials of construction: For C276 alloy, the primary material of construction for the CPC vessels, corrosion rates of either 2 or 20 mpy were reported up to a temperature of 93 deg C; For the austenitic stainless steels, 304L and 316L, variable rates were reported over a range of temperatures, varying from 2 mpy up to 200 mpy (at 100 deg C); For 690, G30, Allcorr, Ultimet and Stellite alloys no data were available; and, For relevant polymers where data are available, the data suggests that exposure to glycolic acid is not detrimental. The literature data had limited application to the DWPF process since only the storage and feed vessels, pumps and piping used to handle the glycolic acid are directly covered by the available data. These components are either 304L or 316L alloys for which the literature data is inconsistent (See Bullet 2 above). Corrosion rates in pure glycolic acid solutions also are not representative of the DWPF process streams. This stream is complex and contains aggressive species, i.e. chlorides, sulfates, mercury, as well as antifoaming agents which cumulatively have an unknown effect on the corrosion rates of the materials of construction. Therefore, testing is recommended to investigate any synergistic effects of the aggressive

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

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.

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.

Materials evaluation programs at the Defense Waste Processing Facility

International Nuclear Information System (INIS)

Gee, J.T.; Iverson, D.C.; Bickford, D.F.

1992-01-01

The Savannah River Site (SRS) has been operating a nuclear fuel cycle since the 1950s to produce nuclear materials in support of the national defense effort. About 83 million gallons of high-level waste produced since operations began has been consolidated by evaporation into 33 million gallons at the waste tank farm. The Department of Energy authorized the construction of the Defense Waste Processing Facility (DWPF), the function of which is to immobilize the waste as a durable borosilicate glass contained in stainless steel canisters prior to the placement of the canisters in a federal repository. The DWPF is now mechanically complete and is undergoing commissioning and run-in activities. A brief description of the DWPF process is provided

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

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

Criticality assessment of initial operations at the Defense Waste Processing Facility

International Nuclear Information System (INIS)

Ha, B.C.; Williamson, T.G.

1993-01-01

At the Savannah River Site (SRS), high level radioactive wastes will be immobilized into borosilicate glass for long term storage and eventual disposal. Since the waste feed streams contain uranium and plutonium, the Defense Waste Processing Facility (DWPF) process has been evaluated to ensure that a subcritical condition is maintained. It was determined that the risk of nuclear criticality in the DWPF during initial, sludge-only operations is minimal due to the dilute concentration of fissile material in the sludge combined with excess neutron absorbers

Waste glass melting stages

International Nuclear Information System (INIS)

Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

1994-01-01

Three simulated nuclear waste glass feeds, consisting of dried waste and glass frit, were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600 degrees C to 1000 degrees C. Simulated melter feeds from the Hanford Waste Vitrification Plant (HWVP), the Defense Waste Processing Facility (DWPF), and Kernforschungszentru Karlsruhe (KfK) in Germany were used. The samples were thin sectioned and examined by optical microscopy to investigate the stages of the conversion from feed to glass. Various phenomena were seen, such as frit softening, bubble formation, foaming, bubble motion and removal, convective mixing, and homogenization. The behavior of different feeds was similar, although the degree of gas generation and melt homogenization varied. 2 refs., 8 tabs

New glass material oxidation and dissolution system facility: Direct conversion of surplus fissile materials, spent nuclear fuel, and other material to high-level-waste glass. Storage and disposition of weapons-usable fissile materials programmatic environmental impact statement data report: Predecisional draft

International Nuclear Information System (INIS)

Forsberg, C.W.; Elam, K.R.; Reich, W.J.

1995-01-01

With the end of the Cold War, countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. It has been recommended that these surplus fissile materials (SFMs) be processed so that they are no more accessible than plutonium in spent nuclear fuel (SNF). This SNF standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. This report provides for the PEIS the necessary input data on a new method for the disposition of SFMs: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptunium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal. The primary form of this SNF is Hanford-N SNF with preirradiation uranium enrichments between 0.95 and 1.08%. The final product is a plutonium, low-enriched-uranium, HLW, borosilicate glass for disposition in a geological repository. The proposed conversion process is the Glass Material Oxidation and Dissolution System (GMODS), which is a new process. The initial analysis of the GMODS process indicates that a MODS facility for this application would be similar in size and environmental impact to the Defense Waste Processing Facility (DWPF) at the Savannah River Site. Because of this, the detailed information available on DWPF was used as the basis for much of the GMODS input into the SFMs PEIS

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

Glass temperatures in free-standing canisters

International Nuclear Information System (INIS)

Hardy, B.J.; Hensel, S.J.

1993-01-01

The waste-forms produced by the Defense Waste Processing Facility (DWPF) are subject to the requirements of the Waste Acceptance Product Specifications (WAPS). The WAPS sets the maximum post cooldown temperature of the waste-form glass at 400 degrees C. This criterion must be satisfied for the ambient conditions and heat generation rates expected for the waste-forms. As part of the work described in task plan, WSRC-RP-93-1177, Rev. 0, a computer model was used to calculate the maximum glass temperatures in free standing wasteforms for a variety of ambient temperatures and heat generation rates

Alternative design concept for the second Glass Waste Storage Building

International Nuclear Information System (INIS)

Rainisch, R.

1992-10-01

This document presents an alternative design concept for storing canisters filled with vitrified waste produced at the Defense Waste Processing Facility (DWPF). The existing Glass Waste Storage Building (GWSB1) has the capacity to store 2,262 canisters and is projected to be completely filled by the year 2000. Current plans for glass waste storage are based on constructing a second Glass Waste Storage Building (GWSB2) once the existing Glass Waste Storage Building (GWSB1) is filled to capacity. The GWSB2 project (Project S-2045) is to provide additional storage capacity for 2,262 canisters. This project was initiated with the issue of a basic data report on March 6, 1989. In response to the basic data report Bechtel National, Inc. (BNI) prepared a draft conceptual design report (CDR) for the GWSB2 project in April 1991. In May 1991 WSRC Systems Engineering issued a revised Functional Design Criteria (FDC), the Rev. I document has not yet been approved by DOE. This document proposes an alternative design for the conceptual design (CDR) completed in April 1991. In June 1992 Project Management Department authorized Systems Engineering to further develop the proposed alternative design. The proposed facility will have a storage capacity for 2,268 canisters and will meet DWPF interim storage requirements for a five-year period. This document contains: a description of the proposed facility; a cost estimate of the proposed design; a cost comparison between the proposed facility and the design outlined in the FDC/CDR; and an overall assessment of the alternative design as compared with the reference FDC/CDR design

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

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

Devitrification of defense nuclear waste glasses: role of melt insolubles

International Nuclear Information System (INIS)

Bickford, D.F.; Jantzen, C.M.

1985-01-01

Time-temperature-transformation (TTT) curves have been determined for simulated nuclear waste glasses bounding the compositional range in the Defense Waste Processing Facility (DWPF). Formulations include all of the minor chemical elements such as ruthenium and chromium which have limited solubility in borosilicate glasses. Heterogeneous nucleation of spinel on ruthenium dioxide, and subsequent nucleation of acmite on spinel is the major devitrification path. Heterogeneous nucleation on melt insolubles causes more rapid growth of crystalline devitrification phases, than in glass free of melt insolubles. These studies point out the importance of simulating waste glass composition and processing as accurately as possible to obtain reliable estimates of glass performance. 11 refs., 8 figs., 1 tab

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

Evaluation of a high-level waste radiological maintenance facility

International Nuclear Information System (INIS)

Collins, K.J.

1998-01-01

The Savannah River Site''s (SRS) Defense Waste Processing Facility (DWPF) near Aiken, SC is the nation''s first and world''s largest high level waste vitrification facility. DWPF began, operations in March 1996 to process radioactive waste, consisting of a matrixed predominantly 137 Cs precipitate and a predominately 90 Sr and alpha emitting sludge, into boro-silicate glass for long term storage. Presently, DWPF is processing only sludge waste and is preparing to process a combination of sludge and precipitate waste. During precipitate operations, canister dose rates are expected to exceed 10 Sv hr -1 (1000 rem hr -1 ). In sludge-only operations, canister contact gamma dose rates are approximately 15 mSv hr -1 (1500 mrem hr -1 ). Transferable contamination levels have been greater than 10 mSv hr -1 (100 cm 2 ) -1 for beta-gamma emitters and into the millions of Bq (100 cm 2 ) -1 for the alpha emitting radionuclides. This paper presents an evaluation of the radiological maintenance areas and their ability to support radiological work

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

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

Defense waste processing facility startup progress report

International Nuclear Information System (INIS)

Iverson, D.C.; Elder, H.H.

1992-01-01

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. About 83 million gallons of high level waste produced since operation began have been consolidated into 33 million gallons by evaporation at the waste tank farm. The Department of Energy has authorized the construction of the Defense Waste Processing Facility (DWPF) to immobilize the waste as a durable borosilicate glass contained in stainless steel canisters, prior to emplacement in a federal repository. The DWPF is now mechanically complete and undergoing commissioning and run-in activities. Cold startup testing using simulated non-radioactive feeds is scheduled to begin in November 1992 with radioactive operation scheduled to begin in May 1994. While technical issues have been identified which can potentially affect DWPF operation, they are not expected to negatively impact the start of non-radioactive startup testing

Nucleation and crystal growth behavior of nepheline in simulated high-level waste glasses

Energy Technology Data Exchange (ETDEWEB)

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

2017-09-26

The Savannah River National Laboratory (SRNL) has been tasked with supporting glass formulation development and process control strategies in key technical areas, relevant to the Department of Energy’s Office of River Protection (DOE-ORP) and related to high-level waste (HLW) vitrification at the Waste Treatment and Immobilization Plant (WTP). Of specific interest is the development of predictive models for crystallization of nepheline (NaAlSiO4) in HLW glasses formulated at high alumina concentrations. This report summarizes recent progress by researchers at SRNL towards developing a predicative tool for quantifying nepheline crystallization in HLW glass canisters using laboratory experiments. In this work, differential scanning calorimetry (DSC) was used to obtain the temperature regions over which nucleation and growth of nepheline occur in three simulated HLW glasses - two glasses representative of WTP projections and one glass representative of the Defense Waste Processing Facility (DWPF) product. The DWPF glass, which has been studied previously, was chosen as a reference composition and for comparison purposes. Complementary quantitative X-ray diffraction (XRD) and optical microscopy confirmed the validity of the methodology to determine nucleation and growth behavior as a function of temperature. The nepheline crystallization growth region was determined to generally extend from ~ 500 to >850 °C, with the maximum growth rates occurring between 600 and 700 °C. For select WTP glass compositions (high Al2O3 and B2O3), the nucleation range extended from ~ 450 to 600 °C, with the maximum nucleation rates occurring at ~ 530 °C. For the DWPF glass composition, the nucleation range extended from ~ 450 to 750 °C with the maximum nucleation rate occurring at ~ 640 °C. The nepheline growth at the peak temperature, as determined by XRD, was between 35 - 75 wt.% /hour. A maximum nepheline growth rate of ~ 0.1 mm/hour at 700 °C was measured for the DWPF

Nucleation and crystal growth behavior of nepheline in simulated high-level waste glasses

International Nuclear Information System (INIS)

Fox, K.; Amoroso, J.; Mcclane, D.

2017-01-01

The Savannah River National Laboratory (SRNL) has been tasked with supporting glass formulation development and process control strategies in key technical areas, relevant to the Department of Energy's Office of River Protection (DOE-ORP) and related to high-level waste (HLW) vitrification at the Waste Treatment and Immobilization Plant (WTP). Of specific interest is the development of predictive models for crystallization of nepheline (NaAlSiO4) in HLW glasses formulated at high alumina concentrations. This report summarizes recent progress by researchers at SRNL towards developing a predicative tool for quantifying nepheline crystallization in HLW glass canisters using laboratory experiments. In this work, differential scanning calorimetry (DSC) was used to obtain the temperature regions over which nucleation and growth of nepheline occur in three simulated HLW glasses - two glasses representative of WTP projections and one glass representative of the Defense Waste Processing Facility (DWPF) product. The DWPF glass, which has been studied previously, was chosen as a reference composition and for comparison purposes. Complementary quantitative X-ray diffraction (XRD) and optical microscopy confirmed the validity of the methodology to determine nucleation and growth behavior as a function of temperature. The nepheline crystallization growth region was determined to generally extend from ~ 500 to >850 °C, with the maximum growth rates occurring between 600 and 700 °C. For select WTP glass compositions (high Al2O3 and B2O3), the nucleation range extended from ~ 450 to 600 °C, with the maximum nucleation rates occurring at ~ 530 °C. For the DWPF glass composition, the nucleation range extended from ~ 450 to 750 °C with the maximum nucleation rate occurring at ~ 640 °C. The nepheline growth at the peak temperature, as determined by XRD, was between 35 - 75 wt.% /hour. A maximum nepheline growth rate of ~ 0.1 mm/hour at 700 °C was measured for the DWPF

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

Composition and property measurements for PHA Phase 4 glasses

International Nuclear Information System (INIS)

Edwards, T.B.

2000-01-01

The results presented in this report are for nine Precipitate Hydrolysis Aqueous (PHA) Phase 4 glasses. Three of the glasses contained HM sludge at 22, 26, and 30 wt% respectively, 10 wt% PHA and 1.25 wt% monosodium titanate (MST), all on an oxide basis. The remaining six glasses were selected from the Phase 1 and Phase 2 studies (Purex sludge) but with an increased amount of MST. The high-end target for MST of 2.5 wt% oxide was missed in Phases 1 and 2 due to ∼30 wt% water content of the MST. A goal of this Phase 4 study was to determine whether this increase in titanium concentration from the MST had any impact on glass quality or processibility. Two of the glasses, pha14c and pha15c, were rebatched and melted due to apparent batching errors with pha14 and pha15. The models currently in the Defense Waste Processing Facility's (DWPF) Product Composition Control System (PCCS) were used to predict durability, homogeneity, liquidus, and viscosity for these nine glasses. All of the HM glasses and half of the Purex glasses were predicted to be phase separated, and consequently prediction of glass durability is precluded with the cument models for those glasses that failed the homogeneity constraint. If one may ignore the homogeneity constraint, the measured durabilities were within the 95% prediction limits of the model. Further efforts will be required to resolve this issue on phase separation (inhomogeneity). The liquidus model predicted unacceptable liquidus temperatures for four of the nine glasses. The approximate, bounding liquidus temperatures measured for all had upper limits of 1,000 C or less. Given the fact that liquidus temperatures were only approximated, the 30 wt% loading of Purex may be near or at the edge of acceptability for liquidus. The measured viscosities were close to the predictions of the model. For the Purex glasses, pha12c and pha15c, the measured viscosities of 28 and 23 poise, respectively, indicate that DWPF processing may be compromised

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

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.

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

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

Chemical analysis of simulated high level waste glasses to support stage III sulfate solubility modeling

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-17

The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms within the DOE complex. These wastes can contain relatively high concentrations of sulfate, which has low solubility in borosilicate glass. This is a significant issue for low-activity waste (LAW) glass and is projected to have a major impact on the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Sulfate solubility has also been a limiting factor for recent high level waste (HLW) sludge processed at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). The low solubility of sulfate in glass, along with melter and off-gas corrosion constraints, dictate that the waste be blended with lower sulfate concentration waste sources or washed to remove sulfate prior to vitrification. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerate mission completion.The objective of the current scope being pursued by SHU is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DWPF and WTP, allowing for enhanced waste loadings and waste throughput at these facilities. A series of targeted glass compositions was selected to resolve data gaps in the model and is identified as Stage III. SHU fabricated these glasses and sent samples to SRNL for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for the Stage III, simulated HLW glasses fabricated by SHU in support of the sulfate solubility model development.

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

The defense waste processing facility: the final processing step for defense high-level waste disposal

International Nuclear Information System (INIS)

Cowan, S.P.; Sprecher, W.M.; Walton, R.D.

1983-01-01

The policy of the U.S. Department of Energy is to pursue an aggressive and credible waste management program that advocates final disposal of government generated (defense) high-level nuclear wastes in a manner consistent with environmental, health, and safety responsibilities and requirements. The Defense Waste Processing Facility (DWPF) is an essential component of the Department's program. It is the first project undertaken in the United States to immobilize government generated high-level nuclear wastes for geologic disposal. The DWPF will be built at the Department's Savannah River Plant near Aiken, South Carolina. When construction is complete in 1989, the DWPF will begin processing the high-level waste at the Savannah River Plant into a borosilicate glass form, a highly insoluble and non-dispersable product, in easily handled canisters. The immobilized waste will be stored on site followed by transportation to and disposal in a Federal repository. The focus of this paper is on the DWPF. The paper discusses issues which justify the project, summarizes its technical attributes, analyzes relevant environmental and insitutional factors, describes the management approach followed in transforming technical and other concepts into concrete and steel, and concludes with observations about the future role of the facility

Defense Waste Processing Facility staged operations: environmental information document

International Nuclear Information System (INIS)

1981-11-01

Environmental information is presented relating to a staged version of the proposed Defense Waste Processing Facility (DWPF) at the Savannah River Plant. The information is intended to provide the basis for an Environmental Impact Statement. In either the integral or the staged design, the DWPF will convert the high-level waste currently stored in tanks into: a leach-resistant form containing about 99.9% of all the radioactivity, and a residual, slightly contaminated salt, which is disposed of as saltcrete. In the first stage of the staged version, the insoluble sludge portion of the waste and the long lived radionuclides contained therein will be vitrified. The waste glass will be sealed in canisters and stored onsite until shipped to a Federal repository. In the second stage, the supernate portion of the waste will be decontaminated by ion exchange. The recovered radionuclides will be transferred to the Stage 1 facility, and mixed with the sludge feed before vitrification. The residual, slightly contaminated salt solution will be mixed with Portland cement to form a concrete product (saltcrete) which will be buried onsite in an engineered landfill. This document describes the conceptual facilities and processes for producing glass waste and decontaminated salt. The environmental effects of facility construction, normal operations, and accidents are then presented. Descriptions of site and environs, alternative sites and waste disposal options, and environmental consultations and permits are given in the base Environmental Information Document

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.

The Defense Waste Processing Facility, from vision to reality

International Nuclear Information System (INIS)

Randall, C.T.

2000-01-01

When the Savannah River Plant began operation in the early 1950's producing nuclear materials for the National defense, liquid, highly radioactive waste was generated as a by-product. Since that time the waste has been stored in large, carbon steel tanks that are buried underground. In 1960 one of the tanks developed a leak, and before recovery measures could be taken, about 25-gallons of radioactive salt solution had overflowed the secondary liner and seeped into the soil surrounding the tank. Significant improvements to the tanks were made, but constant surveillance was still required. Thus, the opinion began forming that storage of the mobile, highly radioactive waste in tanks was not a responsible long-term practice. So in the late 1960's the Savannah River Laboratory began research to find a suitable long-term solution to the waste disposal problem. Several alternative waste forms were evaluated, and in 1972 the first Savannah River waste was vitrified on a laboratory scale. By the mid-1970's, the DuPont Company, prime contractor at the Savannah River Plant, began to develop a vision of constructing America's first vitrification plant to immobilize the high level radioactive waste in borosilicate glass. This vision was later championed by DuPont in the form of a vitrification plant called the Defense Waste Processing Facility (DWPF). Today, the DWPF processes Savannah River High Level Waste sludge turning it into a solid, durable waste form of borosilicate glass. The DWPF is the world's largest vitrification facility. It was brought to reality through over 25-years of research and 13-years of careful construction, tests, and reviews at a cost of approximately $3 billion dollars

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

Countercurrent Flow of Molten Glass and Air during Siphon Tests

International Nuclear Information System (INIS)

Guerrero, H.N.

2001-01-01

Siphon tests of molten glass were performed to simulate potential drainage of a radioactive waste melter, the Defense Waste Processing Facility (DWPF) at the Savannah River Site. Glass is poured from the melter through a vertical downspout that is connected to the bottom of the melter through a riser. Large flow surges have the potential of completely filling the downspout and creating a siphon effect that has the potential for complete draining of the melter. Visual observations show the exiting glass stream starts as a single-phase pipe flow, constricting into a narrow glass stream. Then a half-spherical bubble forms at the exit of the downspout. The bubble grows, extending upwards into the downspout, while the liquid flows counter-currently to one side of the spout. Tests were performed to determine what are the spout geometry and glass properties that would be conducive to siphoning, conditions for terminating the siphon, and the total amount of glass drained

International technology exchange in support of the Defense Waste Processing Facility wasteform production

International Nuclear Information System (INIS)

Kitchen, B.G.

1989-01-01

The nearly completed Defense Waste Processing Facility (DWPF) is a Department of Energy (DOE) facility at the Savannah River Site that is designed to immobilize defense high level radioactive waste (HLW) by vitrification in borosilicate glass and containment in stainless steel canisters suitable for storage in the future DOE HLW repository. The DWPF is expected to start cold operation later this year (1990), and will be the first full scale vitrification facility operating in the United States, and the largest in the world. The DOE has been coordinating technology transfer and exchange on issues relating to HLW treatment and disposal through bi-lateral agreements with several nations. For the nearly fifteen years of the vitrification program at Savannah River Laboratory, over two hundred exchanges have been conducted with a dozen international agencies involving about five-hundred foreign national specialists. These international exchanges have been beneficial to the DOE's waste management efforts through confirmation of the choice of the waste form, enhanced understanding of melter operating phenomena, support for paths forward in political/regulatory arenas, confirmation of costs for waste form compliance programs, and establishing the need for enhancements of melter facility designs. This paper will compare designs and schedules of the international vitrification programs, and will discuss technical areas where the exchanges have provided data that have confirmed and aided US research and development efforts, impacted the design of the DWPF and guided the planning for regulatory interaction and product acceptance

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

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

Processing of tetraphenylborate precipitates in the Savannah River Site Defense Waste Processing Facility

International Nuclear Information System (INIS)

Eibling, R.E.

1990-01-01

The Savannah River Site has generated 77 million gallons of high level radioactive waste since the early 1950's. By 1987, evaporation had reduced the concentration of the waste inventory to 35 million gallons. Currently, the wastes reside in large underground tanks as a soluble fraction stored, crystallized salts, and an insoluble fraction, sludge, which consists of hydrated transition metal oxides. The bulk of the radionuclides, 67 percent, are in the sludge while the crystallized salts and supernate are composed of the nitrates, nitrites, sulfates and hydroxides of sodium, potassium, and cesium. The principal radionuclide in the soluble waste is 137 Cs with traces of 90 Sr. The transformation of the high level wastes into a borosilicate glass suitable for permanent disposal is the goal of the Defense Waste Processing Facility (DWPF). To minimize the volume of glass produced, the soluble fraction of the waste is treated with sodium tetraphenylborate and sodium titanate in the waste tanks to precipitate the radioactive cesium ion and absorb the radioactive strontium ion. The precipitate is washed in the waste tanks and is then pumped to the DWPF. The precipitate, as received, is incompatible with the vitrification process because of the high aromatic carbon content and requires further chemical treatment. Within the DWPF, the precipitate is processed in the Salt Processing Cell to remove the aromatic carbon as benzene. The precipitate hydrolysis process hydrolyzes the tetraphenylborate anion to produce borate anion and benzene. The benzene is removed by distillation, decontaminated and transferred out of the DWPF for disposal

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

Investigation of Sludge Batch 3 (Macrobatch 4) Glass Sample Anomalous Behavior

International Nuclear Information System (INIS)

Bannochie, C. J.; Bibler, N. E.; Peeler, D. K.

2005-01-01

Two Defense Waste Processing Facility (DWPF) glass samples from Sludge Batch 3 (SB3) (Macrobatch 4) were received by the Savannah River National Laboratory (SRNL) on February 23, 2005. One sample, S02244, was designated for the Product Consistency Test (PCT) and elemental and radionuclide analyses. The second sample, S02247, was designated for archival storage. The samples were pulled from the melter pour stream during the feeding of Melter Feed Tank (MFT) Batch 308 and therefore roughly correspond to feed from Slurry Mix Evaporator (SME) Batches 306-308. During the course of preparing sample S02244 for PCT and other analyses two observations were made which were characterized as ''unusual'' or anomalous behavior relative to historical observations of glasses prepared for the PCT. These observations ultimately led to a series of scoping tests in order to determine more about the nature of the behavior and possible mechanisms. The first observation was the behavior of the ground glass fraction (-100 +200 mesh) for PCT analysis when contacted with deionized water during the washing phase of the PCT procedure. The behavior was analogous to that of an organic compound in the presence of water: clumping, floating on the water surface, and crawling up the beaker walls. In other words, the glass sample did not ''wet'' normally, displaying a hydrophobic behavior in water. This had never been seen before in 18 years SRNL PCT tests on either radioactive or non-radioactive glasses. Typical glass behavior is largely to settle to the bottom of the water filled beaker, though there may be suspended fines which result in some cloudiness to the wash water. The typical appearance is analogous to wetting sand. The second observation was the presence of faint black rings at the initial and final solution levels in the Teflon vessels used for the mixed acid digestion of S02244 glass conducted for compositional analysis. The digestion is composed of two stages, and at both the

Reevaluation Of Vitrified High-Level Waste Form Criteria For Potential Cost Savings At The Defense Waste Processing Facility

International Nuclear Information System (INIS)

Ray, J. W.; Marra, S. L.; Herman, C. C.

2013-01-01

At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-level waste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form

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

Analytical Plans Supporting The Sludge Batch 8 Glass Variability Study Being Conducted By Energysolutions And CUA's Vitreous State Laboratory

International Nuclear Information System (INIS)

Edwards, T. B.; Peeler, D. K.

2012-01-01

EnergySolutions (ES) and its partner, the Vitreous State Laboratory (VSL) of The Catholic University of America (CUA), are to provide engineering and technical services support to Savannah River Remediation, LLC (SRR) for ongoing operation of the Defense Waste Processing Facility (DWPF) flowsheet as well as for modifications to improve overall plant performance. SRR has requested via a statement of work that ES/VSL conduct a glass variability study (VS) for Sludge Batch 8. SRR issued a technical task request (TTR) asking that the Savannah River National Laboratory (SRNL) provide planning and data reduction support for the ES/VSL effort. This document provides two analytical plans for use by ES/VSL: one plan is to guide the measurement of the chemical composition of the study glasses while the second is to guide the measurement of the durability of the study glasses. The measurements generated by ES/VSL are to be provided to SRNL for data reduction and evaluation. SRNL is to review the results of its evaluation with ES/VSL and SRR. The results will subsequently be incorporated into a joint report with ES/VSL as a deliverable to SRR to support the processing of SB8 at DWPF

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.

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

APET methodology for Defense Waste Processing Facility: Mode C operation

International Nuclear Information System (INIS)

Taylor, R.P. Jr.; Massey, W.M.

1995-04-01

Safe operation of SRS facilities continues to be the highest priority of the Savannah River Site (SRS). One of these facilities, the Defense Waste Processing Facility or DWPF, is currently undergoing cold chemical runs to verify the design and construction preparatory to hot startup in 1995. The DWPFF is a facility designed to convert the waste currently stored in tanks at the 200-Area tank farm into a form that is suitable for long term storage in engineered surface facilities and, ultimately, geologic isolation. As a part of the program to ensure safe operation of the DWPF, a probabilistic Safety Assessment of the DWPF has been completed. The results of this analysis are incorporated into the Safety Analysis Report (SAR) for DWPF. The usual practice in preparation of Safety Analysis Reports is to include only a conservative analysis of certain design basis accidents. A major part of a Probabilistic Safety Assessment is the development and quantification of an Accident Progression Event Tree or APET. The APET provides a probabilistic representation of potential sequences along which an accident may progress. The methodology used to determine the risk of operation of the DWPF borrows heavily from methods applied to the Probabilistic Safety Assessment of SRS reactors and to some commercial reactors. This report describes the Accident Progression Event Tree developed for the Probabilistic Safety Assessment of the DWPF

TIME-TEMPERATURE-TRANSFORMATION DIAGRAMS FOR THE SLUDGE BATCH 3 - FRIT 418 GLASS SYSTEM

International Nuclear Information System (INIS)

Billings, A.; Edwards, Tommy

2009-01-01

As a part of the Waste Acceptance Product Specifications (WAPS) for Vitrified High-Level Waste Forms defined by the Department of Energy - Office of Environmental Management, the phase stability must be determined for each of the projected high-level waste (HLW) types at the Savannah River Site (SRS). Specifically, WAPS 1.4.1 requires the glass transition temperature (Tg) to be defined and time-temperature-transformation (TTT) diagrams to be developed. The Tg of a glass is an indicator of the approximate temperature where the supercooled liquid converts to a solid on cooling or conversely, where the solid begins to behave as a viscoelastic solid on heating. A TTT diagram identifies the crystalline phases that can form as a function of time and temperature for a given waste type or more specifically, the borosilicate glass waste form. In order to assess durability, the Product Consistency Test (PCT) was used and the durability results compared to the Environmental Assessment (EA) glass. The measurement of glass transition temperature and the development of TTT diagrams have already been performed for the seven Defense Waste Processing Facility (DWPF) projected compositions as defined in the Waste Form Compliance Plan (WCP). These measurements were performed before DWPF start-up and the results were incorporated in Volume 7 of the Waste Form Qualification Report (WQR). Additional information exists for other projected compositions, but overall these compositions did not consider some of the processing scenarios now envisioned for DWPF to accelerate throughput. Changes in DWPF processing strategy have required this WAPS specification to be revisited to ensure that the resulting phases have been bounded. Frit 418 was primarily used to process HLW Sludge Batch 3 (SB3) at 38% waste loading (WL) through the DWPF. The Savannah River National Laboratory (SRNL) fabricated a cache of glass from reagent grade oxides to simulate the SB3-Frit 418 system at a 38 wt % WL for glass

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

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

Reevaluation of Vitrified High-Level Waste Form Criteria for Potential Cost Savings at the Defense Waste Processing Facility - 13598

Energy Technology Data Exchange (ETDEWEB)

Ray, J.W. [Savannah River Remediation (United States); Marra, S.L.; Herman, C.C. [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)

2013-07-01

At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-level waste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form. (authors)

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

Overview - Defense Waste Processing Facility Operating Experience

International Nuclear Information System (INIS)

Norton, M.R.

2002-01-01

The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, SC is the world's largest radioactive waste vitrification facility. Radioactive operations began in March 1996 and over 1,000 canisters have been produced. This paper presents an overview of the DWPF process and a summary of recent facility operations and process improvements. These process improvements include efforts to extend the life of the DWPF melter, projects to increase facility throughput, initiatives to reduce the quantity of wastewater generated, improved remote decontamination capabilities, and improvements to remote canyon equipment to extend equipment life span. This paper also includes a review of a melt rate improvement program conducted by Savannah River Technology Center personnel. This program involved identifying the factors that impacted melt rate, conducting small scale testing of proposed process changes and developing a cost effective implementation plan

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

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

Phenomenological analyses and their application to the Defense Waste Processing Facility probabilistic safety analysis accident progression event tree. Revision 1

International Nuclear Information System (INIS)

Kalinich, D.A.; Thomas, J.K.; Gough, S.T.; Bailey, R.T.; Kearnaghan, D.P.

1995-01-01

In the Defense Waste Processing Facility (DWPF) Safety Analysis Reports (SARs) for the Savannah River Site (SRS), risk-based perspectives have been included per US Department of Energy (DOE) Order 5480.23. The NUREG-1150 Level 2/3 Probabilistic Risk Assessment (PRA) methodology was selected as the basis for calculating facility risk. The backbone of this methodology is the generation of an Accident Progression Event Tree (APET), which is solved using the EVNTRE computer code. To support the development of the DWPF APET, deterministic modeling of accident phenomena was necessary. From these analyses, (1) accident progressions were identified for inclusion into the APET; (2) branch point probabilities and any attendant parameters were quantified; and (3) the radionuclide releases to the environment from accidents were determined. The phenomena of interest for accident progressions included explosions, fires, a molten glass spill, and the response of the facility confinement system during such challenges. A variety of methodologies, from hand calculations to large system-model codes, were used in the evaluation of these phenomena

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

Defense Waste Processing Facility Process Simulation Package Life Cycle

International Nuclear Information System (INIS)

Reuter, K.

1991-01-01

The Defense Waste Processing Facility (DWPF) will be used to immobilize high level liquid radioactive waste into safe, stable, and manageable solid form. The complexity and classification of the facility requires that a performance based operator training to satisfy Department of Energy orders and guidelines. A major portion of the training program will be the application and utilization of Process Simulation Packages to assist in training the Control Room Operators on the fluctionality of the process and the application of the Distribution Control System (DCS) in operating and managing the DWPF process. The packages are being developed by the DWPF Computer and Information Systems Simulation Group. This paper will describe the DWPF Process Simulation Package Life Cycle. The areas of package scope, development, validation, and configuration management will be reviewed and discussed in detail

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

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

Cold crucible induction melters (CCIM) have been proposed as an alternative technology for waste glass melting at the Defense Waste Processing Facility (DWPF) at Savannah River Site (SRS) as well as for other waste vitrification facilities. Proponents of this technology cite high temperature operation, high tolerance for noble metals and aluminum, high waste loading, high throughput capacity, and low equipment cost as the advantages over existing Joule Heated Melter (JHM) technology. The CCIM uses induction heating to maintain molten glass at high temperature. A water-cooled helical induction coil is connected to an AC current supply, typically operating at frequencies from 100 KHz to 5 MHz. The oscillating magnetic field generated by the oscillating current flow through the coil induces eddy currents in conductive materials within the coil. Those oscillating eddy currents, in turn, generate heat in the material. In the CCIM, the induction coil surrounds a 'Cold Crucible' which is formed by metal tubes, typically copper or stainless steel. The tubes are constructed such that the magnetic field does not couple with the crucible. Therefore, the field generated by the induction coil couples primarily with the conductive medium (hot glass) within. The crucible tubes are water cooled to maintain their temperature between 100 C to 200 C so that a protective layer of molten glass and/or batch material, referred to as a 'skull', forms between them and the hot, corrosive melt. Because the protective skull is the only material directly in contact with the molten glass, the CCIM doesn't have the temperature limitations of traditional refractory lined JHM. It can be operated at melt temperatures in excess of 2000 C, allowing processing of high waste loading batches and difficult-to-melt compounds. The CCIM is poured through a bottom drain, typically through a water-cooled slide valve that starts and stops the pour stream. To promote uniform temperature distribution and

Design ampersand construction innovations of the defense waste processing facility

International Nuclear Information System (INIS)

McKibben, J.M.; Pair, C.R.; Bethmann, H.K.

1990-01-01

Construction of the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) is essentially complete. The facility is designed to convert high-level radioactive waste, now contained in large steel tanks as aqueous salts and sludge, into solid borosilicate glass in stainless steel canisters. All processing of the radioactive material and operations in a radioactive environment will be done remotely. The stringent requirements dictated by remote operation and new approaches to the glassification process led to the development of a number of first-of-a-kind pieces of equipment, new construction fabrication and erection techniques, and new applications of old techniques. The design features and construction methods used in the vitrification building and its equipment were to accomplish the objective of providing a state-of-the-art vitrification facility. 3 refs., 10 figs

Assessment of Savannah River borosilicate glass in the repository environment

International Nuclear Information System (INIS)

Plodinec, M.J.; Wicks, G.G.; Bibler, N.E.

1982-04-01

Since 1973, borosilicate glass has been studied as a matrix for the immobilization of high-level radioactive waste generated at the Savannah River Plant (SRP). In 1977, efforts began to develop and test the large-scale equipment necessary to convert the alkaline waste slurries at SRP into a durable borosilicate glass. A process has now been developed for the proposed Defense Waste Processing Facility (DWPF) which will annually produce approximately 500 canisters of SRP waste glass which will be stored on an interim basis on the Savannah River site. Current national policy calls for the permanent disposal of high-level waste in deep geologic repositories. In the repository environment, SRP waste glass will eventually be exposed to such stresses as lithostatic or hydrostatic pressures, radiation fields, and self-heating due to radioactive decay. In addition, producing and handling each canister of glass will also expose the glass to thermal and mechanical stresses. An important objective of the extensive glass characterization and testing programs of the Savannah River Laboratory (SRL) has been to determine how these stresses affect the performance of SRP waste glass. The results of these programs indicate that: these stresses will not significantly affect the performance of borosilicate glass containing SRP waste; and SRP waste glass will effectively immobilize hazardous radionuclides in the repository environment

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

Cesium Hydroxide Fusion Dissolution of Analytical Reference Glass-1 in Both Powder and Shard Form

International Nuclear Information System (INIS)

Coleman, C.J.; Spencer, W.A.

1998-04-01

CsOH has been shown to be an effective and convenient dissolution reagent for Analytical Reference Glass-1 (ARG-1). This glass standard was prepared from nonradioactive DWPF Start-up Glass. Therefore, its composition is similar to DWPF product glass and many of the glass matrices prepared at SRTC.The principal advantage of the CsOH fusion dissolution is that the reagent does not add the alkali metals Li, Na, and K usually needed by SRS customers. Commercially available CsOH is quite pure so that alkali metals can be measured accurately, often without blank corrections. CsOH fusions provide a single dissolution method for applicable glass to replace multiple dissolution schemes used by most laboratories. For example, SRTC glass samples are most commonly dissolved with a Na 2 O 2 -NaOH fusion (ref.1) and a microwave- assisted acid dissolution with HNO 3 -HF-H 3 BO 3 -HCl (ref.2). Othe laboratories use fusion methods based on KOH, LiBO 2 , and Na 2 CO 3 CsOH fusion approach reduces by half not only the work in the dissolution laboratory, but also in the spectroscopy laboratories that must analyze each solution.Experiments also revealed that glass shards or pellets are rapidly attacked if the flux temperature is raised considerably above the glass softening point. The softening point of ARG-1 glass is near 650 degrees C. Fusions performed at 750 degrees C provided complete dissolutions and accurate elemental analyses of shards. Successful dissolution of glass shards was demonstrated with CsOH, Na 2 O 2 , NaOH, KOH, and RbOH. Ability to dissolve glass shards is of considerable practical importance. Crushing glass to a fine powder is a slow and tedious task, especially for radioactive glasses dissolved in shielded cells. CsOH fusion of glass powder or shards is a convenient, cost-effective dissolution scheme applicable in SRTC, the DWPF, and the commercial glass industry

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

Integrated Disposal Facility FY 2012 Glass Testing Summary Report

Energy Technology Data Exchange (ETDEWEB)

Pierce, Eric M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kerisit, Sebastien N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Krogstad, Eirik J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burton, Sarah D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bjornstad, Bruce N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Freedman, Vicky L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cantrell, Kirk J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Snyder, Michelle MV [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2013-03-29

PNNL is conducting work to provide the technical basis for estimating radionuclide release from the engineered portion of the disposal facility for Hanford immobilized low-activity waste (ILAW). Before the ILAW can be disposed, DOE must conduct a performance assessment (PA) for the Integrated Disposal Facility (IDF) that describes the long-term impacts of the disposal facility on public health and environmental resources. As part of the ILAW glass testing program, PNNL is implementing a strategy, consisting of experimentation and modeling, to provide the technical basis for estimating radionuclide release from the glass waste form in support of future IDF PAs. Key activities in FY12 include upgrading the STOMP/eSTOMP codes to do near-field modeling, geochemical modeling of PCT tests to determine the reaction network to be used in the STOMP codes, conducting PUF tests on selected glasses to simulate and accelerate glass weathering, developing a Monte Carlo simulation tool to predict the characteristics of the weathered glass reaction layer as a function of glass composition, and characterizing glasses and soil samples exhumed from an 8-year lysimeter test. The purpose of this report is to summarize the progress made in fiscal year (FY) 2012 and the first quarter of FY 2013 toward implementing the strategy with the goal of developing an understanding of the long-term corrosion behavior of LAW glasses.

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

Defense waste processing facility radioactive operations. Part 1 - operating experience

International Nuclear Information System (INIS)

Little, D.B.; Gee, J.T.; Barnes, W.M.

1997-01-01

The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, SC is the nation's first and the world's largest vitrification facility. Following a ten year construction program and a 3 year non-radioactive test program, DWPF began radioactive operations in March 1996. This paper presents the results of the first 9 months of radioactive operations. Topics include: operations of the remote processing equipment reliability, and decontamination facilities for the remote processing equipment. Key equipment discussed includes process pumps, telerobotic manipulators, infrared camera, Holledge trademark level gauges and in-cell (remote) cranes. Information is presented regarding equipment at the conclusion of the DWPF test program it also discussed, with special emphasis on agitator blades and cooling/heating coil wear. 3 refs., 4 figs

Characterization of Analytical Reference Glass-1 (ARG-1)

International Nuclear Information System (INIS)

Smith, G.L.

1993-12-01

High-level radioactive waste may be immobilized in borosilicate glass at the West Valley Demonstration Project, West Valley, New York, the Defense Waste Processing Facility (DWPF), Aiken, South Carolina, and the Hanford Waste Vitrification Project (HWVP), Richland, Washington. The vitrified waste form will be stored in stainless steel canisters before its eventual transfer to a geologic repository for long-term disposal. Waste Acceptance Product Specifications (WAPS) (DOE 1993), Section 1.1.2 requires that the waste form producers must report the measured chemical composition of the vitrified waste in their production records before disposal. Chemical analysis of glass waste forms is receiving increased attention due to qualification requirements of vitrified waste forms. The Pacific Northwest Laboratory (PNL) has been supporting the glass producers' analytical laboratories by a continuing program of multilaboratory analytical testing using interlaboratory ''round robin'' methods. At the PNL Materials Characterization Center Analytical Round Robin 4 workshop ''Analysis of Nuclear Waste Glass and Related Materials,'' January 16--17, 1990, Pleasanton, California, the meeting attendees decided that simulated nuclear waste analytical reference glasses were needed for use as analytical standards. Use of common standard analytical reference materials would allow the glass producers' analytical laboratories to calibrate procedures and instrumentation, to control laboratory performance and conduct self-appraisals, and to help qualify their various waste forms

The Defense Waste Processing Facility: an innovative process for high-level waste immobilization

International Nuclear Information System (INIS)

Cowan, S.P.

1985-01-01

The Defense Waste Processing Facility (DWPF), under construction at the Department of Energy's Savannah River Plant (SRP), will process defense high-level radioactive waste so that it can be disposed of safely. The DWPF will immobilize the high activity fraction of the waste in borosilicate glass cast in stainless steel canisters which can be handled, stored, transported and disposed of in a geologic repository. The low-activity fraction of the waste, which represents about 90% of the high-level waste HLW volume, will be decontaminated and disposed of on the SRP site. After decontamination the canister will be welded shut by an upset resistance welding technique. In this process a slightly oversized plug is pressed into the canister opening. At the same time a large current is passed through the canister and plug. The higher resistance of the canister/plug interface causes the heat which welds the plug in place. This process provides a high quality, reliable weld by a process easily operated remotely

Systems costs for disposal of Savannah River high-level waste sludge and salt

International Nuclear Information System (INIS)

McDonell, W.R.; Goodlett, C.B.

1984-01-01

A systems cost model has been developed to support disposal of defense high-level waste sludge and salt generated at the Savannah River Plant. Waste processing activities covered by the model include decontamination of the salt by a precipitation process in the waste storage tanks, incorporation of the sludge and radionuclides removed from the salt into glass in the Defense Waste Processing Facility (DWPF), and, after interim storage, final disposal of the DWPF glass waste canisters in a federal geologic repository. Total costs for processing of waste generated to the year 2000 are estimated to be about $2.9 billion (1984 dollars); incremental unit costs for DWPF and repository disposal activities range from $120,000 to $170,000 per canister depending on DWPF processing schedules. In a representative evaluation of process alternatives, the model is used to demonstrate cost effectiveness of adjustments in the frit content of the waste glass to reduce impacts of wastes generated by the salt decontamination operations. 13 references, 8 tables

Defense waste processing facility precipitate hydrolysis process

International Nuclear Information System (INIS)

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

1986-03-01

Sodium tetraphenylborate and sodium titanate are used to assist in the concentration of soluble radionuclide in the Savannah River Plant's high-level waste. In the Defense Waste Processing Facility, concentrated tetraphenylborate/sodium titanate slurry containing cesium-137, strontium-90 and traces of plutonium from the waste tank farm is hydrolyzed in the Salt Processing Cell forming organic and aqueous phases. The two phases are then separated and the organic phase is decontaminated for incineration outside the DWPF building. The aqueous phase, containing the radionuclides and less than 10% of the original organic, is blended with the insoluble radionuclides in the high-level waste sludge and is fed to the glass melter for vitrification into borosilicate glass. During the Savannah River Laboratory's development of this process, copper (II) was found to act as a catalyst during the hydrolysis reactions, which improved the organic removal and simplified the design of the reactor

The defense waste processing facility: A status report

International Nuclear Information System (INIS)

Cowan, S.P.; Fulmer, D.C.

1987-01-01

The Defense Waste Processing Fascility (DWPF) will be the nation's first production scale facility for immobilizing high-level waste for disposal. It will also be the largest facility of its kind in the world. The technology, design, and construction efforts are on schedule for ''hot'' operation in fiscal year 1990. This paper provides a status report on the DWPF technology, design, and construction, and describes some of the challenges that have arisen during design and construction

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

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.

Process technology for vitrification of defense high-level waste at the Savannah River Plant

International Nuclear Information System (INIS)

Boersma, M.D.

1984-01-01

Vitrification in borosilicate glass is now the leading worldwide process for immobilizing high-level radioactive waste. Each vitrification project, however, has its unique mission and technical challenges. The Defense Waste Vitrification Facility (DWPF) now under construction at the Savannah River Plant will concentrate and vitrify a large amount of relatively low-power alkaline waste. Process research and development for the DWPF have produced significant advances in remote chemical operations, glass melting, off-gas treatment, slurry handling, decontamination, and welding. 6 references, 1 figure, 5 tables

Task plan: Temperatures in DWPF Glass Waste Storage Building

International Nuclear Information System (INIS)

Hardy, B.J.

1993-01-01

The Bechtel National, Inc. Detailed Design Instructions for Structural Design (DDI-02) requires that concrete components of the GWSB not exceed 150 degrees F for structural elements and 200 degrees F locally over a 24 hour period. In addition, the Waste Acceptance Product Specifications (WAPS) sets the maximum post cooldown temperature of the glass waste-form at 400 degrees C. Various scenarios can be postulated which result in elevated glass and concrete temperatures in the GWSB. Therefore, it is important to determine the concrete and glass temperatures during both normal and off-normal conditions. This document details specific tasks required to develop a technically defensible and verifiable methodology for determining maximum temperatures for the waste-forms and the GWSB concrete structures. All models used in this analysis will satisfy Quality Assurance requirements and be defensible to review and oversight committees

Integrated Disposal Facility FY2011 Glass Testing Summary Report

International Nuclear Information System (INIS)

Pierce, Eric M.; Bacon, Diana H.; Kerisit, Sebastien N.; Windisch, Charles F.; Cantrell, Kirk J.; Valenta, Michelle M.; Burton, Sarah D.; Westsik, Joseph H.

2011-01-01

Pacific Northwest National Laboratory was contracted by Washington River Protection Solutions, LLC to provide the technical basis for estimating radionuclide release from the engineered portion of the disposal facility (e.g., source term). Vitrifying the low-activity waste at Hanford is expected to generate over 1.6 x 10 5 m 3 of glass (Certa and Wells 2010). The volume of immobilized low-activity waste (ILAW) at Hanford is the largest in the DOE complex and is one of the largest inventories (approximately 8.9 x 10 14 Bq total activity) of long-lived radionuclides, principally 99 Tc (t 1/2 = 2.1 x 10 5 ), planned for disposal in a low-level waste (LLW) facility. Before the ILAW can be disposed, DOE must conduct a performance assessment (PA) for the Integrated Disposal Facility (IDF) that describes the long-term impacts of the disposal facility on public health and environmental resources. As part of the ILAW glass testing program PNNL is implementing a strategy, consisting of experimentation and modeling, in order to provide the technical basis for estimating radionuclide release from the glass waste form in support of future IDF PAs. The purpose of this report is to summarize the progress made in fiscal year (FY) 2011 toward implementing the strategy with the goal of developing an understanding of the long-term corrosion behavior of low-activity waste glasses.

High-level waste processing at the Savannah River Site: An update

International Nuclear Information System (INIS)

Marra, J.E.; Bennett, W.M.; Elder, H.H.; Lee, E.D.; Marra, S.L.; Rutland, P.L.

1997-01-01

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) in Aiken, SC mg began immobilizing high-level radioactive waste in borosilicate glass in 1996. Currently, the radioactive glass is being produced as a ''sludge-only'' composition by combining washed high-level waste sludge with glass frit. The glass is poured in stainless steel canisters which will eventually be disposed of in a permanent, geological repository. To date, DWPF has produced about 100 canisters of vitrified waste. Future processing operations will, be based on a ''coupled'' feed of washed high-level waste sludge, precipitated cesium, and glass frit. This paper provides an update of the processing activities completed to date, operational/flowsheet problems encountered, and programs underway to increase production rates

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

Integrated Disposal Facility FY2010 Glass Testing Summary Report

Energy Technology Data Exchange (ETDEWEB)

Pierce, Eric M.; Bacon, Diana H.; Kerisit, Sebastien N.; Windisch, Charles F.; Cantrell, Kirk J.; Valenta, Michelle M.; Burton, Sarah D.; Serne, R Jeffrey; Mattigod, Shas V.

2010-09-30

Pacific Northwest National Laboratory was contracted by Washington River Protection Solutions, LLC to provide the technical basis for estimating radionuclide release from the engineered portion of the disposal facility (e.g., source term). Vitrifying the low-activity waste at Hanford is expected to generate over 1.6 × 105 m3 of glass (Puigh 1999). The volume of immobilized low-activity waste (ILAW) at Hanford is the largest in the DOE complex and is one of the largest inventories (approximately 0.89 × 1018 Bq total activity) of long-lived radionuclides, principally 99Tc (t1/2 = 2.1 × 105), planned for disposal in a low-level waste (LLW) facility. Before the ILAW can be disposed, DOE must conduct a performance assessement (PA) for the Integrated Disposal Facility (IDF) that describes the long-term impacts of the disposal facility on public health and environmental resources. As part of the ILAW glass testing program PNNL is implementing a strategy, consisting of experimentation and modeling, in order to provide the technical basis for estimating radionuclide release from the glass waste form in support of future IDF PAs. The purpose of this report is to summarize the progress made in fiscal year (FY) 2010 toward implementing the strategy with the goal of developing an understanding of the long-term corrosion behavior of low-activity waste glasses. The emphasis in FY2010 was the completing an evaluation of the most sensitive kinetic rate law parameters used to predict glass weathering, documented in Bacon and Pierce (2010), and transitioning from the use of the Subsurface Transport Over Reactive Multi-phases to Subsurface Transport Over Multiple Phases computer code for near-field calculations. The FY2010 activities also consisted of developing a Monte Carlo and Geochemical Modeling framework that links glass composition to alteration phase formation by 1) determining the structure of unreacted and reacted glasses for use as input information into Monte Carlo

Integrated Disposal Facility FY2010 Glass Testing Summary Report

International Nuclear Information System (INIS)

Pierce, Eric M.; Bacon, Diana H.; Kerisit, Sebastien N.; Windisch, Charles F.; Cantrell, Kirk J.; Valenta, Michelle M.; Burton, Sarah D.; Serne, R. Jeffrey; Mattigod, Shas V.

2010-01-01

Pacific Northwest National Laboratory was contracted by Washington River Protection Solutions, LLC to provide the technical basis for estimating radionuclide release from the engineered portion of the disposal facility (e.g., source term). Vitrifying the low-activity waste at Hanford is expected to generate over 1.6 A - 105 m 3 of glass (Puigh 1999). The volume of immobilized low-activity waste (ILAW) at Hanford is the largest in the DOE complex and is one of the largest inventories (approximately 0.89 A - 1018 Bq total activity) of long-lived radionuclides, principally 99Tc (t1/2 = 2.1 A - 105), planned for disposal in a low-level waste (LLW) facility. Before the ILAW can be disposed, DOE must conduct a performance assessement (PA) for the Integrated Disposal Facility (IDF) that describes the long-term impacts of the disposal facility on public health and environmental resources. As part of the ILAW glass testing program PNNL is implementing a strategy, consisting of experimentation and modeling, in order to provide the technical basis for estimating radionuclide release from the glass waste form in support of future IDF PAs. The purpose of this report is to summarize the progress made in fiscal year (FY) 2010 toward implementing the strategy with the goal of developing an understanding of the long-term corrosion behavior of low-activity waste glasses. The emphasis in FY2010 was the completing an evaluation of the most sensitive kinetic rate law parameters used to predict glass weathering, documented in Bacon and Pierce (2010), and transitioning from the use of the Subsurface Transport Over Reactive Multi-phases to Subsurface Transport Over Multiple Phases computer code for near-field calculations. The FY2010 activities also consisted of developing a Monte Carlo and Geochemical Modeling framework that links glass composition to alteration phase formation by (1) determining the structure of unreacted and reacted glasses for use as input information into Monte Carlo

Nuclear criticality safety analysis summary report: The S-area defense waste processing facility

International Nuclear Information System (INIS)

Ha, B.C.

1994-01-01

The S-Area Defense Waste Processing Facility (DWPF) can process all of the high level radioactive wastes currently stored at the Savannah River Site with negligible risk of nuclear criticality. The characteristics which make the DWPF critically safe are: (1) abundance of neutron absorbers in the waste feeds; (2) and low concentration of fissionable material. This report documents the criticality safety arguments for the S-Area DWPF process as required by DOE orders to characterize and to justify the low potential for criticality. It documents that the nature of the waste feeds and the nature of the DWPF process chemistry preclude criticality

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].

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)

Integrated Disposal Facility FY2011 Glass Testing Summary Report

Energy Technology Data Exchange (ETDEWEB)

Pierce, Eric M.; Bacon, Diana H.; Kerisit, Sebastien N.; Windisch, Charles F.; Cantrell, Kirk J.; Valenta, Michelle M.; Burton, Sarah D.; Westsik, Joseph H.

2011-09-29

Pacific Northwest National Laboratory was contracted by Washington River Protection Solutions, LLC to provide the technical basis for estimating radionuclide release from the engineered portion of the disposal facility (e.g., source term). Vitrifying the low-activity waste at Hanford is expected to generate over 1.6 x 10{sup 5} m{sup 3} of glass (Certa and Wells 2010). The volume of immobilized low-activity waste (ILAW) at Hanford is the largest in the DOE complex and is one of the largest inventories (approximately 8.9 x 10{sup 14} Bq total activity) of long-lived radionuclides, principally {sup 99}Tc (t{sub 1/2} = 2.1 x 10{sup 5}), planned for disposal in a low-level waste (LLW) facility. Before the ILAW can be disposed, DOE must conduct a performance assessment (PA) for the Integrated Disposal Facility (IDF) that describes the long-term impacts of the disposal facility on public health and environmental resources. As part of the ILAW glass testing program PNNL is implementing a strategy, consisting of experimentation and modeling, in order to provide the technical basis for estimating radionuclide release from the glass waste form in support of future IDF PAs. The purpose of this report is to summarize the progress made in fiscal year (FY) 2011 toward implementing the strategy with the goal of developing an understanding of the long-term corrosion behavior of low-activity waste glasses.

Remote telerobotic replacement for master-slave manipulator

International Nuclear Information System (INIS)

Heckendorn, F.M.; Iverson, D.C.; LaValle, D.R.

1997-01-01

A remotely replaceable telerobotic manipulator (TRM) has been developed and deployed at the Defense Waste Processing Facility (DWPF) in support of its radioactive operation. The TRM replaces a Master-Slave Manipulator (MSM). The TRM is in use for both routine and recovery operations for the radioactive waste vitrification melter, the primary production device within the DWPF. The arm was designed for deployment and operation using an existing MSM penetration. This replacement of an existing MSM with a high power robotic device demonstrates the capability to perform similar replacement in other operating facilities. The MSM's were originally deployed in the DWPF to perform routine light capacity tasks. During the testing phase of the DWPF, prior to its radioactive startup in 5/96, the need to remove glass deposits that can form at the melter discharge during filling of glass containment canisters was identified. The combination of high radiation and contamination in the DWPF melter cell during radioactive operation eliminated personnel entry as a recovery option. Therefore remote cleaning methods had to be devised. The MSM's had neither the reach nor the strength required for this task. It became apparent that a robust manipulator arm would be required for recovery from these potential melter discharge pluggage events. The existing wall penetrations, used for the MSM's, could not be altered for seismic and radiological reasons. The new manipulator was required to be of considerable reach, due to existing physical layout, and strength, due to the glass removal requirement. Additionally, the device would have to compatible with high radiation and remote crane installation. The physical size of the manipulator and the weight of components must be consistent with the existing facilities. It was recognized early-on that a manipulator of sufficient strength to recover from a pluggage event would require robotic functions to constrain undesirable motions

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

Process arrangement options for Defense waste immobilization

International Nuclear Information System (INIS)

1980-02-01

Current plans are to immobilize the SRP high-level liquid wastes in a high integrity form. Borosilicate glass was selected in 1977 as the reference waste form and a mjaor effort is currently underway to develop the required technology. A large new facility, referred to as the Defense Waste Processing Facility (DWPF) is being designed to carry out this mission, with project authorization targeted for 1982 and plant startup in 1989. However, a number of other process arrangements or manufacturing strategies, including staging the major elements of the project or using existing SRP facilities for some functions, have been suggested in lieu of building the reference DWPF. This study assesses these various options and compares them on a technical and cost basis with the DWPF. Eleven different manufacturing options for SRP defense waste solidification were examined in detail. These cases are: (1) vitrification of acid waste at current generation rate; (2) vitrification of current rate acid waste and caustic sludge; (3 and 4) vitrification of the sludge portion of neutralized waste; (5) decontamination of salt cake and storage of concentrated cesium and strontium for later immobilization; (6) processing waste in a facility with lower capacity than the DWPF; (7) processing waste in a combination of existing and new facilities; (8) waste immobilization in H Canyon; (9) vitrification of both sludge and salt; (10) DWPF with onsite storage; (11) deferred authorization of DWPF

Plutonium Immobilization Can Loading Conceptual Design

Energy Technology Data Exchange (ETDEWEB)

Kriikku, E.

1999-05-13

'The Plutonium Immobilization Facility will encapsulate plutonium in ceramic pucks and seal the pucks inside welded cans. Remote equipment will place these cans in magazines and the magazines in a Defense Waste Processing Facility (DWPF) canister. The DWPF will fill the canister with glass for permanent storage. This report discusses the Plutonium Immobilization can loading conceptual design and includes a process block diagram, process description, preliminary equipment specifications, and several can loading issues. This report identifies loading pucks into cans and backfilling cans with helium as the top priority can loading development areas.'

Plutonium Immobilization Can Loading Conceptual Design

International Nuclear Information System (INIS)

Kriikku, E.

1999-01-01

'The Plutonium Immobilization Facility will encapsulate plutonium in ceramic pucks and seal the pucks inside welded cans. Remote equipment will place these cans in magazines and the magazines in a Defense Waste Processing Facility (DWPF) canister. The DWPF will fill the canister with glass for permanent storage. This report discusses the Plutonium Immobilization can loading conceptual design and includes a process block diagram, process description, preliminary equipment specifications, and several can loading issues. This report identifies loading pucks into cans and backfilling cans with helium as the top priority can loading development areas.'

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

Accident Fault Trees for Defense Waste Processing Facility

Energy Technology Data Exchange (ETDEWEB)

Sarrack, A.G.

1999-06-22

The purpose of this report is to document fault tree analyses which have been completed for the Defense Waste Processing Facility (DWPF) safety analysis. Logic models for equipment failures and human error combinations that could lead to flammable gas explosions in various process tanks, or failure of critical support systems were developed for internal initiating events and for earthquakes. These fault trees provide frequency estimates for support systems failures and accidents that could lead to radioactive and hazardous chemical releases both on-site and off-site. Top event frequency results from these fault trees will be used in further APET analyses to calculate accident risk associated with DWPF facility operations. This report lists and explains important underlying assumptions, provides references for failure data sources, and briefly describes the fault tree method used. Specific commitments from DWPF to provide new procedural/administrative controls or system design changes are listed in the ''Facility Commitments'' section. The purpose of the ''Assumptions'' section is to clarify the basis for fault tree modeling, and is not necessarily a list of items required to be protected by Technical Safety Requirements (TSRs).

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

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.

Fabrication and characterization of MCC approved testing material: ATM-9 glass

International Nuclear Information System (INIS)

Wald, J.W.

1986-06-01

The Materials Characterization Center ATM-9 glass is designed to be representative of glass to be produced by the Defense Waste Processing Facility at the Savannah River Plant, Aiken, South Carolina. ATM-9 glass contains all of the major components of the DWPF glass and corresponds to a waste loading of 29 wt %. The feedstock material for this glass was supplied by Savannah River Laboratory, Aiken, SC, as SRL-165 Black Frit to which was added Ba, Cs, Md, Nd, Zr, as well as 99 Tc, depleted U, 237 Np, 239+240 Pu, and 243 Am. The glass was produced under reducing conditions by the addition of 0.7 wt % graphite during the final melting process. Three kilograms of the glass were produced from April to May of 1984. On final melting, the glass was formed into stress-annealed rectangular bars of two sizes: 1.9 x 1.9 x 10 cm and 1.3 x 1.3 x 10 cm. Seventeen bars of each size were made. The analyzed composition of ATM-9 glass is listed. Examination by optical microscopy of a single transverse section from one bar showed random porosity estimated at 0.36 vol % with nominal pore diameters ranging from approx. 5 μm to 200 μm. Only one distinct second phase was observed and it was at a low concentraction level in the glass matrix. The phase appeared as spherical metallic particles. X-ray diffraction analysis of this same sample did not show any diffraction peaks from crystalline components, indicating that the glass contained less than 5 wt % of crystalline devitrification products. The even shading on the radiograph exposure indicated a generally uniform distribution of radioactivity throughout the glass matrix, with no distinct high-concentration regions

Defense waste processing facility at Savannah River Plant. Instrument and power jumpers

International Nuclear Information System (INIS)

Heckendorm, F.M. II.

1983-06-01

The Defense Waste Processing Facility (DWPF) for waste vitrification at the Savannah River Plant is in the final design stage. Development of equipment interconnecting devices or jumpers for use within the remotely operated processing canyon is now complete. These devices provide for the specialized instrument and electrical requirements of the DWPF process for low-voltage, high-frequency, and high-power interconnections

DEFENSE WASTE PROCESSING FACILITY ANALYTICAL METHOD VERIFICATION FOR THE SLUDGE BATCH 5 QUALIFICATION SAMPLE

International Nuclear Information System (INIS)

Click, D; Tommy Edwards, T; Henry Ajo, H

2008-01-01

For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) performs confirmation of the applicability of the digestion method to be used by the DWPF lab for elemental analysis of Sludge Receipt and Adjustment Tank (SRAT) receipt samples and SRAT product process control samples. DWPF SRAT samples are typically dissolved using a room temperature HF-HNO3 acid dissolution (i.e., DWPF Cold Chem Method, see Procedure SW4-15.201) and then analyzed by inductively coupled plasma - atomic emission spectroscopy (ICP-AES). This report contains the results and comparison of data generated from performing the Aqua Regia (AR), Sodium Peroxide/Hydroxide Fusion (PF) and DWPF Cold Chem (CC) method digestion of Sludge Batch 5 (SB5) SRAT Receipt and SB5 SRAT Product samples. The SB5 SRAT Receipt and SB5 SRAT Product samples were prepared in the SRNL Shielded Cells, and the SRAT Receipt material is representative of the sludge that constitutes the SB5 Batch composition. This is the sludge in Tank 51 that is to be transferred into Tank 40, which will contain the heel of Sludge Batch 4 (SB4), to form the SB5 Blend composition. The results for any one particular element should not be used in any way to identify the form or speciation of a particular element in the sludge or used to estimate ratios of compounds in the sludge. A statistical comparison of the data validates the use of the DWPF CC method for SB5 Batch composition. However, the difficulty that was encountered in using the CC method for SB4 brings into question the adequacy of CC for the SB5 Blend. Also, it should be noted that visible solids remained in the final diluted solutions of all samples digested by this method at SRNL (8 samples total), which is typical for the DWPF CC method but not seen in the other methods. Recommendations to the DWPF for application to SB5 based on studies to date: (1) A dissolution study should be performed on the WAPS

Fiber Fabrication Facility for Non-Oxide and Specialty Glasses

Data.gov (United States)

Federal Laboratory Consortium — FUNCTION: Unique facility for the research, development, and fabrication of non-oxide and specialty glasses and fibers in support of Navy/DoD programs.DESCRIPTION:...

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

Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories

International Nuclear Information System (INIS)

1983-06-01

The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

The Impact of the Proposed delta Gp Limits on Glass Formulation Efforts: Part II. Experimental Results

International Nuclear Information System (INIS)

PEELER, DAVID

2004-01-01

The Savannah River National Laboratory (SRNL) has initiated studies to assess alternative durability options that may provide access to compositional regions of interest in support of the accelerated cleanup mission at the Defense Waste Processing Facility (DWPF). One of the options being pursued is the redefinition of the durability model acceptability limits. Some of the conservative steps used in establishing the current limits without compromising the high confidence required for meeting the specification on the waste form quality were identified and eliminated. The results led to a set of three new Property Acceptability Region (PAR) values for the preliminary glass dissolution estimator that has the potential to allow access to compositional regions of interest to improve melt rate or waste loading. Although these limits are available for implementation, there is no driving force to do so with the current sludge batch (i.e., the current Frit 418 - Sludge Batch 3 (SB3) system is TL limited). The objectives of this task were to investigate (and generate) the incentive of applying the proposed durability limits in the Product Composition Control System (PCCS) from a glass formulation perspective. Glass compositions were identified or developed to transition into and through the region of GP acceptability as defined by the current and proposed durability limits. The progression through the newly defined acceptability region was accomplished by increasing the total alkali in the glass via higher alkali frits and/or waste loading (WL). The focus of this report is on the measured durability response as it compares to model predictions to assess the applicability and/or potential conservatism of the various limits or durability approaches. The normalized boron release values (NL [B] g/L) for the study glasses ranged from approximately 1.0 g/L to 2.0 g/L. The Product Consistency Test (PCT) responses provide evidence that implementation of the proposed GP limits will

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.

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

Cesium glass irradiation sources

International Nuclear Information System (INIS)

Plodinec, M.J.

1982-01-01

The precipitation process for the decontamination of soluble SRP wastes produces a material whose radioactivity is dominated by 137 Cs. Potentially, this material could be vitrified to produce irradiation sources similar to the Hanford CsCl sources. In this report, process steps necessary for the production of cesium glass irradiation sources (CGS), and the nature of the sources produced, are examined. Three options are considered in detail: direct vitrification of precipitation process waste; direct vitrification of this waste after organic destruction; and vitrification of cesium separated from the precipitation process waste. Direct vitrification is compatible with DWPF equipment, but process rates may be limited by high levels of combustible materials in the off-gas. Organic destruction would allow more rapid processing. In both cases, the source produced has a dose rate of 2 x 10 4 rads/hr at the surface. Cesium separation produces a source with a dose rate of 4 x 10 5 at the surface, which is nearer that of the Hanford sources (2 x 10 6 rads/hr). Additional processing steps would be required, as well as R and D to demonstrate that DWPF equipment is compatible with this intensely radioactive material

Glass Frit Clumping And Dusting

Energy Technology Data Exchange (ETDEWEB)

Steimke, J. L.

2013-09-26

DWPF mixes a slurry of glass frit (Frit 418) and dilute (1.5 wt%) formic acid solution with high level waste in the Slurry Mix Evaporator (SME). There would be advantages to introducing the frit in a non-slurry form to minimize water addition to the SME, however, adding completely dry frit has the potential to generate dust which could clog filters or condensers. Prior testing with another type of frit, Frit 320, and using a minimal amount of water reduced dust generation, however, the formation of hard clumps was observed. To examine options and behavior, a TTQAP [McCabe and Stone, 2013] was written to initiate tests that would address these concerns. Tests were conducted with four types of glass frit; Frit 320, DWPF Frit 418, Bekeson Frit 418 and Multi-Aspirator Frit 418. The last two frits are chemically identical to DWPF Frit 418 but smaller particles were removed by the respective vendors. Test results on Frit Clumping and Dusting are provided in this report. This report addresses the following seven questions. Short answers are provided below with more detailed answers to follow. 1. Will the addition of a small amount of water, 1.5 wt%, to dry DWPF Frit 418 greatly reduce the dust generation during handling at DWPF? a. Yes, a small scale test showed that adding a little water to the frit greatly reduced dust generation during handling. 2. Will the addition of small amounts of water to the frit cause clumping that will impair frit handling at DWPF? a. No, not with Frit 418. Although clumps were observed to form when 1.5 wt% water was mixed with DWPF Frit 418, then compressed and air-dried overnight, the clumps were easily crushed and did not form the hardened material noted when Frit 320 was tested. 3. What is the measured size distribution of dust generated when dry frit is handled? (This affects the feasibility and choice of processing equipment for removing the dust generating fraction of the frit before it is added to the SME.) a. The size distribution for

Glass Frit Clumping And Dusting

International Nuclear Information System (INIS)

Steimke, J. L.

2013-01-01

DWPF mixes a slurry of glass frit (Frit 418) and dilute (1.5 wt%) formic acid solution with high level waste in the Slurry Mix Evaporator (SME). There would be advantages to introducing the frit in a non-slurry form to minimize water addition to the SME, however, adding completely dry frit has the potential to generate dust which could clog filters or condensers. Prior testing with another type of frit, Frit 320, and using a minimal amount of water reduced dust generation, however, the formation of hard clumps was observed. To examine options and behavior, a TTQAP [McCabe and Stone, 2013] was written to initiate tests that would address these concerns. Tests were conducted with four types of glass frit; Frit 320, DWPF Frit 418, Bekeson Frit 418 and Multi-Aspirator Frit 418. The last two frits are chemically identical to DWPF Frit 418 but smaller particles were removed by the respective vendors. Test results on Frit Clumping and Dusting are provided in this report. This report addresses the following seven questions. Short answers are provided below with more detailed answers to follow. 1. Will the addition of a small amount of water, 1.5 wt%, to dry DWPF Frit 418 greatly reduce the dust generation during handling at DWPF? a. Yes, a small scale test showed that adding a little water to the frit greatly reduced dust generation during handling. 2. Will the addition of small amounts of water to the frit cause clumping that will impair frit handling at DWPF? a. No, not with Frit 418. Although clumps were observed to form when 1.5 wt% water was mixed with DWPF Frit 418, then compressed and air-dried overnight, the clumps were easily crushed and did not form the hardened material noted when Frit 320 was tested. 3. What is the measured size distribution of dust generated when dry frit is handled? (This affects the feasibility and choice of processing equipment for removing the dust generating fraction of the frit before it is added to the SME.) a. The size distribution for

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

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

Remote instrument/electrical wall nozzle replaement in the Defense Waste Processing Facility

International Nuclear Information System (INIS)

Heckendorn, F.M. II.

1983-09-01

The Defense Waste Processing Facility (DWPF) for waste vitrification at the Savannah River Plant is in the final design stage. Development of remotely replaceable instrument and electrical through-wall wiring is now complete. These assemblies connect the power and control signals from the high radiation environment to the personnel access areas. The ability to replace them will extend the life and lower the cost of the DWPF. 3 references, 22 figures, 2 tables

Impact of Alkali Source on Vitrification of SRS High Level Waste

International Nuclear Information System (INIS)

LAMBERT, D. P.; MILLER, D. H.; PEELER, D. K.; SMITH, M. E.; STONE, M. E.

2005-01-01

The Defense Waste Processing Facility (DWPF) Savannah River Site is currently immobilizing high level nuclear waste sludge by vitrification in borosilicate glass. The processing strategy involves blending a large batch of sludge into a feed tank, washing the sludge to reduce the amount of soluble species, then processing the large ''sludge batch'' through the DWPF. Each sludge batch is tested by the Savannah River National Laboratory (SRNL) using simulants and tests with samples of the radioactive waste to ''qualify'' the batch prior to processing in the DWPF. The DWPF pretreats the sludge by first acidifying the sludge with nitric and formic acid. The ratio of nitric to formic acid is adjusted as required to target a final glass composition that is slightly reducing (the target is for ∼20% of the iron to have a valence of two in the glass). The formic acid reduces the mercury in the feed to elemental mercury which is steam stripped from the feed. After a concentration step, the glass former (glass frit) is added as a 50 wt% slurry and the batch is concentrated to approximately 50 wt% solids. The feed slurry is then fed to a joule heated melter maintained at 1150 C. The glass must meet both processing (e.g., viscosity and liquidus temperature) and product performance (e.g., durability) constraints The alkali content of the final waste glass is a critical parameter that affects key glass properties (such as durability) as well as the processing characteristics of the waste sludge during the pretreatment and vitrification processes. Increasing the alkali content of the glass has been shown to improve the production rate of the DWPF, but the total alkali in the final glass is limited by constraints on glass durability and viscosity. Two sources of alkali contribute to the final alkali content of the glass: sodium salts in the waste supernate and sodium and lithium oxides in the glass frit added during pretreatment processes. Sodium salts in the waste supernate can

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

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)

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

Nitric acid flowsheet with late wash PHA testing

International Nuclear Information System (INIS)

Zamecnik, J.R.

1993-01-01

This Task Technical Plan outlines the activities to be conducted in the Integrated DWPF Melter System (IDMS) in ongoing support of the Defense Waste Processing Facility (DWPF) Chemical Process Cell (CPC) utilizing the Nitric Acid Flowsheet in the Sludge Receipt and Adjustment Tank (SRAT) and Precipitate Hydrolysis Aqueous (PHA) produced by the Late Wash Flowsheet. The IDMS facility is to be operated over a series of runs (2 to 4) using the Nitric Acid Flowsheet. The PHA will be produced with the Late Wash Flowsheet in the Precipitate Hydrolysis Experimental Facility (PHEF). All operating conditions shall simulate the expected DWPF operating conditions as closely as possible. The task objectives are to perform at least two IDMS runs with as many operating conditions as possible at nominal DWPF conditions. The major purposes of these runs are twofold: verify that the combined Late Wash and Nitric Acid flowsheets produce glass of acceptable quality without additional changes to process equipment, and determine the reproducibility of data from run to run. These runs at nominal conditions will be compared to previous runs made with PHA produced from the Late Wash flowsheet and with the Nitric Acid flowsheet in the SRAT (Purex 4 and Purex 5)

Plutonium Immobilization Can Loading Concepts

International Nuclear Information System (INIS)

Kriikku, E.; Ward, C.; Stokes, M.; Randall, B.; Steed, J.; Jones, R.; Hamilton, L.; Rogers, L.; Fiscus, J.; Dyches, G.

1998-05-01

The Plutonium Immobilization Facility will encapsulate plutonium in ceramic pucks and seal the pucks inside welded cans. Remote equipment will place these cans in magazines and the magazines in a Defense Waste Processing Facility (DWPF) canister. The DWPF will fill the canister with glass for permanent storage. This report discusses five can loading conceptual designs and the lists the advantages and disadvantages for each concept. This report identifies loading pucks into cans and backfilling cans with helium as the top priority can loading development areas. The can loading welder and cutter are very similar to the existing Savannah River Site (SRS) FB-Line bagless transfer welder and cutter and thus they are a low priority development item

Defense Waste Processing Facility (DWPF): The vitrification of high-level nuclear waste. (Latest citations from the Bibliographic database). Published Search

International Nuclear Information System (INIS)

1993-09-01

The bibliography contains citations concerning a production-scale facility and the world's largest plant for the vitrification of high-level radioactive nuclear wastes (HLW) located in the United States. Initially based on the selection of borosilicate glass as the reference waste form, the citations present the history of the development including R ampersand D projects and the actual construction of the production facility at the DOE Savannah River Plant (SRP). (Contains a minimum of 177 citations and includes a subject term index and title list.)

Testing of the Defense Waste Processing Facility Cold Chemical Dissolution Method in Sludge Batch 9 Qualification

Energy Technology Data Exchange (ETDEWEB)

Edwards, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Coleman, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Young, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Brown, L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-05-10

For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) tests the applicability of the digestion methods used by the DWPF Laboratory for elemental analysis of Sludge Receipt and Adjustment Tank (SRAT) Receipt samples and SRAT Product process control samples. DWPF SRAT samples are typically dissolved using a method referred to as the DWPF Cold Chemical or Cold Chem Method (CC), (see DWPF Procedure SW4- 15.201). Testing indicates that the CC method produced mixed results. The CC method did not result in complete dissolution of either the SRAT Receipt or SRAT Product with some fine, dark solids remaining. However, elemental analyses did not reveal extreme biases for the major elements in the sludge when compared with analyses obtained following dissolution by hot aqua regia (AR) or sodium peroxide fusion (PF) methods. The CC elemental analyses agreed with the AR and PF methods well enough that it should be adequate for routine process control analyses in the DWPF after much more extensive side-by-side tests of the CC method and the PF method are performed on the first 10 SRAT cycles of the Sludge Batch 9 (SB9) campaign. The DWPF Laboratory should continue with their plans for further tests of the CC method during these 10 SRAT cycles.

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

Paper Study Evaluations Of The Introduction Of Small Column Ion Exchange Waste Streams To The Defense Waste Processing Facility

International Nuclear Information System (INIS)

Fox, K.; Edwards, T.; Stone, M.; Koopman, D.

2010-01-01

The objective of this paper study is to provide guidance on the impact of Monosodium Titanate (MST) and Crystalline Silicotitanate (CST) streams from the Small Column Ion Exchange (SCIX) process on the Defense Waste Processing Facility (DWPF) flowsheet and glass waste form. A series of waste processing scenarios was evaluated, including projected compositions of Sludge Batches 8 through 17 (SB8 through SB17), MST additions, CST additions to Tank 40 or to a sludge batch preparation tank (Tank 42 or Tank 51, referred to generically as Tank 51 in this report), streams from the Salt Waste Processing Facility (SWPF), and two canister production rates. A wide array of potential glass frit compositions was used to support this assessment. The sludge and frit combinations were evaluated using the predictive models in the current DWPF Product Composition Control System (PCCS). The results were evaluated based on the number of frit compositions available for a particular sludge composition scenario. A large number of candidate frit compositions (e.g., several dozen to several hundred) is typically a good indicator of a sludge composition for which there is flexibility in forming an acceptable waste glass and meeting canister production rate commitments. The MST and CST streams will significantly increase the concentrations of certain components in glass, such as Nb 2 O 5 , TiO 2 , and ZrO 2 , to levels much higher than have been previously processed at DWPF. Therefore, several important assumptions, described in detail in the report, had to be made in performing the evaluations. The results of the paper studies, which must be applied carefully given the assumptions made concerning the impact of higher Ti, Zr, and Nb concentrations on model validity, provided several observations: (1) There was difficulty in identifying a reasonable number of candidate frits (and in some cases an inability to identify any candidate frits) when a waste loading of 40% is targeted for Sludge

Technical Report on the Impact of MgO on Defense Waste Processing Facility

International Nuclear Information System (INIS)

Schultz, R.L.

2000-01-01

The purpose of this study was to determine the effect(s) of removing MgO from DWPF frits to assess the impact on liquidus temperature and the durability of the glass product. Removal of MgO from the frit was hypothesized to lead to a decrease in liquidus temperature and thereby allow increased waste loading

Ecological studies related to construction of the Defense Waste Processing Facility on the Savannah River Site

International Nuclear Information System (INIS)

Scott, D.E.; Pechmann, J.H.K.; Knox, J.N.; Estes, R.A.; McGregor, J.H.; Bailey, K.

1988-12-01

The Savannah River Ecology Laboratory has completed 10 years of ecological studies related to the construction of the Defense Waste Processing Facility (DWPF) on the Savannah River Site. This progress report examines water quality studies on streams peripheral to the DWPF construction site and examines the effectiveness of ''refuge ponds'' in ameliorating the effects of construction on local amphibians. Individual papers on these topics are indexed separately. 93 refs., 15 figs., 15 tabs

Nuclear hazardous waste cost control management

International Nuclear Information System (INIS)

Selg, R.A.

1991-01-01

The effects of the waste content of glass waste forms on Savannah River high-level waste disposal costs are currently under study to adjust the glass frit content to optimize the glass waste loadings and therefore significantly reduce the overall waste disposal cost. Changes in waste content affect onsite Defense Waste Changes in waste contents affect onsite Defense Waste Processing Facility (DWPF) costs as well as offsite shipping and repository emplacement charges. A nominal 1% increase over the 28 wt% waste loading of DWPF glass would reduce disposal costs by about $50 million for Savannah River wastes generated to the year 2000. Optimization of the glass waste forms to be produced in the SWPF is being supported by economic evaluations of the impact of the forms on waste disposal costs. Glass compositions are specified for acceptable melt processing and durability characteristics, with economic effects tracked by the number of waste canisters produced. This paper presents an evaluation of the effects of variations in waste content of the glass waste forms on the overall cost of the disposal, including offsite shipment and repository emplacement, of the Savannah River high-level wastes

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

Basic Data Report -- Defense Waste Processing Facility Sludge Plant, Savannah River Plant 200-S Area

Energy Technology Data Exchange (ETDEWEB)

Amerine, D.B.

1982-09-01

This Basic Data Report for the Defense Waste Processing Facility (DWPF)--Sludge Plant was prepared to supplement the Technical Data Summary. Jointly, the two reports were intended to form the basis for the design and construction of the DWPF. To the extent that conflicting information may appear, the Basic Data Report takes precedence over the Technical Data Summary. It describes project objectives and design requirements. Pertinent data on the geology, hydrology, and climate of the site are included. Functions and requirements of the major structures are described to provide guidance in the design of the facilities. Revision 9 of the Basic Data Report was prepared to eliminate inconsistencies between the Technical Data Summary, Basic Data Report and Scopes of Work which were used to prepare the September, 1982 updated CAB. Concurrently, pertinent data (material balance, curie balance, etc.) have also been placed in the Basic Data Report. It is intended that these balances be used as a basis for the continuing design of the DWPF even though minor revisions may be made in these balances in future revisions to the Technical Data Summary.

Independent technical review of Savannah River Site Defense Waste Processing Facility technical issues

International Nuclear Information System (INIS)

1992-07-01

The Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) will vitrify high-level radioactive waste that is presently stored as liquid, salt-cake, and sludge in 51 waste-storage tanks. Construction of the DWPF began in 1984, and the Westinghouse Savannah Company (WSRC) considers the plant to be 100% turned over from construction and 91% complete. Cold-chemical runs are scheduled to begin in November 1992, and hot start up is projected for June 1994. It is estimated that the plant lifetime must exceed 15 years to complete the vitrification of the current, high-level tank waste. In a memo to the Assistant Secretary for Defense Programs (DP-1), the Assistant Secretary for Environmental Restoration and Waste management (EM-1) established the need for an Independent Technical Review (ITR), or the Red Team, to ''review process technology issues preventing start up of the DWPF.'' This report documents the findings of an Independent Technical Review (ITR) conducted by the Department of Energy (DOE), Office of Environmental Restoration and Waste Management (EM), at the request of the Assistant Secretary for Environmental Restoration and Waste Management, of specified aspects of Defense Waste Process Facility (DWPF) process technology. Information for the assessment was drawn from documents provided to the ITR Team by the Westinghouse Savannah River Company (WSRC), and presentations, discussions, interviews, and tours held at the Savannah River Site (SRS) during the weeks of February and March 9, 1992

Ecological studies related to construction of the Defense Waste Processing Facility on the Savannah River Site

Energy Technology Data Exchange (ETDEWEB)

Scott, D.E.; Pechmann, J.H.K.; Knox, J.N.; Estes, R.A.; McGregor, J.H.; Bailey, K. (ed.)

1988-12-01

The Savannah River Ecology Laboratory has completed 10 years of ecological studies related to the construction of the Defense Waste Processing Facility (DWPF) on the Savannah River Site. This progress report examines water quality studies on streams peripheral to the DWPF construction site and examines the effectiveness of refuge ponds'' in ameliorating the effects of construction on local amphibians. Individual papers on these topics are indexed separately. 93 refs., 15 figs., 15 tabs. (MHB)

Application of accident progression event tree technology to the Savannah River Site Defense Waste Processing Facility SAR analysis

International Nuclear Information System (INIS)

Brandyberry, M.D.; Baker, W.H.; Wittman, R.S.; Amos, C.N.

1993-01-01

The Accident Analysis in the Safety Analysis Report (SAR) for the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) has recently undergone an upgrade. Non-reactor SARs at SRS (and other Department of Energy (DOE) sites) use probabilistic techniques to assess the frequency of accidents at their facilities. This paper describes the application of an extension of the Accident Progression Event Tree (APET) approach to accidents at the SRS DWPF. The APET technique allows an integrated model of the facility risk to be developed, where previous probabilistic accident analyses have been limited to the quantification of the frequency and consequences of individual accident scenarios treated independently. Use of an APET allows a more structured approach, incorporating both the treatment of initiators that are common to more than one accident, and of accident progression at the facility

Final Report - Crystal Settling, Redox, and High Temperature Properties of ORP HLW and LAW Glasses, VSL-09R1510-1, Rev. 0, dated 6/18/09

Energy Technology Data Exchange (ETDEWEB)

Kruger, Albert A.; Wang, C.; Gan, H.; Pegg, I. L.; Chaudhuri, M.; Kot, W.; Feng, Z.; Viragh, C.; McKeown, D. A.; Joseph, I.; Muller, I. S.; Cecil, R.; Zhao, W.

2013-11-13

The radioactive tank waste treatment programs at the U. S. Department of Energy (DOE) have featured joule heated ceramic melter technology for the vitrification of high level waste (HLW). The Hanford Tank Waste Treatment and Immobilization Plant (WTP) employs this same basic technology not only for the vitrification of HLW streams but also for the vitrification of Low Activity Waste (LAW) streams. Because of the much greater throughput rates required of the WTP as compared to the vitrification facilities at the West Valley Demonstration Project (WVDP) or the Defense Waste Processing Facility (DWPF), the WTP employs advanced joule heated melters with forced mixing of the glass pool (bubblers) to improve heat and mass transport and increase melting rates. However, for both HLW and LAW treatment, the ability to increase waste loadings offers the potential to significantly reduce the amount of glass that must be produced and disposed and, therefore, the overall project costs. This report presents the results from a study to investigate several glass property issues related to WTP HLW and LAW vitrification: crystal formation and settling in selected HLW glasses; redox behavior of vanadium and chromium in selected LAW glasses; and key high temperature thermal properties of representative HLW and LAW glasses. The work was conducted according to Test Plans that were prepared for the HLW and LAW scope, respectively. One part of this work thus addresses some of the possible detrimental effects due to considerably higher crystal content in waste glass melts and, in particular, the impact of high crystal contents on the flow property of the glass melt and the settling rate of representative crystalline phases in an environment similar to that of an idling glass melter. Characterization of vanadium redox shifts in representative WTP LAW glasses is the second focal point of this work. The third part of this work focused on key high temperature thermal properties of

Formic Acid Free Flowsheet Development To Eliminate Catalytic Hydrogen Generation In The Defense Waste Processing

Energy Technology Data Exchange (ETDEWEB)

Lambert, Dan P.; Stone, Michael E.; Newell, J. David; Fellinger, Terri L.; Bricker, Jonathan M.

2012-09-14

The Defense Waste Processing Facility (DWPF) processes legacy nuclear waste generated at the Savannah River Site (SRS) during production of plutonium and tritium demanded by the Cold War. The nuclear waste is first treated via a complex sequence of controlled chemical reactions and then vitrified into a borosilicate glass form and poured into stainless steel canisters. Converting the nuclear waste into borosilicate glass canisters is a safe, effective way to reduce the volume of the waste and stabilize the radionuclides. Testing was initiated to determine whether the elimination of formic acid from the DWPF's chemical processing flowsheet would eliminate catalytic hydrogen generation. Historically, hydrogen is generated in chemical processing of alkaline High Level Waste sludge in DWPF. In current processing, sludge is combined with nitric and formic acid to neutralize the waste, reduce mercury and manganese, destroy nitrite, and modify (thin) the slurry rheology. The noble metal catalyzed formic acid decomposition produces hydrogen and carbon dioxide. Elimination of formic acid by replacement with glycolic acid has the potential to eliminate the production of catalytic hydrogen. Flowsheet testing was performed to develop the nitric-glycolic acid flowsheet as an alternative to the nitric-formic flowsheet currently being processed at the DWPF. This new flowsheet has shown that mercury can be reduced and removed by steam stripping in DWPF with no catalytic hydrogen generation. All processing objectives were also met, including greatly reducing the Slurry Mix Evaporator (SME) product yield stress as compared to the baseline nitric/formic flowsheet. Ten DWPF tests were performed with nonradioactive simulants designed to cover a broad compositional range. No hydrogen was generated in testing without formic acid.

Defense Waste Processing Facility Recycle Stream Evaporation

International Nuclear Information System (INIS)

STONE, MICHAEL

2006-01-01

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) stabilizes high level radioactive waste (HLW) by vitrification of the waste slurries. DWPF currently produces approximately five gallons of dilute recycle for each gallon of waste vitrified. This recycle stream is currently sent to the HLW tank farm at SRS where it is processed through the HLW evaporators with the concentrate eventually sent back to the DWPF for stabilization. Limitations of the HLW evaporators and storage space constraints in the tank farm have the potential to impact the operation of the DWPF and could limit the rate that HLW is stabilized. After an evaluation of various alternatives, installation of a dedicated evaporator for the DWPF recycle stream was selected for further evaluation. The recycle stream consists primarily of process condensates from the pretreatment and vitrification processes. Other recycle streams consist of process samples, sample line flushes, sump flushes, and cleaning solutions from the decontamination and filter dissolution processes. The condensate from the vitrification process contains some species, such as sulfate, that are not appreciably volatile at low temperature and could accumulate in the system if 100% of the evaporator concentrate was returned to DWPF. These species are currently removed as required by solids washing in the tank farm. The cleaning solutions are much higher in solids content than the other streams and are generated 5-6 times per year. The proposed evaporator would be required to concentrate the recycle stream by a factor of 30 to allow the concentrate to be recycled directly to the DWPF process, with a purge stream sent to the tank farm as required to prevent buildup of sulfate and similar species in the process. The overheads are required to meet stringent constraints to allow the condensate to be sent directly to an effluent treatment plant. The proposed evaporator would nearly de-couple the DWPF process from the

Ecological studies related to construction of the Defense Waste Processing Facility on the Savannah River Site

Energy Technology Data Exchange (ETDEWEB)

Pechmann, J.H.K.; Scott, D.E.; McGregor, J.H.; Estes, R.A.; Chazal, A.C.

1993-02-01

The Defense Waste Processing Facility (DWPF) was built on the Savannah River Site (SRS) during the mid-1980's. The Savannah River Ecology Laboratory (SREL) has completed 12 years of ecological studies related to the construction of the DWPF complex. Prior to construction, the 600-acre site (S-Area) contained a Carolina bay and the headwaters of a stream. Research conducted by the SREL has focused primarily on four questions related to these wetlands: (1) Prior to construction, what fauna and flora were present at the DWPF site and at similar, yet undisturbed, alternative sites (2) By comparing the Carolina bay at the DWPF site (Sun Bay) with an undisturbed control Carolina bay (Rainbow Bay), what effect is construction having on the organisms that inhabited the DWPF site (3) By comparing control streams with streams on the periphery of the DWPF site, what effect is construction having on the peripheral streams (4) How effective have efforts been to lessen the impacts of construction, both with respect to erosion control measures and the construction of refuge ponds'' as alternative breeding sites for amphibians that formerly bred at Sun Bay Through the long-term census-taking of biota at the DWPF site and Rainbow Bay, SREL has begun to evaluate the impact of construction on the biota and the effectiveness of mitigation efforts. Similarly, the effects of erosion from the DWPF site on the water quality of S-Area peripheral streams are being assessed. This research provides supporting data relevant to the National Environmental Policy Act (NEPA) of 1969, the Endangered Species Act of 1973, Executive Orders 11988 (Floodplain Management) and 11990 (Protection of Wetlands), and United States Department of Energy (DOE) Guidelines for Compliance with Floodplain/Wetland Environmental Review Requirements (10CFR1022).

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

Plutonium Immobilization Program cold pour tests

International Nuclear Information System (INIS)

Hovis, G.L.; Stokes, M.W.; Smith, M.E.; Wong, J.W.

1999-01-01

The Plutonium Immobilization Program (PIP) is a joint venture between the Savannah River Site, Lawrence Livermore National Laboratory, Argonne National Laboratory, and Pacific Northwest National Laboratory to carry out the disposition of excess weapons-grade plutonium. This program uses the can-in-canister (CIC) approach. CIC involves encapsulating plutonium in ceramic forms (or pucks), placing the pucks in sealed stainless steel cans, placing the cans in long cylindrical magazines, latching the magazines to racks inside Defense Waste Processing Facility (DWPF) canisters, and filling the DWPF canisters with high-level waste glass. This process puts the plutonium in a stable form and makes it attractive for reuse. At present, the DWPF pours glass into empty canisters. In the CIC approach, the addition of a stainless steel rack, magazines, cans, and ceramic pucks to the canisters introduces a new set of design and operational challenges: All of the hardware installed in the canisters must maintain structural integrity at elevated (molten-glass) temperatures. This suggests that a robust design is needed. However, the amount of material added to the DWPF canister must be minimized to prevent premature glass cooling and excessive voiding caused by a large internal thermal mass. High metal temperatures, minimizing thermal mass, and glass flow paths are examples of the types of technical considerations of the equipment design process. To determine the effectiveness of the design in terms of structural integrity and glass-flow characteristics, full-scale testing will be conducted. A cold (nonradioactive) pour test program is planned to assist in the development and verification of a baseline design for the immobilization canister to be used in the PIP process. The baseline design resulting from the cold pour test program and CIC equipment development program will provide input to Title 1 design for second-stage immobilization. The cold pour tests will be conducted in two

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

SPEEDUP simulation of liquid waste batch processing. Revision 1

International Nuclear Information System (INIS)

Shannahan, K.L.; Aull, J.E.; Dimenna, R.A.

1994-01-01

The Savannah River Site (SRS) has accumulated radioactive hazardous waste for over 40 years during the time SRS made nuclear materials for the United States Department of Energy (DOE) and its predecessors. This waste is being stored as caustic slurry in a large number of 1 million gallon steel tanks, some of which were initially constructed in the early 1950's. SRS and DOE intend to clean up the Site and convert this waste into stable forms which then can be safely stored. The liquid waste will be separated into a partially decontaminated low-level and radioactive high-level waste in one feed preparation operation, In-Tank Precipitation. The low-level waste will be used to make a concrete product called saltstone in the Saltstone Facility, a part of the Defense Waste Processing Facility (DWPF). The concrete will be poured into large vaults, where it will be permanently stored. The high-level waste will be added to glass-formers and waste slurry solids from another feed preparation operation, Extended Sludge Processing. The mixture will then be converted to a stable borosilicate glass by a vitrification process that is the other major part of the DWPF. This glass will be poured into stainless steel canisters and sent to a temporary storage facility prior to delivery to a permanent underground storage site

Ecological studies related to the construction of the Defense Waste Processing Facility on the Savannah River Site

International Nuclear Information System (INIS)

Scott, D.E.; Chazel, A.C.; Pechmann, J.H.K.; Estes, R.A.

1993-06-01

The Defense Waste Processing Facility (DWPF) was built on the Savannah River Site (SRS) during the mid-1980's. The Savannah River Ecology Laboratory (SREL) has completed 14 years of ecological studies related to the construction of the DWPF complex. Prior to construction, the 600-acre site (S-Area) contained a Carolina bay and the headwaters of a stream. Research conducted by the SREL has focused primarily on four questions related to these wetlands: (1) Prior to construction, what fauna and flora were present at the DWPF site and at similar, yet undisturbed, alternative sites? (2) By comparing the Carolina bay at the DWPF site (Sun Bay) with an undisturbed control Carolina bay (Rainbow Bay), what effect is construction having on the organisms that inhabited the DWPF site? (3) By comparing control streams with streams on the periphery of the DWPF site, what effect is construction having on the peripheral streams? (4) How effective have efforts been to lessen the impacts of construction, both with respect to erosion control measures and the construction of ''refuge ponds'' as alternative breeding sites for amphibians that formerly bred at Sun Bay? Through the long-term census-taking of biota at the DWPF site and Rainbow Bay, SREL has begun to evaluate the impact of construction on the biota and the effectiveness of mitigation efforts. Similarly, the effects of erosion from the DWPF site on the water quality of S-Area peripheral streams are being assessed. This research provides supporting data relevant to the National Environmental Policy Act (NEPA) of 1969, the Endangered Species Act of 1973, Executive Orders 11988 (Floodplain Management) and 11990 (Protection of Wetlands), and United States Department of Energy (DOE) Guidelines for Compliance with Floodplain/Wetland Environmental Review Requirements (10 CFR 1022)

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

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

Recent results on the effect of gamma radiation on the durability and microstructure of DWPF glass

International Nuclear Information System (INIS)

Bibler, N.E.; Tosten, M.H.; Beam, D.C.

1989-01-01

The effect of gamma radiation on the durability and microstructure of a simulated nuclear waste glass from the Savannah River Site has been carefully investigated. Three large pieces of glass were irradiated with a Co-60 source to three doses up to a maximum dose of 3.1 x 10 10 rad. Internal samples of the large pieces of irradiated and unirradiated glass were leached in deionized water to investigate durability changes and were examined by transmission electron microscopy (TEM) to investigate microstructure changes. Leach tests were performed in triplicate at 90 degree C with crushed glass samples in deionized water. A statistical analysis of the results indicated to the 95% confidence level that the radiation did not affect the glass durability. Careful examination by TEM indicated no effect of gamma radiation on the microstructure of the glass although severe damage could be induced by the electron beam from the microscope. 19 refs., 2 figs., 3 tabs

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

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

Defense-in-depth evaluation for the New Waste Transfer Facility

International Nuclear Information System (INIS)

Hayes, T.G.; Kelly, J.L.

1995-01-01

This report fulfills part of the requirements of References 2 and 3 by documenting a Defense-In-Depth evaluation for the New Waste Transfer Facility (NWTF). This evaluation was performed using methodology similar to that used in an evaluation for the Defense Waste Processing Facility (DWPF). It differs because the DWPF evaluation was based on an existing Process Hazards Analysis (PHA) while NWTF's is based on a Preoperational Process Hazards Review (PHR) (Ref. 1). The accidents in the Process Hazards Review (PHR) were reviewed to determine those that might have significant consequences. Significance was based on the findings of the PHR, The facility design was reviewed to determine the Structures, Systems, and Components (SSCs) and administrative controls available before and after each accident. From this was developed a list of the Lines of Defense (LODs) available to contain the hazard associated with the accident. A summary of these LODs is given in Appendix C. Items are tabulated that are suggested for consideration in the functional classification as worker protection items. The specific criteria used in the evaluation is given in the methodology section of this report. The results are documented in Appendices A, B, C, and D

Defense Waste Processing Facility Nitric- Glycolic Flowsheet Chemical Process Cell Chemistry: Part 2

Energy Technology Data Exchange (ETDEWEB)

Zamecnik, J. [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)

2017-06-06

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 Savannah River National Laboratory (SRNL) from 2011 to 2016. The goal of this work was to develop empirical correlation models to predict these values from measureable variables from the chemical process so that these quantities could be predicted a-priori from the sludge or simulant 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 work on these correlations based on the aforementioned data. Previous work on these correlations was documented in a technical report covering data from 2011-2015. This current report supersedes this previous report. Further refinement of the models as additional data are collected is recommended.

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

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)

Estimates of radionuclide release from glass waste forms in a tuff repository and the effects on regulatory compliance

International Nuclear Information System (INIS)

Aines, R.D.

1986-04-01

This paper discusses preliminary estimates of the release of radionuclides from waste packages containing glass-based waste forms under the expected conditions at Yucca Mountain. These estimates can be used to evaluate the contribution of waste package performance toward meeting repository regulatory restrictions on radionuclide release. Glass waste will be held in double stainless steel canisters. After failure of the container sometime after the 300 to 1000 year containment period, the open headspace in these cans will provide the only area where standing water can accumulate and react with the glass. A maximum release rate of 0.177 g/m 2 x year or 1.3 grams per year was obtained. Normalized loss of 1.3 grams per year corresponds to 0.08 parts in 100,000 per year of the 1660 kg reference weight of DWPF glass

Ecological studies related to construction of the Defense Waste Processing Facility on the Savannah River Site. FY 1989--1990 annual report

Energy Technology Data Exchange (ETDEWEB)

Pechmann, J.H.K.; Scott, D.E.; McGregor, J.H.; Estes, R.A.; Chazal, A.C.

1993-02-01

The Defense Waste Processing Facility (DWPF) was built on the Savannah River Site (SRS) during the mid-1980`s. The Savannah River Ecology Laboratory (SREL) has completed 12 years of ecological studies related to the construction of the DWPF complex. Prior to construction, the 600-acre site (S-Area) contained a Carolina bay and the headwaters of a stream. Research conducted by the SREL has focused primarily on four questions related to these wetlands: (1) Prior to construction, what fauna and flora were present at the DWPF site and at similar, yet undisturbed, alternative sites? (2) By comparing the Carolina bay at the DWPF site (Sun Bay) with an undisturbed control Carolina bay (Rainbow Bay), what effect is construction having on the organisms that inhabited the DWPF site? (3) By comparing control streams with streams on the periphery of the DWPF site, what effect is construction having on the peripheral streams? (4) How effective have efforts been to lessen the impacts of construction, both with respect to erosion control measures and the construction of ``refuge ponds`` as alternative breeding sites for amphibians that formerly bred at Sun Bay? Through the long-term census-taking of biota at the DWPF site and Rainbow Bay, SREL has begun to evaluate the impact of construction on the biota and the effectiveness of mitigation efforts. Similarly, the effects of erosion from the DWPF site on the water quality of S-Area peripheral streams are being assessed. This research provides supporting data relevant to the National Environmental Policy Act (NEPA) of 1969, the Endangered Species Act of 1973, Executive Orders 11988 (Floodplain Management) and 11990 (Protection of Wetlands), and United States Department of Energy (DOE) Guidelines for Compliance with Floodplain/Wetland Environmental Review Requirements (10CFR1022).

Letter Report. Defense Waste Processing Facility Pour Spout Heaters - Conceptual Designs and Modeling

Energy Technology Data Exchange (ETDEWEB)

SK Sundaram; JM Perez, Jr.

2000-09-06

The Tanks Focus Area (TFA) identified a major task to address performance limitations and deficiencies of the Defense Waste Processing Facility (DWPF) now in its sixth year of operation. Design, installation, testing, monitoring, operability, and a number of other characteristics were studied by research personnel collaboratively at a number of facilities: Savannah River Technology Center (SRTC), Clemson Environmental Technologies Laboratory (CETL), Pacific Northwest National Laboratory (PNNL), and the Idaho National Engineering and Environmental Laboratory (INEEL). Because the potential limiting feature to the DWPF was identified as the pour spout/riser heater, researches on alternative design concepts originally proposed in the past were revisited. In the original works, finite element modeling was performed to evaluate temperature distribution and stress of the design currently used at the DWPF. Studies were also made to define the requirements of the design and to consider the approaches for remote removal/replacement. Their heater type/location, their remotely replaceable thermocouples, and their capabilities for remote handling characterized the five alternative designs proposed. Review comments on the alternative designs indicated a relatively wide range of advantages and disadvantages of the designs. The present report provides an overview of the design criteria, modeling results, and alternative designs. Based on a review of the past design optimization activities and an assessment of recent experience, recommendations are proposed for future consideration and improvement.

Letter Report. Defense Waste Processing Facility Pour Spout Heaters - Conceptual Designs and Modeling

International Nuclear Information System (INIS)

Sundaram, S.K.; Perez, J.M. Jr.

2000-01-01

The Tanks Focus Area (TFA) identified a major task to address performance limitations and deficiencies of the Defense Waste Processing Facility (DWPF) now in its sixth year of operation. Design, installation, testing, monitoring, operability, and a number of other characteristics were studied by research personnel collaboratively at a number of facilities: Savannah River Technology Center (SRTC), Clemson Environmental Technologies Laboratory (CETL), Pacific Northwest National Laboratory (PNNL), and the Idaho National Engineering and Environmental Laboratory (INEEL). Because the potential limiting feature to the DWPF was identified as the pour spout/riser heater, researches on alternative design concepts originally proposed in the past were revisited. In the original works, finite element modeling was performed to evaluate temperature distribution and stress of the design currently used at the DWPF. Studies were also made to define the requirements of the design and to consider the approaches for remote removal/replacement. Their heater type/location, their remotely replaceable thermocouples, and their capabilities for remote handling characterized the five alternative designs proposed. Review comments on the alternative designs indicated a relatively wide range of advantages and disadvantages of the designs. The present report provides an overview of the design criteria, modeling results, and alternative designs. Based on a review of the past design optimization activities and an assessment of recent experience, recommendations are proposed for future consideration and improvement

Materials performance in a high-level radioactive waste vitrification system

International Nuclear Information System (INIS)

Imrich, K.J.; Chandler, G.T.

1996-01-01

The Defense Waste Processing Facility (DWPF) is a Department of Energy Facility designed to vitrify highly radioactive waste. An extensive materials evaluation program has been completed on key components in the DWPF after twelve months of operation using nonradioactive simulated wastes. Results of the visual inspections of the feed preparation system indicate that the system components, which were fabricated from Hastelloy C-276, should achieve their design lives. Significant erosion was observed on agitator blades that process glass frit slurries; however, design modifications should mitigate the erosion. Visual inspections of the DWPF melter top head and off gas components, which were fabricated from Inconel 690, indicated that varying degrees of degradation occurred. Most of the components will perform satisfactorily for their two year design life. The components that suffered significant attack were the borescopes, primary film cooler brush, and feed tubes. Changes in the operation of the film cooler brush and design modifications to the feed tubes and borescopes is expected to extend their service lives to two years. A program to investigate new high temperature engineered materials and alloys with improved oxidation and high temperature corrosion resistance will be initiated

Evaluation of quartz melt rate furnace with the nitric-glycolic flowsheet

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)

2017-08-03

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 is supplemental to the Cold Cap Evaluation Furnace (CEF) testing conducted in 20141 and the Slurry-fed Melt Rate Furnace (SMRF) testing conducted in 20162 that supported Deliverable 4 of the DWPF & Saltstone Facility Engineering Technical Task Request (TTR).3 The Quartz Melt Rate Furnace (QMRF) was evaluated as a bench-scale scoping tool to potentially be used in lieu of or simply prior to the use of the larger-scale SMRF or CEF. The QMRF platform has been used previously to evaluate melt rate behavior and offgas compositions of DWPF glasses prepared from the Nitric-Formic (NF) flowsheet but not for the NG flowsheet and not with continuous feeding.4 The overall objective of the 2016-2017 testing was to evaluate the efficacy of the QMRF as a lab-scale platform for steady state, continuously fed melter testing with the NG flowsheet as an alternative to more expensive and complex testing with the SMRF or CEF platforms.

Solidification of Savannah River Plant high-level waste

International Nuclear Information System (INIS)

Maher, R.; Shafranek, L.F.; Stevens, W.R. III.

1983-01-01

The Department of Energy, in accord with recommendations from the Du Pont Company, has started construction of a Defense Waste Processing Facility (DWPF) at the Savannah River Plant. The facility should be completed by the end of 1988, and full-scale operation should begin in 1990. This facility will immobilize in borosilicate glass the large quantity of high-level radioactive waste now stored at the plant plus the waste to be generated from continued chemical reprocessing operations. The existing wastes at the Savannah River Plant will be completely converted by about 2010. 21 figures

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

Statistical experimental design for saltstone mixtures

International Nuclear Information System (INIS)

Harris, S.P.; Postles, R.L.

1991-01-01

We used a mixture experimental design for determining a window of operability for a process at the Savannah River Site Defense Waste Processing Facility (DWPF). The high-level radioactive waste at the Savannah River Site is stored in large underground carbon steel tanks. The waste consists of a supernate layer and a sludge layer. 137 Cs will be removed from the supernate by precipitation and filtration. After further processing, the supernate layer will be fixed as a grout for disposal in concrete vaults. The remaining precipitate will be processed at the DWPF with treated waste tank sludge and glass-making chemicals into borosilicate glass. The leach rate properties of the supernate grout, formed from various mixes of solidified salt waste, needed to be determined. The effective diffusion coefficients for NO 3 and Cr were used as a measure of leach rate. Various mixes of cement, Ca(OH) 2 , salt, slag and flyash were used. These constituents comprise the whole mix. Thus, a mixture experimental design was used

Defense waste processing facility project at the Savannah River Plant

International Nuclear Information System (INIS)

Baxter, R.G.; Maher, R.; Mellen, J.B.; Shafranek, L.F.; Stevens, W.R. III.

1984-01-01

The Du Pont Company is building for the Department of Energy a facility to vitrify high-level waste at the Savannah River Plant near Aiken, South Carolina. The Defense Waste Processing Facility (DWPF) will solidify existing and future radioactive wastes produced by defense activities at the site. At the present time engineering and design are 45% complete, the site has been cleared, and startup is expected in 1989. This paper will describe project status as well as features of the design. 9 figures

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

Ecological studies related to the construction of the Defense Waste Processing Facility on the Savannah River Site. Annual report, FY-1991 and FY-1992

Energy Technology Data Exchange (ETDEWEB)

Scott, D.E.; Chazel, A.C.; Pechmann, J.H.K.; Estes, R.A.

1993-06-01

The Defense Waste Processing Facility (DWPF) was built on the Savannah River Site (SRS) during the mid-1980`s. The Savannah River Ecology Laboratory (SREL) has completed 14 years of ecological studies related to the construction of the DWPF complex. Prior to construction, the 600-acre site (S-Area) contained a Carolina bay and the headwaters of a stream. Research conducted by the SREL has focused primarily on four questions related to these wetlands: (1) Prior to construction, what fauna and flora were present at the DWPF site and at similar, yet undisturbed, alternative sites? (2) By comparing the Carolina bay at the DWPF site (Sun Bay) with an undisturbed control Carolina bay (Rainbow Bay), what effect is construction having on the organisms that inhabited the DWPF site? (3) By comparing control streams with streams on the periphery of the DWPF site, what effect is construction having on the peripheral streams? (4) How effective have efforts been to lessen the impacts of construction, both with respect to erosion control measures and the construction of ``refuge ponds`` as alternative breeding sites for amphibians that formerly bred at Sun Bay? Through the long-term census-taking of biota at the DWPF site and Rainbow Bay, SREL has begun to evaluate the impact of construction on the biota and the effectiveness of mitigation efforts. Similarly, the effects of erosion from the DWPF site on the water quality of S-Area peripheral streams are being assessed. This research provides supporting data relevant to the National Environmental Policy Act (NEPA) of 1969, the Endangered Species Act of 1973, Executive Orders 11988 (Floodplain Management) and 11990 (Protection of Wetlands), and United States Department of Energy (DOE) Guidelines for Compliance with Floodplain/Wetland Environmental Review Requirements (10 CFR 1022).

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

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.

Analysis of cesium extracting solvent using GCMS and HPLC

International Nuclear Information System (INIS)

White, T.L.; Herman, C.C.; Crump, S.L.; Marinik, A.R.; Lambert, D.P.; Eibling, R.E.

2007-01-01

A high-level waste (HLW) remediation process scheduled to begin in 2007 at the Savannah River Site is the Modular Caustic Side Solvent Extraction (CSSX) Unit (MCU). The MCU will use a hydrocarbon solvent (diluent) containing a cesium extractant, a calix[4]arene compound, to extract radioactive cesium from caustic HLW. The resulting decontaminated HLW waste or raffinate will be processed into grout at the Saltstone Production Facility (SPF). The cesium containing CSSX stream will undergo washing with dilute nitric acid followed by stripping of the cesium nitrate into a very dilute nitric acid or the strip effluent stream and the CSSX solvent will be recycled. The Defense Waste Processing Facility (DWPF) will receive the strip effluent stream and immobilize the cesium into borosilicate glass. Excess CSSX solvent carryover from the MCU creates a potential flammability problem during DWPF processing. Bench-scale DWPF process testing was performed with simulated waste to determine the fate of the CSSX solvent components. A simple high performance liquid chromatography (HPLC) method was developed to identify the modifier (which is used to increase Cs extraction and extractant solubility) and extractant within the DWPF process. The diluent and trioctylamine (which is used to suppress impurity effect and ion-pair disassociation) were determined using gas chromatography mass spectroscopy (GCMS). To close the organic balance, two types of sample preparation methods were needed. One involved extracting aqueous samples with methylene chloride or hexane, and the second was capturing the off gas of the DWPF process using carbon tubes and rinsing the tubes with carbon disulfide for analysis. This paper addresses the development of the analytical methods and the bench-scale simulated waste study results. (author)

SUMMARY OF 2010 DOE EM INTERNATIONAL PROGRAM STUDIES OF WASTE GLASS STRUCTURE AND PROPERTIES

Energy Technology Data Exchange (ETDEWEB)

Fox, K.; Choi, A.; Marra, J.; Billings, A.

2011-02-07

Collaborative work between the Savannah River National Laboratory (SRNL) and SIA Radon in Russia was divided among three tasks for calendar year 2010. The first task focused on the study of simplified high level waste glass compositions with the objective of identifying the compositional drivers that lead to crystallization and poor chemical durability. The second task focused on detailed characterization of more complex waste glass compositions with unexpectedly poor chemical durabilities. The third task focused on determining the structure of select high level waste glasses made with varying frit compositions in order to improve models under development for predicting the melt rate of the Defense Waste Processing Facility (DWPF) glasses. The majority of these tasks were carried out at SIA Radon. Selection and fabrication of the glass compositions, along with chemical composition measurements and evaluations of durability were carried out at SRNL and are described in this report. SIA Radon provided three summary reports based on the outcome of the three tasks. These reports are included as appendices to this document. Briefly, the result of characterization of the Task 1 glasses may indicate that glass compositions where iron is predominantly tetrahedrally coordinated have more of a tendency to crystallize nepheline or nepheline-like phases. For the Task 2 glasses, the results suggested that the relatively low fraction of tetrahedrally coordinated boron and the relatively low concentrations of Al{sub 2}O{sub 3} available to form [BO{sub 4/2}]{sup -}Me{sup +} and [AlO{sub 4/2}]{sup -}Me{sup +} tetrahedral units are not sufficient to consume all of the alkali ions, and thus these alkali ions are easily leached from the glasses. All of the twelve Task 3 glass compositions were determined to be mainly amorphous, with some minor spinel phases. Several key structural units such as metasilicate chains and rings were identified, which confirms the current modeling

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.

Investigation of Rheological Impacts on the Defense Waste Processing Facility's Sludge Slurry Feed as Insoluble Solids and Wash Endpoints are Adjusted

International Nuclear Information System (INIS)

Fellinger, T. L.; Howard, S.J.; Lee, M.C.; Galloway, R.H.

2006-01-01

The Savannah River Site (SRS) is currently pursuing an aggressive program to empty its High Level Waste (HLW) tanks and immobilize its radioactive waste into a durable borosilicate glass in the Defense Waste Processing Facility (DWPF). To create a batch of feed for the DWPF, several tanks of radioactive sludge slurry are combined into one of the million gallon (i.e. 3.79 E06 liters) feed tanks for DWPF. Once these sludge slurries are combined, the soluble sodium and weight percent total solids are adjusted by a 'washing' process. The 'washing' process involves diluting the soluble sodium of the sludge slurry with inhibited water (0.015 M NaOH and 0.015 M NaNO 2 ) and allowing the sludge slurry to settle into two layers. The two layers in the tank consist of a clear supernate on top and a layer of settled sludge solids on the bottom. The clear supernate layer is then decanted to another hold tank. This 'washing' process is repeated until the desired wash endpoint (i.e. sodium concentration in the supernate) and weight percent total solids are achieved. A final washed batch of feed consists of approximately 500,000 gallons (i.e. 1.89 E06 liters). DWPF has already processed three batches of feed and is currently processing a fourth. Prior to processing a batch of feed in the DWPF, it must be well characterized. Samples of the prepared feed batch are sent to the Savannah River National Laboratory (SRNL) for this characterization. As a part of the SRNL characterization for the fourth batch, rheology measurements were performed. Measurements were performed at different weight percent insoluble solids loadings to mimic potential facility processing scenarios (i.e. mixing/pumping of concentrated sludge slurry). In order to determine the influence of the soluble Na on the rheological properties of the sample, the supernate of the 'as received' sample was adjusted from 1 M soluble Na to 0.5 M soluble Na by using a lab scale version of the 'washing' process. Rheology

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)

Socioeconomic assessment of defense waste processing facility impacts in the Savannah River Plant region

Energy Technology Data Exchange (ETDEWEB)

Peelle, E.; Reed, J.H.; Stevenson, R.H.

1981-09-01

The DWPF will immobilize highly radioactive defense wastes for storage on site until shipment to an approved federal repository for radioactive wastes. This document assesses the socioeconomic impacts of constructing and operating the proposed facility and presents the assessment methodology. Because various schedules and various ways of staging the construction of the DWPF are considered and because in some of these instances a large nearby construction project (the Vogtle Nuclear Power Station) may influence the socioeconomic impacts, four scenarios involving different facility options and schedules are assessed. In general, the impacts were found not to be large. In the scenario where the socioeconomic effects were the greatest, it was found that there are likely to be some impacts on schools in Barnwell County as well as a shortage of mobile homes in that county. Aiken, Allendale, and Bamberg counties are also likely to experience slight-to-moderate housing shortages. Minor impacts are anticipated for fire and police services, roads, traffic, and land use. There will be noticeable economic impact from the project. Other scenarios had fewer socioeconomic impacts.

Socioeconomic assessment of defense waste processing facility impacts in the Savannah River Plant region

International Nuclear Information System (INIS)

Peelle, E.; Reed, J.H.; Stevenson, R.H.

1981-09-01

The DWPF will immobilize highly radioactive defense wastes for storage on site until shipment to an approved federal repository for radioactive wastes. This document assesses the socioeconomic impacts of constructing and operating the proposed facility and presents the assessment methodology. Because various schedules and various ways of staging the construction of the DWPF are considered and because in some of these instances a large nearby construction project (the Vogtle Nuclear Power Station) may influence the socioeconomic impacts, four scenarios involving different facility options and schedules are assessed. In general, the impacts were found not to be large. In the scenario where the socioeconomic effects were the greatest, it was found that there are likely to be some impacts on schools in Barnwell County as well as a shortage of mobile homes in that county. Aiken, Allendale, and Bamberg counties are also likely to experience slight-to-moderate housing shortages. Minor impacts are anticipated for fire and police services, roads, traffic, and land use. There will be noticeable economic impact from the project. Other scenarios had fewer socioeconomic impacts

Analytical study plan: Shielded Cells batch 1 campaign; Revision 1

International Nuclear Information System (INIS)

Bibler, N.E.; Ha, B.C.; Hay, M.S.; Ferrara, D.M.; Andrews, M.K.

1993-01-01

Radioactive operations in the Defense Waste Processing Facility (DWPF) will require that the Savannah River Technology Center (SRTC) perform analyses and special studies with actual Savannah River Site (SRS) high-level waste sludge. SRS Tank 42 and Tank 51 will comprise the first batch of sludge to be processed in the DWPF. Approximately 25 liters of sludge from each of these tanks will be characterized and processed in the Shielded Cells of SRTC. During the campaign, processes will include sludge characterization, sludge washing, rheology determination, mixing, hydrogen evolution, feed preparation, and vitrification of the waste. To complete the campaign, the glass will be characterized to determine its durability and crystallinity. This document describes the types of samples that will be produced, the sampling schedule and analyses required, and the methods for sample and analytical control

Reduction Of Constraints For Coupled Operations

International Nuclear Information System (INIS)

Raszewski, F.; Edwards, T.

2009-01-01

The homogeneity constraint was implemented in the Defense Waste Processing Facility (DWPF) Product Composition Control System (PCCS) to help ensure that the current durability models would be applicable to the glass compositions being processed during DWPF operations. While the homogeneity constraint is typically an issue at lower waste loadings (WLs), it may impact the operating windows for DWPF operations, where the glass forming systems may be limited to lower waste loadings based on fissile or heat load limits. In the sludge batch 1b (SB1b) variability study, application of the homogeneity constraint at the measurement acceptability region (MAR) limit eliminated much of the potential operating window for DWPF. As a result, Edwards and Brown developed criteria that allowed DWPF to relax the homogeneity constraint from the MAR to the property acceptance region (PAR) criterion, which opened up the operating window for DWPF operations. These criteria are defined as: (1) use the alumina constraint as currently implemented in PCCS (Al 2 O 3 (ge) 3 wt%) and add a sum of alkali constraint with an upper limit of 19.3 wt% (ΣM 2 O 2 O 3 constraint to 4 wt% (Al 2 O 3 (ge) 4 wt%). Herman et al. previously demonstrated that these criteria could be used to replace the homogeneity constraint for future sludge-only batches. The compositional region encompassing coupled operations flowsheets could not be bounded as these flowsheets were unknown at the time. With the initiation of coupled operations at DWPF in 2008, the need to revisit the homogeneity constraint was realized. This constraint was specifically addressed through the variability study for SB5 where it was shown that the homogeneity constraint could be ignored if the alumina and alkali constraints were imposed. Additional benefit could be gained if the homogeneity constraint could be replaced by the Al 2 O 3 and sum of alkali constraint for future coupled operations processing based on projections from Revision 14 of

Ultra-stable, diode-pumped Nd-doped glass regenerative amplifier for the National Ignition Facility (NIF)

International Nuclear Information System (INIS)

Crane, J.K.; Martinez, M.; Beach, R.J.; Mitchell, S.; Pratt, G.; Christensen, J.J.

1995-12-01

We describe a diode laser-pumped Nd:glass regenerative amplifier that amplifies temporally shaped pulses with low distortion, high pulse-to- pulse stability, and high gain. This laser amplifier is a prototype subsystem for the National Ignition Facility (NIF) laser system. 2 refs., 1 fig

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

Defense Waste Processing Facility Simulant Chemical Processing Cell Studies for Sludge Batch 9

International Nuclear Information System (INIS)

Smith, Tara E.; Newell, J. David; Woodham, Wesley H.

2016-01-01

The Savannah River National Laboratory (SRNL) received a technical task request from Defense Waste Processing Facility (DWPF) and Saltstone Engineering to perform simulant tests to support the qualification of Sludge Batch 9 (SB9) and to develop the flowsheet for SB9 in the DWPF. These efforts pertained to the DWPF Chemical Process Cell (CPC). CPC experiments were performed using SB9 simulant (SB9A) to qualify SB9 for sludge-only and coupled processing using the nitric-formic flowsheet in the DWPF. Two simulant batches were prepared, one representing SB8 Tank 40H and another representing SB9 Tank 51H. The simulant used for SB9 qualification testing was prepared by blending the SB8 Tank 40H and SB9 Tank 51H simulants. The blended simulant is referred to as SB9A. Eleven CPC experiments were run with an acid stoichiometry ranging between 105% and 145% of the Koopman minimum acid equation (KMA), which is equivalent to 109.7% and 151.5% of the Hsu minimum acid factor. Three runs were performed in the 1L laboratory scale setup, whereas the remainder were in the 4L laboratory scale setup. Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles were performed on nine of the eleven. The other two were SRAT cycles only. One coupled flowsheet and one extended run were performed for SRAT and SME processing. Samples of the condensate, sludge, and off-gas were taken to monitor the chemistry of the CPC experiments.

Defense Waste Processing Facility Simulant Chemical Processing Cell Studies for Sludge Batch 9

Energy Technology Data Exchange (ETDEWEB)

Smith, Tara E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Newell, J. David [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Woodham, Wesley H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-08-10

The Savannah River National Laboratory (SRNL) received a technical task request from Defense Waste Processing Facility (DWPF) and Saltstone Engineering to perform simulant tests to support the qualification of Sludge Batch 9 (SB9) and to develop the flowsheet for SB9 in the DWPF. These efforts pertained to the DWPF Chemical Process Cell (CPC). CPC experiments were performed using SB9 simulant (SB9A) to qualify SB9 for sludge-only and coupled processing using the nitric-formic flowsheet in the DWPF. Two simulant batches were prepared, one representing SB8 Tank 40H and another representing SB9 Tank 51H. The simulant used for SB9 qualification testing was prepared by blending the SB8 Tank 40H and SB9 Tank 51H simulants. The blended simulant is referred to as SB9A. Eleven CPC experiments were run with an acid stoichiometry ranging between 105% and 145% of the Koopman minimum acid equation (KMA), which is equivalent to 109.7% and 151.5% of the Hsu minimum acid factor. Three runs were performed in the 1L laboratory scale setup, whereas the remainder were in the 4L laboratory scale setup. Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles were performed on nine of the eleven. The other two were SRAT cycles only. One coupled flowsheet and one extended run were performed for SRAT and SME processing. Samples of the condensate, sludge, and off-gas were taken to monitor the chemistry of the CPC experiments.

EXPERIMENTAL RESULTS OF THE NEPHELINE PHASE III STUDY

Energy Technology Data Exchange (ETDEWEB)

Fox, K.; Edwards, T.

2009-11-09

This study is the third phase in a series of experiments designed to reduce conservatism in the model that predicts the formation of nepheline, a crystalline phase that can reduce the durability of high level waste glass. A Phase I study developed a series of glass compositions that were very durable while their nepheline discriminator values were well below the current nepheline discriminator limit of 0.62, where nepheline is predicted to crystallize upon slow cooling. A Phase II study selected glass compositions to identify any linear effects of composition on nepheline crystallization and that were restricted to regions that fell within the validation ranges of the Defense Waste Processing Facility (DWPF) Product Composition Control System (PCCS) models. However, it was not possible to identify any linear effects of composition on chemical durability performance for this set of study glasses. The results of the Phase II study alone were not sufficient to recommend modification of the current nepheline discriminator. It was recommended that the next series of experiments continue to focus not only on compositional regions where the PCCS models are considered applicable (i.e., the model validation ranges), but also be restricted to compositional regions where the only constraint limiting processing is the current nepheline discriminator. Two methods were used in selecting glasses for this Phase III nepheline study. The first was based on the relationship of the current nepheline discriminator model to the other DWPF PCCS models, and the second was based on theory of crystallization in mineral and glass melts. A series of 29 test glass compositions was selected for this study using a combination of the two approaches. The glasses were fabricated and characterized in the laboratory. After reviewing the data, the study glasses generally met the target compositions with little issue. Product Consistency Test results correlated well with the crystallization analyses in

A facile method to fabricate close-packed concave microlens array on cylindrical glass

International Nuclear Information System (INIS)

Deng, Zefang; Chen, Feng; Yang, Qing; Liu, Hewei; Bian, Hao; Du, Guangqing; Hu, Yang; Si, Jinhai; Meng, Xiangwei; Hou, Xun

2012-01-01

This work presents a facile method to fabricate concave microlens arrays (MLAs) with controllable shape and high fill factor on cylindrical silica glass by a femtosecond laser-enhanced chemical wet etching process. The hexagonal and rectangular MLAs are flexibly fabricated on the silica glass cylinder with a diameter of 3 mm. The morphological characteristics of MLAs are measured by a scanning electron microscope and a laser scanning confocal microscope. The measurements show that the good uniformity and high packing density MLA structures are generated. It has also been demonstrated that the shape and size of the concave structures could be easily tuned by changing laser power and the arrangement of laser exposure spots. The convex MLAs replicated by the polymer casting method experience excellent image quality. (paper)

Complementary methods to study glasses and melts at large scale facilities

International Nuclear Information System (INIS)

Leydier, M.

2010-01-01

In this work, large scale facilities (neutron and synchrotron sources) were used for studying the structure and dynamic of disordered materials (liquids and glasses). In particular, three studies are presented. The first is a structural study of Ln 2 O 3 -Al 2 O 3 -SiO 2 glasses where Ln represents the cations Sc, Y and La. We combined the results obtained from x-ray and neutron diffraction and x-ray absorption experiments. This work is focused on the determination of the interatomic distances and coordination numbers for the three pairs Si-O, Al-O and Ln-O. The second is a study of the iron oxide FeO in the liquid state. Photoemission experiments at the iron absorption edge were associated with x-ray and neutron diffraction measurements. The results obtained made it possible to define a consistent structural model for liquid FeO. The third is a study of the dynamics in CaAl 2 O 4 melts. From inelastic x-ray scattering experiments, it was possible to determine the apparent and isothermal sound velocities as well as the longitudinal viscosity. These measurements were complemented by quasielastic neutron scattering experiments from which atomic diffusion coefficients were determined. This work shows the interest of combining various experimental techniques for studying glasses and melts and points out the need to associate also modelling techniques such as molecular dynamics simulations. (author)

DWPF SB6 Initial CPC Flowsheet Testing SB6-1 TO SB6-4L Tests Of SB6-A And SB6-B Simulants

International Nuclear Information System (INIS)

Lambert, D.; Pickenheim, B.; Best, D.

2009-01-01

The Defense Waste Processing Facility (DWPF) will transition from Sludge Batch 5 (SB5) processing to Sludge Batch 6 (SB6) processing in late fiscal year 2010. Tests were conducted using non-radioactive simulants of the expected SB6 composition to determine the impact of varying the acid stoichiometry during the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) processes. The work was conducted to meet the Technical Task Request (TTR) HLW/DWPF/TTR-2008-0043, Rev.0 and followed the guidelines of a Task Technical and Quality Assurance Plan (TT and QAP). The flowsheet studies are performed to evaluate the potential chemical processing issues, hydrogen generation rates, and process slurry rheological properties as a function of acid stoichiometry. These studies were conducted with the estimated SB6 composition at the time of the study. This composition assumed a blend of 101,085 kg of Tank 4 insoluble solids and 179,000 kg of Tank 12 insoluble solids. The current plans are to subject Tank 12 sludge to aluminum dissolution. Liquid Waste Operations assumed that 75% of the aluminum would be dissolved during this process. After dissolution and blending of Tank 4 sludge slurry, plans included washing the contents of Tank 51 to ∼1M Na. After the completion of washing, the plan assumes that 40 inches on Tank 40 slurry would remain for blending with the qualified SB6 material. There are several parameters that are noteworthy concerning SB6 sludge: (1) This is the second batch DWPF will be processing that contains sludge that has had a significant fraction of aluminum removed through aluminum dissolution; (2) The sludge is high in mercury, but the projected concentration is lower than SB5; (3) The sludge is high in noble metals, but the projected concentrations are lower than SB5; and(4) The sludge is high in U and Pu - components that are not added in sludge simulants. Six DWPF process simulations were completed in 4-L laboratory-scale equipment using

Incorporating functional requirements into the structural design of the Defense Waste Processing Facility

International Nuclear Information System (INIS)

Hsiu, F.J.; Ng, C.K.; Almuti, A.M.

1986-01-01

Vitrification Building-type structures have unique features and design needs. The structural design requires new concepts and custom detailing. The above special structural designs have demonstrated the importance of the five design considerations listed in the introduction. Innovative ideas and close coordination are required to achieve the design objectives. Many of these innovations have been applied to the DWPF facility which is a first of a kind

DETERMINATION OF HLW GLASS MELT RATE USING X-RAY COMPUTED TOMOGRAPHY

Energy Technology Data Exchange (ETDEWEB)

Choi, A.; Miller, D.; Immel, D.

2011-10-06

The purpose of the high-level waste (HLW) glass melt rate study is two-fold: (1) to gain a better understanding of the impact of feed chemistry on melt rate through bench-scale testing, and (2) to develop a predictive tool for melt rate in support of the on-going frit development efforts for the Defense Waste Processing Facility (DWPF). In particular, the focus is on predicting relative melt rates, not the absolute melt rates, of various HLW glass formulations solely based on feed chemistry, i.e., the chemistry of both waste and glass-forming frit for DWPF. Critical to the successful melt rate modeling is the accurate determination of the melting rates of various HLW glass formulations. The baseline procedure being used at the Savannah River National Laboratory (SRNL) is to; (1) heat a 4 inch-diameter stainless steel beaker containing a mixture of dried sludge and frit in a furnace for a preset period of time, (2) section the cooled beaker along its diameter, and (3) measure the average glass height across the sectioned face using a ruler. As illustrated in Figure 1-1, the glass height is measured for each of the 16 horizontal segments up to the red lines where relatively large-sized bubbles begin to appear. The linear melt rate (LMR) is determined as the average of all 16 glass height readings divided by the time during which the sample was kept in the furnace. This 'visual' method has proved useful in identifying melting accelerants such as alkalis and sulfate and further ranking the relative melt rates of candidate frits for a given sludge batch. However, one of the inherent technical difficulties of this method is to determine the glass height in the presence of numerous gas bubbles of varying sizes, which is prevalent especially for the higher-waste-loading glasses. That is, how the red lines are drawn in Figure 1-1 can be subjective and, therefore, may influence the resulting melt rates significantly. For example, if the red lines are drawn too low

MATRIX 2 RESULTS OF THE FY07 ENHANCED DOE HIGH-LEVEL WASTE MELTER THROUGHPUT STUDIES AT SRNL

Energy Technology Data Exchange (ETDEWEB)

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

2008-10-23

High-level waste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). The objective of this study was to generate supplemental validation data that could be used to determine the applicability of the current liquidus temperature (TL) model to expanded DWPF glass composition regions of interest based on higher WLs. Two specific flowsheets were used in this study to provide such insight: (1) Higher WL glasses (45 and 50%) based on future sludge batches that have (and have not) undergone the Al-dissolution process. (2) Coupled operations supported by the Salt Waste Processing Facility (SWPF), which increase the TiO{sub 2} concentration in glass to greater than 2 wt%. Glasses were also selected to address technical issues associated with Al{sub 2}O{sub 3} solubility, nepheline formation, and homogeneity issues for coupled operations. A test matrix of 28 glass compositions was developed to provide insight into these issues. The glasses were fabricated and characterized using chemical composition analysis, X-ray Diffraction (XRD), TL measurement and the Product Consistency Test (PCT). The results of this study are summarized below: (1) TiO{sub 2} concentrations up to {approx} 3.5 wt% were retained in DWPF type glasses, where retention is defined as the absence of crystalline TiO{sub 2} (i.e., unreacted or undissolved) in the as-fabricated glasses. Although this TiO{sub 2} content does not bound the projected SWPF high output flowsheet (up to 6 wt% TiO{sub 2} may be required in glass), these data demonstrate the potential for increasing the TiO{sub 2} limit in glass above the current limit of 2 wt

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.

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

Glass formulation development and testing for the vitrification of DWPF HLW sludge coupled with crystalline silicotitanate (CST)

International Nuclear Information System (INIS)

Andrews, M.K.; Workman, P.J.

1997-01-01

An alternative to the In Tank Precipitation and sodium titanate processes at the Savannah River Site is the removal of cesium, strontium, and plutonium from the tank supernate by ion exchange using crystalline silicotitanate (CST). This inorganic material has been shown to effectively and selectively sorb these elements from supernate. The loaded CST could then be immobilized with High-Level Waste (HLW) sludge during vitrification. Initial efforts on the development of a glass formulation for a coupled waste stream indicate that reasonable loadings of both sludge and CST can be achieved in glass

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.

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

Reduction of Constraints: Applicability of the Homogeneity Constraint for Macrobatch 3

International Nuclear Information System (INIS)

Peeler, D.K.

2001-01-01

The Product Composition Control System (PCCS) is used to determine the acceptability of each batch of Defense Waste Processing Facility (DWPF) melter feed in the Slurry Mix Evaporator (SME). This control system imposes several constraints on the composition of the contents of the SME to define acceptability. These constraints relate process or product properties to composition via prediction models. A SME batch is deemed acceptable if its sample composition measurements lead to acceptable property predictions after accounting for modeling, measurement and analytic uncertainties. The baseline document guiding the use of these data and models is ''SME Acceptability Determination for DWPF Process Control (U)'' by Brown and Postles [1996]. A minimum of three PCCS constraints support the prediction of the glass durability from a given SME batch. The Savannah River Technology Center (SRTC) is reviewing all of the PCCS constraints associated with durability. The purpose of this review is to revisit these constraints in light of the additional knowledge gained since the beginning of radioactive operations at DWPF and to identify any supplemental studies needed to amplify this knowledge so that redundant or overly conservative constraints can be eliminated or replaced by more appropriate constraints

[Glass Development Laser (GDL) Facility upgrade.] LLE Review. Quarterly report, October-December 1984. Volume 21

International Nuclear Information System (INIS)

Kim, H.

1984-01-01

This volume of the LLE Review contains articles on the upgrade of the GDL (Glass Development) system, theoretical advances in the laser fusion effort, improved target fabrication capabilities, x-ray laser research, developments in the picosecond optics research of the LLE advanced technology program, and on the National Laser Users Facility activities for October-December 1984. 56 refs., 31 figs., 3 tabs

Sludge Batch Variability Study With Frit 418

International Nuclear Information System (INIS)

Johnson, F.; Edwards, T.

2010-01-01

The Defense Waste Processing Facility (DWPF) initiated processing Sludge Batch 6 (SB6) in the summer of 2010. In support of processing, the Savannah River National Laboratory (SRNL) provided a recommendation to utilize Frit 418 to process SB6. This recommendation was based on assessments of the compositional projections for SB6 available at the time from the Liquid Waste Organization (LWO) and SRNL (using a model-based approach). To support qualification of SB6, SRNL executed a variability study to assess the applicability of the current durability models for SB6. The durability models were assessed over the expected Frit 418-SB6 composition range. Seventeen glasses were selected for the variability study based on the sludge projections used in the frit recommendation. Five of the glasses are based on the centroid of the compositional region, spanning a waste loading (WL) range of 32 to 40%. The remaining twelve glasses are extreme vertices (EVs) of the sludge region of interest for SB6 combined with Frit 418 and are all at 36% WL. These glasses were fabricated and characterized using chemical composition analysis, X-ray diffraction (XRD) and the Product Consistency Test (PCT). After initiating the SB6 variability study, the measured composition of the SB6 Tank 51 qualification glass produced at the SRNL Shielded Cells Facility indicated that thorium was present in the glass at an appreciable concentration (1.03 wt%), which made it a reportable element for SB6. This concentration of ThO 2 resulted in a second phase of experimental studies. Five glasses were formulated that were based on the centroid of the new sludge compositional region combined with Frit 418, spanning a WL range of 32 to 40%. These glasses were fabricated and characterized using chemical composition analysis and the PCT. Based on the measured PCT response, all of the glasses (with and without thorium) were acceptable with respect to the Environmental Assessment (EA) reference glass regardless of

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

THE USE OF DI WATER TO MITIGATE DUSTING FOR ADDITION OF DWPF FRIT TO THE SLURRY MIX EVAPORATOR

Energy Technology Data Exchange (ETDEWEB)

Hansen, E.

2010-07-21

The Defense Waste Processing Facility (DPWF) presently is in the process to determine means to reduce water utilization in the Slurry Mix Evaporator (SME) process, thus reducing effluent and processing times. The frit slurry addition system mixes the dry frit with water, yielding approximately a 50 weight percent slurry containing frit and the other fraction water. This slurry is discharged into the SME and excess water is removed via boiling. To reduce this water load to the SME, DWPF has proposed using a pneumatic system in conveying the frit to the SME, in essence a dry delivery system. The problem associated with utilizing a dry delivery system with the existing frit is the generation of dust when discharged into the SME. The use of water has been shown to be effective in the mining industry as well in the DOE complex to mitigate dusting. The method employed by SRNL to determine the quantity of water to mitigate dusting in dry powders was effective, between a lab and bench scale tests. In those tests, it was shown that as high as five weight percent (wt%) of water addition was required to mitigate dust from batches of glass forming minerals used by the Waste Treatment Plant at Hanford, Washington. The same method used to determine the quantity of water to mitigate dusting was used in this task to determine the quantity of water to mitigate this dusting using as-received frit. The ability for water to mitigate dusting is due to its adhesive properties as shown in Figure 1-1. Wetting the frit particles allows for the smaller frit particles (including dust) to adhere to the larger frit particles or to agglomerate into large particles. Fluids other than water can also be used, but their adhesive properties are different than water and the quantity required to mitigate dusting is different, as was observed in reference 1. Excessive water, a few weight percentages greater than that required to mitigate dusting can cause the resulting material not to flow. The primary

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

MAR Assessments Of The High Level Waste System Plan Revision 16

International Nuclear Information System (INIS)

Peeler, D.; Edwards, T.

2011-01-01

High-level waste (HLW) throughput (i.e., the amount of waste processed per unit of time) is primarily a function of two critical parameters: waste loading (WL) and melt rate. For the Defense Waste Processing Facility (DWPF), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). Significant increases in waste throughput have been achieved at DWPF since initial radioactive operations began in 1996. Key technical and operational initiatives that supported increased waste throughput included improvements in facility attainment, the Chemical Processing Cell (CPC) flowsheet, process control models and frit formulations. As a result of these key initiatives, DWPF increased WLs from a nominal 28% for Sludge Batch 2 (SB2) to ∼34 to 38% for SB3 through SB6 while maintaining or slightly improving canister fill times. Although considerable improvements in waste throughput have been obtained, future contractual waste loading targets are nominally 40%, while canister production rates are also expected to increase (to a rate of 325 to 400 canisters per year). Although implementation of bubblers have made a positive impact on increasing melt rate for recent sludge batches targeting WLs in the mid30s, higher WLs will ultimately make the feeds to DWPF more challenging to process. Savannah River Remediation (SRR) recently requested the Savannah River National Laboratory (SRNL) to perform a paper study assessment using future sludge projections to evaluate whether the current Process Composition Control System (PCCS) algorithms would provide projected operating windows to allow future contractual WL targets to be met. More specifically, the objective of this study was to evaluate future sludge batch projections (based on Revision 16 of the HLW Systems Plan) with respect to projected operating windows using current PCCS models and associated constraints. Based on the assessments, the waste loading interval over

The Impact of Waste Loading on Viscosity in the Frit 418-SB3 System

International Nuclear Information System (INIS)

PEELER, DAVID

2004-01-01

In this report, data are provided to gain insight into the potential impact of a lower viscosity glass on melter stability (i.e., pressure spikes, cold cap behavior) and/or pour stream stability. High temperature viscosity data are generated for the Frit 418-SB3 system as a function of waste loading (from 30 to 45 percent) and compared to similar data from other systems that have been (or are currently being) processed through the Defense Waste Processing Facility (DWPF) melter. The data are presented in various formats to potentially align the viscosity data with physical observations at various points in the melter system or critical DWPF processing unit operations. The expectations is that the data will be provided adequate insight into the vitrification parameters which might evolve into working solutions as DWPF strives to maximize waste throughput. This report attempts to provide insight into a physical interpretation of the data from a DWPF perspective. The theories present ed are certainly not an all inclusive list and the order in which they are present does imply a ranking, probability, or likelihood that the proposed theory is even plausible. The intent of this discussion is to provide a forum in which the viscosity data can be discussed in relation to possible mechanisms which could potentially lead to a workable solution as discussed in relation to possible solution as higher overall attainment is striven for during processing of the current or future sludge batches

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

An update on the quality assurance for the waste vitrification plants

Energy Technology Data Exchange (ETDEWEB)

Caplinger, W.H.; Shugars, D.L.; Carlson, M.K.

1990-01-01

Immobilization of high-level defense production wastes is an important step in environmental restoration. The best available technology for immobilization of this waste currently is by incorporation into borosilicate glass, i.e., vitrification. Three US sites are active in the design, construction, or operation of vitrification facilities. The status, facility description and Quality Assurance (QA) development for each facility was presented at the 1989 Energy Division Conference. This paper presents the developments since that time. The West Valley Demonstration Project (WVDP) in northwestern New York State has demonstrated the technology. At the Savannah River Site (SRS) in South Carolina the Defense Waste Processing Facility (DWPF) has completed design, construction is essentially complete, and preparation for operation is underway. The Hanford Waste Vitrification Plant (HWVP) in Washington State is in initial Detailed Design. 4 refs.

An update on the quality assurance for the waste vitrification plants

International Nuclear Information System (INIS)

Caplinger, W.H.; Shugars, D.L.; Carlson, M.K.

1990-01-01

Immobilization of high-level defense production wastes is an important step in environmental restoration. The best available technology for immobilization of this waste currently is by incorporation into borosilicate glass, i.e., vitrification. Three US sites are active in the design, construction, or operation of vitrification facilities. The status, facility description and Quality Assurance (QA) development for each facility was presented at the 1989 Energy Division Conference. This paper presents the developments since that time. The West Valley Demonstration Project (WVDP) in northwestern New York State has demonstrated the technology. At the Savannah River Site (SRS) in South Carolina the Defense Waste Processing Facility (DWPF) has completed design, construction is essentially complete, and preparation for operation is underway. The Hanford Waste Vitrification Plant (HWVP) in Washington State is in initial Detailed Design. 4 refs

Reconnaissance hydrogeologic investigation of the Defense Waste Processing Facility and Vicinity, Savannah River Plant, South Carolina

International Nuclear Information System (INIS)

Dennehy, K.F.; Prowell, D.C.; McMahon, P.B.

1989-01-01

The purposes of this report are two-fold: (1) to define the hydrogeologic conditions in the vicinity of the defense waste processing facility (DWPF) and, (2) to evaluate the potential for movement of a concentrated salt-solution waste if released at or near the DWPF. These purposes were accomplished by assembling and evaluating existing hydrogeologic data; collecting additional geologic, hydrologic, and water-quality data; developing a local geologic framework; developing a conceptual model of the local ground-water flow system; and by performing laboratory experiments to determine the mobility of salt-solution waste in surface and near-surface sediments. Although the unconsolidated sediments are about 1000 ft thick in the study area, only the Tertiary age sediments, or upper 300 ft are discussed in this report. The top of the Ellenton Formation acts as the major confining unit between the overlying aquifers in Tertiary sediments and the underlying aquifers in Cretaceous sediments; therefore, the Ellenton Formation is the vertical limit of our hydrogeologic investigation. The majority of the hydrologic data for this study come from monitoring wells at the saltstone disposal site (SDS) in Z Area (fig. 3). No recent water-level data were collected in S Area owing to the removal of S Area monitoring wells prior to construction at the DWPF. 46 refs., 26 figs., 7 tabs

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

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

High-level waste glass compendium; what it tells us concerning the durability of borosilicate waste glass

International Nuclear Information System (INIS)

Cunnane, J.C.; Allison, J.

1993-01-01

Facilities for vitrification of high-level nuclear waste in the United States are scheduled for startup in the next few years. It is, therefore, appropriate to examine the current scientific basis for understanding the corrosion of high-level waste borosilicate glass for the range of service conditions to which the glass products from these facilities may be exposed. To this end, a document has been prepared which compiles worldwide information on borosilicate waste glass corrosion. Based on the content of this document, the acceptability of canistered waste glass for geological disposal is addressed. Waste glass corrosion in a geologic repository may be due to groundwater and/or water vapor contact. The important processes that determine the glass corrosion kinetics under these conditions are discussed based on experimental evidence from laboratory testing. Testing data together with understanding of the long-term corrosion kinetics are used to estimate radionuclide release rates. These rates are discussed in terms of regulatory performance standards

High Level Waste (HLW) Processing Experience with Increased Waste Loading

International Nuclear Information System (INIS)

JANTZEN, CAROL

2004-01-01

The Defense Waste Processing Facility (DWPF) Engineering requested characterization of glass samples that were taken after the second melter had been operational for about 5 months. After the new melter had been installed, the waste loading had been increased to about 38 weight percentage after a new quasicrystalline liquidus model had been implemented. The DWPF had also switched from processing with refractory Frit 200 to a more fluid Frit 320. The samples were taken after DWPF observed very rapid buildup of deposits in the upper pour spout bore and on the pour spout insert while processing the high waste loading feedstock. These samples were evaluated using various analytical techniques to determine the cause of the crystallization. The pour stream sample was homogeneous, amorphous, and representative of the feed batch from which it was derived. Chemical analysis of the pour stream sample indicated that a waste loading of 38.5 weight per cent had been achieved. The data analysis indicated that surface crystallization, induced by temperature and oxygen fugacity gradients in the pour spout, caused surface crystallization to occur in the spout and on the insert at the higher waste loadings even though there was no crystallization in the pour stream

Antifoam Degradation Products in Off Gas and Condensate of Sludge Batch 9 Simulant Nitric-Formic Flowsheet Testing for the Defense Waste Processing Facility

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-14

Ten chemical processing cell (CPC) experiments were performed using simulant to evaluate Sludge Batch 9 for sludge-only and coupled processing using the nitric-formic flowsheet in the Defense Waste Processing Facility (DWPF). Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles were performed on eight of the ten. The other two were SRAT cycles only. Samples of the condensate, sludge, and off gas were taken to monitor the chemistry of the CPC experiments. The Savannah River National Laboratory (SRNL) has previously shown antifoam decomposes to form flammable organic products, (hexamethyldisiloxane (HMDSO), trimethylsilanol (TMS), and propanal), that are present in the vapor phase and condensate of the CPC vessels. To minimize antifoam degradation product formation, a new antifoam addition strategy was implemented at SRNL and DWPF to add antifoam undiluted.

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.

SRL canister impact tests

International Nuclear Information System (INIS)

Kelker, J.W. Jr.

1986-05-01

The Defense Waste Processing Facility (DWPF) is being constructed at the SRP for the containerization of high-level nuclear waste as a waste form for eventual permanent disposal. The waste will be incorporated in molten glass and solidified in Type 304L stainless steel canisters 2 feet in diameter x 9 feet 10 inches long. The canisters have a minimum wall thickness of 3/8 inch. Over a three-year period, nineteen drop-tests of nine canisters, filled with simulated waste glass, were made in support of the DWPF containerization program. Eight of the canister evaluation tests were of Type 304L stainless steel material and one was of commercially pure titanium. Three different length (9.44, 5.06, and 7.88 inch) nozzle configurations containing final closure upset welds were evaluated for the stainless steel canisters. All impact tests of the stainless steel canisters, which included bottom-, side-, and top-drops, were acceptable. The bottom-drop test of the titanium canister, which contained a final closure upset weld, was acceptable; however, the top-drop resulted in a breaching of the top head where it joins the nozzle. The final closure titanium upset weld was acceptable. The titanium canister wall thickness was 1/4 inch

Mercury removal from SRP radioactive waste streams using ion exchange

International Nuclear Information System (INIS)

Bibler, J.P.; Wallace, R.M.; Ebra, M.A.

1986-01-01

Mercury is present in varying concentrations in some Savannah River Plant (SRP) waste streams as a result of its use as a catalyst in the dissolution of fuel elements composed of uranium-aluminum alloys. It may be desirable to remove mercury from these streams before treatment of the waste for incorporation in glass for long-term storage. The glass forming process will also create waste from which mercury will have to be removed. The goal of mercury would be to eliminate ultimate emission of the toxic substance into the environment. This paper describes tests that demonstrate the feasibility of using a specific cation exchange resin, Duolite GT-73 for the removal of mercury from five waste streams generated at the SRP. Two of these streams are dilute; one is the condensate from a waste evaporator while the other is the effluent from an effluent treatment plant now under development. The three other streams are related to the Defense Waste Processing Facility (DWPF) that is being built at SRP. One of these streams is a concentrated salt solution (principally sodium nitrate and sodium hydroxide) that constitutes the soluble fraction of SRP waste and contains 20% mercury in the waste. The second stream is a slurry of the insoluble components in SRP waste and contains 80% of the mercury. The third stream is the offgas condensate from the glass melter system in the DWPF

Elastic modulus measurements of LDEF glasses and glass-ceramics using a speckle technique

International Nuclear Information System (INIS)

Wiedlocher, D.E.; Kinser, D.L.

1992-01-01

Elastic moduli of five glass types and the glass-ceramic Zerodur, exposed to a near-earth orbit environment on the Long Duration Exposure Facility (LDEF), were compared to that of unexposed samples. A double exposure speckle photography technique utilizing 633 nm laser light was used in the production of the speckle pattern. Subsequent illumination of a double exposed negative using the same wavelength radiation produces Young's fringes from which the in-plane displacements are measured. Stresses imposed by compressive loading produced measurable strains in the glasses and glass-ceramic

Errors of DWPF frit analysis: Final report

International Nuclear Information System (INIS)

Schumacher, R.F.

1993-01-01

Glass frit will be a major raw material for the operation of the Defense Waste Processing Facility. The frit will be controlled by certificate of conformance and a confirmatory analysis from a commercial analytical laboratory. The following effort provides additional quantitative information on the variability of frit chemical analyses at two commercial laboratories. Identical samples of IDMS Frit 202 were chemically analyzed at two commercial laboratories and at three different times over a period of four months. The SRL-ADS analyses, after correction with the reference standard and normalization, provided confirmatory information, but did not detect the low silica level in one of the frit samples. A methodology utilizing elliptical limits for confirming the certificate of conformance or confirmatory analysis was introduced and recommended for use when the analysis values are close but not within the specification limits. It was also suggested that the lithia specification limits might be reduced as long as CELS is used to confirm the analysis

Saltstone processing startup at the Savannah River Plant

International Nuclear Information System (INIS)

Wilhite, E.L.; Langton, C.A.; Sturm, H.F.; Hooker, R.L.; Occhipinti, E.S.

1988-01-01

High-level nuclear wastes are stored in large underground tanks at the Savannah River Plant. Processing of this waste in preparation for ultimate disposal will begin in 1988. The waste will be processed to separate the high-level radioactive fraction from the low-level radioactive fraction. The separation will be made in existing waste tanks by a process combining precipitation, adsorption, and filtration. The high-level fraction will be vitrified into borosilicate glass in the Defense Waste Processing Facility (DWPF) for permanent disposal in a federal repository. The low-level fraction (decontaminated salt solution) will be mixed with a cementitious slag-flyash blend. The resulting wasteform, saltstone, will be disposed of onsite by emplacement in an engineered facility. Waste properties, disposal facility details, and wasteform characteristics are discussed. In particular, details of saltstone processing, focusing on experience obtained from facility startup, are presented

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.

Experimental study of glass sampling devices

International Nuclear Information System (INIS)

Jouan, A.; Moncouyoux, J.P.; Meyere, A.

1992-01-01

Two high-level liquid waste containment glass sampling systems have been designed and built. The first device fits entirely inside a standard glass storage canister, and may thus be used in facilities not initially designed for this function. It has been tested successfully in the nonradioactive prototype unit at Marcoule. The work primarily covered the design and construction of an articulated arm supporting the sampling vessel, and the mechanisms necessary for filling the vessel and recovering the sample. System actuation and operation are fully automatic, and the resulting sample is representative of the glass melt. Implementation of the device is delicate however, and its reliability is estimated at about 75%. A second device was designed specifically for new vitrification facilities. It is installed directly on the glass melting furnace, and meets process operating and quality control requirements. Tests conducted at the Marcoule prototype vitrification facility demonstrated the feasibility of the system. Special attention was given to the sampling vessel transfer mechanisms, with two filling and controlled sample cooling options

ANL Technical Support Program for DOE Environmental Restoration and Waste Management. Annual report, October 1992--September 1993

International Nuclear Information System (INIS)

Bates, J.K.; Bourcier, W.L.; Bradley, C.R.

1994-06-01

This report is an overview of the progress during FY 1993 for the Technical Support Program that is part of the ANL Technology Support Activity for DOE Environmental Restoration and Waste Management (EM). The purpose is to evaluate, before hot start-up of the Defense Waste Processing Facility (DWPF) and the West Valley Demonstration Project (WVDP), factors that are anticipated to affect glass reaction in an unsaturated environment typical of what may be expected for the candidate Yucca Mountain repository site. Specific goals for the testing program include the following: reviewing and evaluating available data on parameters that will be important in establishing the long-term performance of glass in a repository environment; performing tests to further quantify the effects of important variables where there are deficiencies in the available data; and initiating long-term tests to determine glass performance under a range of conditions applicable to repository disposal

Glass formulation requirements for Hanford coupled operations using crystalline silicotitanates (CST)

International Nuclear Information System (INIS)

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

1997-01-01

The U.S. Department of Energy (DOE) through the Richland Operations Office has requested proposals from the private sector for the treatment of waste from the Hanford Waste Tanks. Phase I of this privatization initiative may include a demonstration for treatment and immobilization of both low activity and high-level waste. If the demonstration includes high-level waste, then the Cs-137 waste stream most likely will be combined with the high-level waste sludge to produce a coupled feed for immobilization (most likely vitrification using a borosilicate glass). It appears that pretreatment will involve the removal of cesium (and perhaps strontium and some transuranic radionuclides) from the supernate using an ion exchange material such as crystalline silicotitanate (CST). The ion exchange sorbent (or the eluted Cs-137) can then be combined with the sludge and vitrified in a coupled operation similar to the DWPF process. Alternatively, the cesium-loaded ion exchange sorbent can be vitrified directly to produce a separate glass waste form. SRTC has been involved in an Office of Science and Technology (EM-50) funded project to determine if Cs-137 loaded CST can be successfully incorporated into glass at significant levels. 1 For a waste form which would include only Cs-137 loaded CST, concentrations up to 60 wt% of CST in glass have been achieved. 2 The glass produced from this demonstration is both processable and durable. This CST-only waste form could be used at Hanford if the cesium-loaded CST is vitrified in a separate melter. For coupled feed operations, the CST would be mixed with high-level radioactive sludge from the Hanford tanks. This report provides the basis and the path forward for SRTC's efforts at developing a glass frit formulation which will incorporate both Hanford sludge and cesium-loaded CST for a coupled flowsheet. The goal of this work is to demonstrate the feasibility of vitrification as a method for immobilization of coupled feed (specifically

A truck cask design for shipping defense high-level waste

International Nuclear Information System (INIS)

Madsen, M.M.; Zimmer, A.

1985-01-01

The Defense High-Level Waste (DHLW) cask is a Type B packaging currently under development by the U.S. Department of Energy (DOE). This truck cask has been designed to initially transport borosilicate glass waste from the Defense Waste Processing Facility (DWPF) to the Waste Isolation Pilot Plant (WIPP). Specific program activities include designing, testing, certifying, and fabricating a prototype legal-weight truck cask system. The design includes such state-of-the-art features as integral impact limiters and remote handling features. A replaceable shielding liner provides the flexibility for shipping a wide range of waste types and activity levels

Large-scale continuous process to vitrify nuclear defense waste: operating experience with nonradioactive waste

International Nuclear Information System (INIS)

Cosper, M.B.; Randall, C.T.; Traverso, G.M.

1982-01-01

The developmental program underway at SRL has demonstrated the vitrification process proposed for the sludge processing facility of the DWPF on a large scale. DWPF design criteria for production rate, equipment lifetime, and operability have all been met. The expected authorization and construction of the DWPF will result in the safe and permanent immobilization of a major quantity of existing high level waste. 11 figures, 4 tables

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.'

Comparison of SRP high-level waste disposal costs for borosilicate glass and crystalline ceramic waste forms

International Nuclear Information System (INIS)

McDonell, W.R.

1982-04-01

An evaluation of costs for the immobilization and repository disposal of SRP high-level wastes indicates that the borosilicate glass waste form is less costly than the crystalline ceramic waste form. The wastes were assumed immobilized as glass with 28% waste loading in 10,300 reference 24-in.-diameter canisters or as crystalline ceramic with 65% waste loading in either 3400 24-in.-diameter canisters or 5900 18-in.-diameter canisters. After an interim period of onsite storage, the canisters would be transported to the federal repository for burial. Total costs in undiscounted 1981 dollars of the waste disposal operations, excluding salt processing for which costs are not yet well defined, were about $2500 million for the borosilicate glass form in reference 24-in.-diameter canisters, compared to about $2900 million for the crystalline ceramic form in 24-in.-diameter canisters and about $3100 million for the crystalline ceramic form in 18-in.-diameter canisters. No large differences in salt processing costs for the borosilicate glass and crystalline ceramic forms are expected. Discounting to present values, because of a projected 2-year delay in startup of the DWPF for the crystalline ceramic form, preserved the overall cost advantage of the borosilicate glass form. The waste immobilization operations for the glass form were much less costly than for the crystalline ceramic form. The waste disposal operations, in contrast, were less costly for the crystalline ceramic form, due to fewer canisters requiring disposal; however, this advantage was not sufficient to offset the higher development and processing costs of the crystalline ceramic form. Changes in proposed Nuclear Regulatory Commission regulations to permit lower cost repository packages for defense high-level wastes would decrease the waste disposal costs of the more numerous borosilicate glass forms relative to the crystalline ceramic forms

Defense Waste Processing Facility: Savannah River Plant, Aiken, SC. Final environmental impact statement

International Nuclear Information System (INIS)

1982-02-01

The purpose of this Environmental Impact Statement (EIS) is to provide environmental input into both the selection of an appropriate strategy for the permanent disposal of the high-level radioactive waste (HLW) currently stored at the Savannah River Plant (SRP) and the subsequent decision to construct and operate a Defense Waste Processing Facility (DWPF) at the SRP site. The SRP is a major US Department of Envgy (DOE) installation for the production of nuclear materials for national defense. Approximately 83 x 10 3 m 3 (22 million gal) of HLW currently are stored in tanks at the SRP site. The proposed DWPF would process the liquid HLW generated by SRP operations into a stable form for ultimate disposal. This EIS assesses the effects of the proposed immobilization project on land use, air quality, water quality, ecological systems, health risk, cultural resources, endangered species, wetlands protection, resource depletion, and regional social and economic systems. The radiological and nonradiological risks of transporting the immobilized wastes are assessed. The environmental impacts of disposal alternatives have recently been evaluated in a previous EIS and are therefore only summarized in this EIS

Solidification of Savannah River Plant high level waste

International Nuclear Information System (INIS)

Maher, R.; Shafranek, L.F.; Kelley, J.A.; Zeyfang, R.W.

1981-11-01

Authorization for construction of the Defense Waste Processing Facility (DWPF) is expected in FY 83. The optimum time for stage 2 authorization is about three years later. Detailed design and construction will require approximately five years for stage 1, with stage 2 construction completed about two to three years later. Production of canisters of waste glass would begin in 1988, and the existing backlog of high level waste sludge stored at SRP would be worked off by about the year 2000. Stage 2 operation could begin in 1990. The technology and engineering are ready for construction and eventual operation of the DWPF for immobilizing high level radioactive waste at Savannah River Plant (SRP). Proceeding with this project will provide the public, and the leadership of this country, with a crucial demonstration that a major quantity of existing high level nuclear wastes can be safely and permanently immobilized. Early demonstration will both expedite and facilitate rational decision making on this aspect of the nuclear program. Delay in providing these facilities will result in significant DOE expenditures at SRP for new tanks just for continued temporary storage of wastes, and would probably result in dissipation of the intellectual and planning momentum that has built up in developing the project

Silicate glasses. Chapter 1

International Nuclear Information System (INIS)

Lutze, W.

1988-01-01

This chapter is a survey of world-wide research and development efforts in nuclear waste glasses and its production technology. The principal glasses considered are silicate glasses which contain boron, i.e. borosilicate glass. A historical overview of waste form development programs in nine countries is followed by a summary of the design criteria for borosilicate glass compositions glass compositions. In the sections on glass properties the waste form is characterized in terms of potential alterations under the influence of heat, thermal gradients, radiation, aqueous solutions and combinations thereof. The topics are phase transformations, mechanical properties, radiation effects and chemical durability. The results from studies of volcanic glasses, as natural analogues for borosilicate nuclear waste glasses in order to verify predictions obtained from short-term tests in the laboratory, have been compiled in a special section on natural analogues. A special section on advanced vitrification techniques summarizes the various actual and potential processing schemes and describes the facilities. The literature has been considered until 1985. (author). 430 refs.; 68 figs.; 29 tabs

Characterization of the R7T7 LWR reference glass

International Nuclear Information System (INIS)

Pacaud, F.; Fillet, C.; Baudin, G.; Bastien-Thiry, H.

1990-01-01

Characterization describes the glass properties by means of standard tests with no attempt to assess its long-term behavior. Characterization involved complementary comparative investigations of nonradioactive laboratory glass specimens, radioactive glass specimens prepared in laboratory hot cells, and nonradioactive industrial glass samples fabricated in the full-scale continuous vitrification prototype facility (specimens were taken from the casting stream and core-samples were taken from a 200 kg glass block after cooling in the canister). Additional measurements are planned on actual radioactive glass samples fabricated in the R7 facility at La Hague. The results are indicated for each of the properties studied: physical, thermal and mechanical properties; structure and homogeneity examination; thermal stability and crystallization; resistance to chemical corrosion; irradiation resistance and volatilization. Comparative examination of glass samples of different origins showed consistent properties

Proposed Use of a Constructed Wetland for the Treatment of Metals in the S-04 Outfall of the Defense Waste Processing Facility at the Savannah River Site

International Nuclear Information System (INIS)

Glover, T.

1999-01-01

The DWPF is part of an integrated waste treatment system at the SRS to treat wastes containing radioactive contaminants. In the early 1980s the DOE recognized that there would be significant safety and cost advantages associated with immobilizing the radioactive waste in a stable solid form. The Defense Waste Processing Facility was designed and constructed to accomplish this task

Proposed Use of a Constructed Wetland for the Treatment of Metals in the S-04 Outfall of the Defense Waste Processing Facility at the Savannah River Site

Energy Technology Data Exchange (ETDEWEB)

Glover, T.

1999-11-23

The DWPF is part of an integrated waste treatment system at the SRS to treat wastes containing radioactive contaminants. In the early 1980s the DOE recognized that there would be significant safety and cost advantages associated with immobilizing the radioactive waste in a stable solid form. The Defense Waste Processing Facility was designed and constructed to accomplish this task.

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

Statistical experimental design for saltstone mixtures

International Nuclear Information System (INIS)

Harris, S.P.; Postles, R.L.

1992-01-01

The authors used a mixture experimental design for determining a window of operability for a process at the U.S. Department of Energy, Savannah River Site, Defense Waste Processing Facility (DWPF). The high-level radioactive waste at the Savannah River Site is stored in large underground carbon steel tanks. The waste consists of a supernate layer and a sludge layer. Cesium-137 will be removed from the supernate by precipitation and filtration. After further processing, the supernate layer will be fixed as a grout for disposal in concrete vaults. The remaining precipitate will be processed at the DWPF with treated waste tank sludge and glass-making chemicals into borosilicate glass. The leach-rate properties of the supernate grout formed from various mixes of solidified coefficients for NO 3 and chromium were used as a measure of leach rate. Various mixes of cement, Ca(OH) 2 , salt, slag, and fly ash were used. These constituents comprise the whole mix. Thus, a mixture experimental design was used. The regression procedure (PROC REG) in SAS was used to produce analysis of variance (ANOVA) statistics. In addition, detailed model diagnostics are readily available for identifying suspicious observations. For convenience, trillinear contour (TLC) plots, a standard graphics tool for examining mixture response surfaces, of the fitted model were produced using ECHIP

Separation of aromatic precipitates from simulated high level radioactive waste by hydrolysis, evaporation and liquid-liquid extraction

International Nuclear Information System (INIS)

Young, S.R.; Shah, H.B.; Carter, J.T.

1991-01-01

The Defense Waste Processing Facility (DWPF) at the SRS will be the United States' first facility to process High Level radioactive Waste (HLW) into a borosilicate glass matrix. The removal of aromatic precipitates by hydrolysis, evaporation and liquid-liquid extraction will be a key step in the processing of the HLW. This step, titled the Precipitate Hydrolysis Process, has been demonstrated by the Savannah River Laboratory with the Precipitate Hydrolysis Experimental Facility (PHEF). The mission of the PHEF is to demonstrate processing of simulated high level radioactive waste which contains tetraphenylborate precipitates and nitrite. Reduction of nitrite by hydroxylamine nitrate and hydrolysis of the tetraphenylborate by formic acid is discussed. Gaseous production, which is primarily benzene, nitrous oxide and carbon dioxide, has been quantified. Production of high-boiling organic compounds and the accumulation of these organic compounds within the process are addressed

Flow measurement and control in the defense waste process

International Nuclear Information System (INIS)

Heckendorn, F.M. II.

1985-01-01

The Defense Waste Processing Facility (DWPF) for immobilizing Savannah River Plant (SRP) high-level radioactive waste is now under construction. Previously stored waste is retrieved and processed into a glass matrix for permanent storage. The equipment operates in an entirely remote environment for both processing and maintenance due to the highly radioactive nature of the waste. A fine powdered glass frit is mixed with the waste prior to its introduction as a slurry into an electric glass furnace. The slurry is Bingham plastic in nature and of high viscosity. This combination of factors has created significant problems in flow measurement and control. Specialized pieces of equipment have been demonstrated that will function properly in a highly abrasive environment while receiving no maintenance during their lifetime. Included are flow meters, flow control technology, flow switching, and remote connections. No plastics or elastomers are allowed in contact with fluids and all electronic components are mounted remotely. Both two- and three-way valves are used. Maintenance is by crane replacement of process sections, utilizing specialized connectors. All portions of the above are now operating full scale (radioactively cold) at the test facility at SRP. 4 references, 8 figures

ANL Technical Support Program for DOE Environmental Restoration and Waste Management

International Nuclear Information System (INIS)

Bates, J.K.; Bradley, C.R.; Buck, E.C.; Cunnane, J.C.; Dietz, N.L.; Ebert, W.L.; Emery, J.W.; Feng, X.; Gerding, T.J.; Gong, M.; Hoh, J.C.; Mazer, J.J.; Wronkiewicz, D.J.; Bourcier, W.L.; Morgan, L.E.; Nielsen, J.K.; Steward, S.A.; Ewing, R.C.; Wang, L.M.; Han, W.T.; Tomozawa, M.

1992-03-01

This report provides an overview of progress during FY 1991 for the Technical Support Program that is part of the ANL Technology Support Activity for DOE, Environmental Restoration and Waste Management (EM). The purpose is to evaluate, before hot start-up of the Defenses Waste Processing Facility (DWPF) and the West Valley Demonstration Project (WVDP), factors that are likely to affect glass reaction in an unsaturated environment typical of what may be expected for the candidate Yucca Mountain repository site. Specific goals for the testing program include the following: (1) to review and evaluate available information on parameters that will be important in establishing the long-term performance of glass in a repository environment; (2) to perform testing to further quantify the effects of important variables where there are deficiencies in the available data; and (3) to initiate long-term testing that will bound glass performance under a range of conditions applicable to repository disposal

The potential impacts of sodium management on Frit Development for Coupled Operations

Energy Technology Data Exchange (ETDEWEB)

Johnson, F. C. [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); Peeler, D. K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-06-10

In this report, Section 2.0 provides a description of sodium management and its impact on the glass waste form, Section 3.0 provides background information on phase separation, Section 4.0 provides the impact of sodium management on SB9 frit development efforts and the results of a limited scoping study investigating phase separation in potential DWPF frits, and Section 5.0 discusses potential technical issues associated with using a phase separated frit for DWPF operations.

Development of the DWPF canister temporary shrink-fit seal

International Nuclear Information System (INIS)

Kelker, J.W. Jr.

1986-04-01

The Defense Waste Processing Facility is being constructed at The Savannah River Plant for the containerization of high-level nuclear waste in a wasteform for eventual permanent disposal. The waste will be incorporated in molten glass and solidified in type 304L stainless steel canisters, 2-feet in diameter x 9-feet 10-inches long, containing a flanged 6-in.-diam pipe fill-nozzle. The canisters have a minimum wall thickness of 3/8 in. Utilizing the heat from the glass filling operation, a shrink-fit seal for a plug in the end of the canister fill nozzle was developed that: will withstand the radioactive environment; will prevent the spread of contamination, and will keep moisture and water from entering the canister during storage and decontamination of the canister by wet-frit blasting to remove smearable and oxide-film fixed radioactive nuclides; is removable and can be replaced by a new oversize plug in the event the seal fails the pressure decay leakage test ( -4 atm cc/sec helium); will keep the final weld closure clean and free of nuclear contamination; will withstand being pressed into the nozzle without exposing external contamination or completely breaking the seal; is reliable; and is easily installed. The seal consists of: a removable sleeve (with a tapered bore) which is shrink-fitted into the nozzle bore during canister fabrication; and a tapered plug which is placed into the sleeved nozzle after the canister is filled with radioactive molten glass. A leak-tight shrink-fit seal is formed between the nozzle, sleeve, and plug upon temperature equilibrium. The temporarily sealed canister is transferred from the Melt cell to the Decon cell, and the surface is decontaminated. Next it is transferred to the Weld/Test cell where the temporary seal is pressed down into the nozzle, revealing a clean cavity where the canister final closure weld is made

Advanced High-Level Waste Glass Research and Development Plan

Energy Technology Data Exchange (ETDEWEB)

Peeler, David K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vienna, John D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schweiger, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fox, Kevin M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-07-01

The U.S. Department of Energy Office of River Protection (ORP) has implemented an integrated program to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. The integrated ORP program is focused on providing a technical, science-based foundation from which key decisions can be made regarding the successful operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) facilities. The fundamental data stemming from this program will support development of advanced glass formulations, key process control models, and tactical processing strategies to ensure safe and successful operations for both the low-activity waste (LAW) and high-level waste (HLW) vitrification facilities with an appreciation toward reducing overall mission life. The purpose of this advanced HLW glass research and development plan is to identify the near-, mid-, and longer-term research and development activities required to develop and validate advanced HLW glasses and their associated models to support facility operations at WTP, including both direct feed and full pretreatment flowsheets. This plan also integrates technical support of facility operations and waste qualification activities to show the interdependence of these activities with the advanced waste glass (AWG) program to support the full WTP mission. Figure ES-1 shows these key ORP programmatic activities and their interfaces with both WTP facility operations and qualification needs. The plan is a living document that will be updated to reflect key advancements and mission strategy changes. The research outlined here is motivated by the potential for substantial economic benefits (e.g., significant increases in waste throughput and reductions in glass volumes) that will be realized when advancements in glass formulation continue and models supporting facility operations are implemented. Developing and applying advanced

Remote crane control techniques and closed-circuit television for the U.S. Department of Energy, Defense Waste Processing Facility

International Nuclear Information System (INIS)

DaSilva, D.A.

1988-01-01

The Defense Waste Processing Facility (DWPF) located at the Savannah River Plant (SRP), South Carolina is a nuclear waste facility being built to vitrify and containerize high level radioactive waste products. DWPF has a unique requirement for an unmanned crane system to install and replace equipment in the high humidity, high radiation and harsh chemical environment of permanently inaccessible processing cells. A radio control system is provided to control a 117 ton capacity bridge crane that is equipped with various power tools for remote handling of crane replaceable and maintained equipment. High resolution black and white closed circuit television (CTV) assemblies mounted on the crane and on the walls of the various processing cells are provided for viewing the equipment during normal operations and maintenance. The main process cell (MPC) crane is designed as the vehicle to be the eyes, ears and hands for remote in-cell operations and maintenance. The crane runs on elevated rails above the process cells. A trailer-like-electrical equipment compartment (EEC) containing control and radio communications equipment for the crane; is dragged along on rails in a heavily shielded corridor by a drag bar mounted to the crane. A two way RF system is the communications link for all control and video signals between the crane and two stationary crane control consoles

Actinide, Elemental, and Fission Product Measurements by ICPMS at the Savannah River Site

International Nuclear Information System (INIS)

Tovo, L.L.; Waller, P.R.; Clymire, J.; Jones, V.D.; Boyce, W.T.

1998-03-01

VG Elemental Inductively coupled plasma-mass spectrometer (ICPMS), PlasmaQuad 1 (PQ1) Model No. 4, installed in a radiohood, is used by the Savannah River Technology Center to provide non-routine mass measurements for environmental monitoring, waste tank characterization studies, isotope ratios for criticality determinations, and the measurement of elemental, fission product, and actinide mass distributions of the glass product from the Defense Waste Processing Facility (DWPF). Modifications to improve instrument reliability, sample preparation, and data handling, as well as modifications to the laboratory that permit measurements in a radioactive environment will be discussed. Based on our operating experience, two laboratory facilities are being prepared for additional instruments to operate in a radioactive environment. A separate instrument is being installed for non-radioactive measurements and method development

Sol-Gel Glasses

Science.gov (United States)

Mukherjee, S. P.

1985-01-01

Multicomponent homogeneous, ultrapure noncrystalline gels/gel derived glasses are promising batch materials for the containerless glass melting experiments in microgravity. Hence, ultrapure, homogeneous gel precursors could be used to: (1) investigate the effect of the container induced nucleation on the glass forming ability of marginally glass forming compositions; and (2) investigate the influence of gravity on the phase separation and coarsening behavior of gel derived glasses in the liquid-liquid immiscibility zone of the nonsilicate systems having a high density phase. The structure and crystallization behavior of gels in the SiO2-GeO2 as a function of gel chemistry and thermal treatment were investigated. As are the chemical principles involved in the distribution of a second network former in silica gel matrix being investigated. The procedures for synthesizing noncrystalline gels/gel-monoliths in the SiO2-GeO2, GeO2-PbO systems were developed. Preliminary investigations on the levitation and thermal treatment of germania silicate gel-monoliths in the Pressure Facility Acoustic Levitator were done.

Impact of Salt Waste Processing Facility Streams on the Nitric-Glycolic Flowsheet in the Chemical Processing Cell

Energy Technology Data Exchange (ETDEWEB)

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

2017-08-08

An evaluation of the previous Chemical Processing Cell (CPC) testing was performed to determine whether the planned concurrent operation, or “coupled” operations, of the Defense Waste Processing Facility (DWPF) with the Salt Waste Processing Facility (SWPF) has been adequately covered. Tests with the nitricglycolic acid flowsheet, which were both coupled and uncoupled with salt waste streams, included several tests that required extended boiling times. This report provides the evaluation of previous testing and the testing recommendation requested by Savannah River Remediation. The focus of the evaluation was impact on flammability in CPC vessels (i.e., hydrogen generation rate, SWPF solvent components, antifoam degradation products) and processing impacts (i.e., acid window, melter feed target, rheological properties, antifoam requirements, and chemical composition).

INTERNATIONAL STUDY OF ALUMINUM IMPACTS ON CRYSTALLIZATION IN U.S. HIGH LEVEL WASTE GLASS

International Nuclear Information System (INIS)

Fox, K; David Peeler, D; Tommy Edwards, T; David Best, D; Irene Reamer, I; Phyllis Workman, P; James Marra, J

2008-01-01

The objective of this task was to develop glass formulations for (Department of Energy) DOE waste streams with high aluminum concentrations to avoid nepheline formation while maintaining or meeting waste loading and/or waste throughput expectations as well as satisfying critical process and product performance related constraints. Liquidus temperatures and crystallization behavior were carefully characterized to support model development for higher waste loading glasses. The experimental work, characterization, and data interpretation necessary to meet these objectives were performed among three partnering laboratories: the V.G. Khlopin Radium Institute (KRI), Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL). Projected glass compositional regions that bound anticipated Defense Waste Processing Facility (DWPF) and Hanford high level waste (HLW) glass regions of interest were developed and used to generate glass compositions of interest for meeting the objectives of this study. A thorough statistical analysis was employed to allow for a wide range of waste glass compositions to be examined while minimizing the number of glasses that had to be fabricated and characterized in the laboratory. The glass compositions were divided into two sets, with 45 in the test matrix investigated by the U.S. laboratories and 30 in the test matrix investigated by KRI. Fabrication and characterization of the US and KRI-series glasses were generally handled separately. This report focuses mainly on the US-series glasses. Glasses were fabricated and characterized by SRNL and PNNL. Crystalline phases were identified by X-ray diffraction (XRD) in the quenched and canister centerline cooled (CCC) glasses and were generally iron oxides and spinels, which are not expected to impact durability of the glass. Nepheline was detected in five of the glasses after the CCC heat treatment. Chemical composition measurements for each of the glasses were conducted

INTERNATIONAL STUDY OF ALUMINUM IMPACTS ON CRYSTALLIZATION IN U.S. HIGH LEVEL WASTE GLASS

Energy Technology Data Exchange (ETDEWEB)

Fox, K; David Peeler, D; Tommy Edwards, T; David Best, D; Irene Reamer, I; Phyllis Workman, P; James Marra, J

2008-09-23

The objective of this task was to develop glass formulations for (Department of Energy) DOE waste streams with high aluminum concentrations to avoid nepheline formation while maintaining or meeting waste loading and/or waste throughput expectations as well as satisfying critical process and product performance related constraints. Liquidus temperatures and crystallization behavior were carefully characterized to support model development for higher waste loading glasses. The experimental work, characterization, and data interpretation necessary to meet these objectives were performed among three partnering laboratories: the V.G. Khlopin Radium Institute (KRI), Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL). Projected glass compositional regions that bound anticipated Defense Waste Processing Facility (DWPF) and Hanford high level waste (HLW) glass regions of interest were developed and used to generate glass compositions of interest for meeting the objectives of this study. A thorough statistical analysis was employed to allow for a wide range of waste glass compositions to be examined while minimizing the number of glasses that had to be fabricated and characterized in the laboratory. The glass compositions were divided into two sets, with 45 in the test matrix investigated by the U.S. laboratories and 30 in the test matrix investigated by KRI. Fabrication and characterization of the US and KRI-series glasses were generally handled separately. This report focuses mainly on the US-series glasses. Glasses were fabricated and characterized by SRNL and PNNL. Crystalline phases were identified by X-ray diffraction (XRD) in the quenched and canister centerline cooled (CCC) glasses and were generally iron oxides and spinels, which are not expected to impact durability of the glass. Nepheline was detected in five of the glasses after the CCC heat treatment. Chemical composition measurements for each of the glasses were conducted

Dynamic fatigue testing of Zerodur glass-ceramic

Science.gov (United States)

Tucker, Dennis S.

1988-01-01

The inherent brittleness of glass invariably leads to a large variability in strength data and a time dependence in strength. Leading rate plays a large role in strength values. Glass is found to be weaker when supporting loads over long periods of time as compared to glass which undergoes rapid leading. These properties complicate the structural design allowables for the utilization of glass components in an application such as Advanced X-ray Astrophysics Facility (AXAF). The test methodology to obtain parameters which can be used to predict the reliability and life time of Zerodur glass-ceramic which is to be used for the mirrors in the AXAF is described.

Sludge Batch 7B Qualification Activities With SRS Tank Farm Sludge

International Nuclear Information System (INIS)

Pareizs, J.; Click, D.; Lambert, D.; Reboul, S.

2011-01-01

Waste Solidification Engineering (WSE) has requested that characterization and a radioactive demonstration of the next batch of sludge slurry - Sludge Batch 7b (SB7b) - 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 typically simulates the Tank Farm and DWPF processes with a Tank 51 sample (referred to as the qualification sample). With the tight schedule constraints for SB7b and the potential need for caustic addition to allow for an acceptable glass processing window, the qualification for SB7b was approached differently than past batches. For SB7b, SRNL prepared a Tank 51 and a Tank 40 sample for qualification. SRNL did not receive the qualification sample from Tank 51 nor did it simulate all of the Tank Farm washing and decanting operations. Instead, SRNL prepared a Tank 51 SB7b sample from samples of Tank 7 and Tank 51, along with a wash solution to adjust the supernatant composition to the final SB7b Tank 51 Tank Farm projections. SRNL then prepared a sample to represent SB7b in Tank 40 by combining portions of the SRNL-prepared Tank 51 SB7b sample and a Tank 40 Sludge Batch 7a (SB7a) sample. The blended sample was 71% Tank 40 (SB7a) and 29% Tank 7/Tank 51 on an insoluble solids basis. This sample is referred to as the SB7b Qualification Sample. The blend represented the highest projected Tank 40 heel (as of May 25, 2011), and thus, the highest

Neural network analysis of nuclear waste glass composition vs durability

International Nuclear Information System (INIS)

Seibel, C.K.

1994-01-01

The relationship between the chemical composition of oxide glasses and their physical properties is poorly understood, but it is becoming more important as vitrification (transformation into glass) of high-level nuclear waste becomes the favored method for long-term storage. The vitrified waste will be stored deep in geologic repositories where it must remain intact for at least 10,000 years. A strong resistance to groundwater exposure; i.c. a slow rate of glass dissolution, is of great importance. This project deals specifically with glass samples developed and tested for the nuclear fuel reprocessing facility near West Valley, New York. This facility needs to dispose of approximately 2.2 million liters of high-level radioactive liquid waste currently stored in stainless steel tanks. A self-organizing, artificial neural network was used to analyze the trends in the glass dissolution data for the effects of composition and the resulting durability of borosilicate glasses in an aqueous environment. This durability data can be used to systematically optimize the properties of the complex nuclear glasses and slow the dissolution rate of radionuclides into the environment

TECHNOLOGY DEMONSTRATION OF SLUDGE MASS REDUCTION VIA ALUMINUM DISSOLUTION: GLASS FORMULATION PROCESSING WINDOW PREDICTIONS FOR SB5

International Nuclear Information System (INIS)

Fox, K.; Tommy Edwards, T.; David Peeler, D.

2007-01-01

Composition projections for Sludge Batch 5 (SB5) were developed, based on a modeling approach at the Savannah River National Laboratory (SRNL), to evaluate possible impacts of the Al-dissolution process on the availability of viable frit compositions for vitrification at the Defense Waste Processing Facility (DWPF). The study included two projected SB5 compositions that bound potential outcomes (or degrees of effectiveness) of the Al-dissolution process, as well as a nominal SB5 composition projection based on the results of the recent Al-dissolution demonstration at SRNL. The three SB5 projections were the focus of a two-stage paper study assessment. A Nominal Stage assessment combined each of the SB5 composition projections with an array of 19,305 frit compositions over a wide range of waste loading (WL) values and evaluated them against the DWPF process control models. The Nominal Stage results allowed for the down-selection of a small number of frits that provided reasonable projected operating windows (typically 27 to 42 wt% WL). The frit/sludge systems were mostly limited by process related constraints, with only one system being limited by predictions of nepheline crystallization, a waste form affecting constraint. The criteria applied in selecting the frit compositions somewhat restricted the compositional flexibility of the candidate frits for each individual SB5 composition projection, which may limit the ability to further tailor the frit for improved melt rate. Variation Stage assessments were then performed using the down-selected frits and the three SB5 composition projections with variation applied to each sludge component. The Variation Stage results showed that the operating windows were reduced in width, as expected when variation in the sludge composition is applied. However, several of the down-selected frits exhibited a relatively high degree of robustness to the applied sludge variation, providing WL windows of approximately 30 to 39 wt%. The

Calculation of radionuclides in the defense waste processing facility

International Nuclear Information System (INIS)

Chandler, J.R.; Finch, D.R.; Becker, G.W. Jr.

1979-01-01

SHIELD system calculations yield the isotopic inventory, activity, decay heat, and multigroup radiation source spectra for all of the DWPF process streams and for the solidified waste products. One application of these results is the analysis of the radiation emissions of the stored waste. Another application is the analysis of time dependent properties of the solidified waste. Initially, gamma radiation from /sup 137m/Ba decay contributes approximately one-third of the total energy. As the 137 Cs content decreases, the gamma contribution declines. The major producers of beta radiation are the 90 Sr, 137 Cs, and 144 Pr decay chains. As the glass age increases, however, the contribution from the actinides dominates increasingly. The inital activity level in the glass is 2000 curies per gallon. The activity and decay heat decrease by a factor of 2 in about fifteen years, and by a factor of 4 in fifty years. A similar analysis was made for the salt cake. Initially, the salt cake produces 0.01 watts per gallon from 2.4 curies per gallon of activity. In five years, the activity is reduced by a factor of 19, and the decay heat declines by a factor of 24. After ten years, both the activity and decay heat levels are less than 1% of their initial values. 7 figures, 4 tables

Materials Characterization Facility

Data.gov (United States)

Federal Laboratory Consortium — The Materials Characterization Facility enables detailed measurements of the properties of ceramics, polymers, glasses, and composites. It features instrumentation...

Annual report to the Laser Facility Committee 1979

International Nuclear Information System (INIS)

1979-03-01

The report covers the work done at the Central Laser Facility, Rutherford Laboratory during the year preceding 31 March 1979. Preliminary work already undertaken on the upgrade of the glass laser and target areas consisting of the relocation of the two beam target chamber and tests on phosphate glass and also the completion of the electron beam generator for use by researchers on high power gas laser systems, are described. Work of the groups using the glass laser facility are considered under the headings; glass laser development, gas laser development, laser plasma interactions, transport and particle emission, ablative compression studies, atomic and radiation physics, XUV lasers, theory and computation. (U.K.)

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

BNFL Report Glass Formers Characterization

Energy Technology Data Exchange (ETDEWEB)

Schumacher, R.F.

2000-07-27

The objective of this task was to obtain powder property data on candidate glass former materials, sufficient to guide conceptual design and estimate the cost of glass former handling facilities as requested under Part B1 of BNFL Technical and Development Support. Twenty-nine glass forming materials were selected and obtained from vendors for the characterization of their physical properties, durability in caustic solution, and powder flow characteristics. A glass former was selected based on the characterization for each of the ten oxide classes required for Envelope A, B, and C mixtures. Three blends (A, B, and C) were prepared based on formulations provided by Vitreous State Laboratory and evaluated with the same methods employed for the glass formers. The properties obtained are presented in a series of attached Tables. It was determined that five of the ten glass formers, (kyanite, iron oxide, titania, zircon, and zinc oxide) have the potential to cause some level of solids f low problems. The problems might include arching or ratholing in the silo/hopper. In addition, all of the blends may require consideration for their handling.

BNFL Report Glass Formers Characterization

International Nuclear Information System (INIS)

Schumacher, R.F.

2000-01-01

The objective of this task was to obtain powder property data on candidate glass former materials, sufficient to guide conceptual design and estimate the cost of glass former handling facilities as requested under Part B1 of BNFL Technical and Development Support. Twenty-nine glass forming materials were selected and obtained from vendors for the characterization of their physical properties, durability in caustic solution, and powder flow characteristics. A glass former was selected based on the characterization for each of the ten oxide classes required for Envelope A, B, and C mixtures. Three blends (A, B, and C) were prepared based on formulations provided by Vitreous State Laboratory and evaluated with the same methods employed for the glass formers. The properties obtained are presented in a series of attached Tables. It was determined that five of the ten glass formers, (kyanite, iron oxide, titania, zircon, and zinc oxide) have the potential to cause some level of solids f low problems. The problems might include arching or ratholing in the silo/hopper. In addition, all of the blends may require consideration for their handling

Crystallization in high-level waste glass: A review of glass theory and noteworthy literature

Energy Technology Data Exchange (ETDEWEB)

Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-08-18

There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-level waste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO₄) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe, Cr)₂O₄), while not detrimental to glass durability, can cause an array of processing problems inside HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic, thermodynamic, and viscosity arguments, and several highlights of glass-crystallization research, as it pertains to high-level waste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies. Higher waste loadings and more efficient processing strategies will reduce the overall HLW Hanford Tank Waste Treatment and Immobilization Plant (WTP) vitrification facilities mission life.

Vitrification facility at the West Valley Demonstration Project

International Nuclear Information System (INIS)

DesCamp, V.A.; McMahon, C.L.

1996-07-01

This report is a description of the West Valley Demonstration Project's vitrification facilities from the establishment of the West Valley, NY site as a federal and state cooperative project to the completion of all activities necessary to begin solidification of radioactive waste into glass by vitrification. Topics discussed in this report include the Project's background, high-level radioactive waste consolidation, vitrification process and component testing, facilities design and construction, waste/glass recipe development, integrated facility testing, and readiness activities for radioactive waste processing

Science and Engineering Research Council Central Laser Facility

International Nuclear Information System (INIS)

1981-03-01

This report covers the work done at, or in association with, the Central Laser Facility during the year April 1980 to March 1981. In the first chapter the major reconstruction and upgrade of the glass laser, which has been undertaken in order to increase the versatility of the facility, is described. The work of the six groups of the Glass Laser Scientific Progamme and Scheduling Committee is described in further chapters entitled; glass laser development, laser plasma interactions, transport and particle emission studies, ablative acceleration and compression studies, spectroscopy and XUV lasers, and theory and computation. Publications based on the work of the facility which have either appeared or been accepted for publication during the year are listed. (U.K.)

Turning nuclear waste into glass

Energy Technology Data Exchange (ETDEWEB)

Pegg, Ian L.

2015-02-15

Vitrification has emerged as the treatment option of choice for the most dangerous radioactive waste. But dealing with the nuclear waste legacy of the Cold War will require state-of-the-art facilities and advanced glass formulations.

REFINEMENT OF THE NEPHELINE DISCRIMINATOR: RESULTS OF A PHASE I STUDY

Energy Technology Data Exchange (ETDEWEB)

Fox, K; James Newell, J; Tommy Edwards, T; David Best, D; Irene Reamer, I; Phyllis Workman, P

2008-02-13

The performance of a glass used for immobilization of high-level nuclear waste (HLW) is generally quantified by its resistance to chemical degradation, or durability. The durability of a HLW glass is dependent on its composition. If crystalline phases form within a glass during cooling, the composition of the residual glass network is altered, therefore affecting the durability of the glass. Crystallization of nepheline (NaAlSiO{sub 4}) has been shown to adversely impact the durability of HLW glasses since it removes glass forming species (in this case, Al and Si) from the glass network. The propensity for nepheline crystallization in a HLW glass increases with increasing concentrations of Al{sub 2}O{sub 3} and Na{sub 2}O in the glass. Nepheline crystallization is therefore of concern for processing of HLW at the Defense Waste Processing Facility (DWPF) since the sludge waste streams at the Savannah River Site (SRS) can contain high concentrations of Al{sub 2}O{sub 3} and Na{sub 2}O. Currently, a 'nepheline discriminator' is included as a process control constraint at the DWPF. The nepheline discriminator relates the concentrations of SiO{sub 2}, Na{sub 2}O and Al{sub 2}O{sub 3} (as weight percentages in glass) to a critical value of 0.62. The discriminator defines a boundary line on the SiO{sub 2}-Na{sub 2}O-Al{sub 2}O{sub 3} ternary diagram above which (or toward the SiO{sub 2} corner of the ternary) nepheline is not predicted to crystallize in the glass upon quenching or slow cooling. The current equation uses only the concentrations of the SiO{sub 2}, Na{sub 2}O and Al{sub 2}O{sub 3} components in the glass in predicting whether or not nepheline is likely to crystallize. However, several other components have been shown to impact the propensity for nepheline crystallization, including B{sub 2}O{sub 3} and CaO among others. Therefore, the potential exists to further refine the nepheline discriminator to include these components. In addition, recently

Late washing filter cleaning cycle demonstration

International Nuclear Information System (INIS)

Meyer, M.L.; McCabe, D.J.

1992-01-01

The DWPF Late Washing Facility will filter cesium and potassium tetraphenyl borate (TPB) solids using a Mott sintered metal filter, identical to the filter now used in the In-tank Precipitation Facility. The purpose of the late wash step is primarily to remove the nitrite salts from the slurry prior to delivery to DWPF. Periodic chemical cleaning of the filter will be required, presumably after each batch although the actual required frequency could not be determined on the lab-scale. Minimization of chemical cleaning solution volumes is key to maximizing the attainment of the Late Wash facility. This report summarizes work completed in experiments designed to identify minimum cleaning solution requirements

Defense Waste Processing Facility Canister Closure Weld Current Validation Testing

Energy Technology Data Exchange (ETDEWEB)

Korinko, P. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Maxwell, D. N. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2018-01-29

Two closure welds on filled Defense Waste Processing Facility (DWPF) canisters failed to be within the acceptance criteria in the DWPF operating procedure SW4-15.80-2.3 (1). In one case, the weld heat setting was inadvertently provided to the canister at the value used for test welds (i.e., 72%) and this oversight produced a weld at a current of nominally 210 kA compared to the operating procedure range (i.e., 82%) of 240 kA to 263 kA. The second weld appeared to experience an instrumentation and data acquisition upset. The current for this weld was reported as 191 kA. Review of the data from the Data Acquisition System (DAS) indicated that three of the four current legs were reading the expected values, approximately 62 kA each, and the fourth leg read zero current. Since there is no feasible way by further examination of the process data to ascertain if this weld was actually welded at either the target current or the lower current, a test plan was executed to provide assurance that these Nonconforming Welds (NCWs) meet the requirements for strength and leak tightness. Acceptance of the welds is based on evaluation of Test Nozzle Welds (TNW) made specifically for comparison. The TNW were nondestructively and destructively evaluated for plug height, heat tint, ultrasonic testing (UT) for bond length and ultrasonic volumetric examination for weld defects, burst pressure, fractography, and metallography. The testing was conducted in agreement with a Task Technical and Quality Assurance Plan (TTQAP) (2) and applicable procedures.

The impact of the MCU life extension solvent on sludge batch 8 projected operating windows

International Nuclear Information System (INIS)

Peeler, D.K.; Edwards, T.B.

2013-01-01

As a part of the Actinide Removal Process (ARP)/Modular Caustic Side Solvent Extraction Unit (MCU) Life Extension Project, a next generation solvent (NGS) and a new strip acid 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 or evaluations with the next generation solvent are required to determine the impact of these changes (if any) to Chemical Process Cell (CPC) activities, glass formulation strategies, and melter operations at the Defense Waste Processing Facility (DWPF). The introduction of the dilute (0.01 M) boric acid stream into the DWPF flowsheet has a potential impact on glass formulation and frit development efforts since B203 is a major oxide in frits developed for DWPF. Prior knowledge of this stream can be accounted for during frit development efforts but that was not the case for Sludge Batch 8 (SB8). Frit 803 has already been recommended and procured for SB8 processing; altering the frit to account for the incoming boron from the strip effluent (SE) is not an option for SB8. Therefore, the operational robustness of Frit 803 to the introduction of SE including its compositional tolerances (i.e., up to 0.0125M boric acid) is of interest and was the focus of this study. The primary question to be addressed in the current study was: What is the impact (if any) on the projected operating windows for the Frit 803 - SB8 flowsheet to additions of B203 from the SE in the Sludge Receipt and Adjustment Tank (SRAT)? More specifically, will Frit 803 be robust to the potential compositional changes occurring in the SRAT due to sludge variation, varying additions of ARP and/or the introduction of SE by providing access to waste loadings (WLs) of interest to DWPF? The Measurement Acceptability Region (MAR) results indicate there is very little, if any, impact on the projected operating windows for the Frit 803 - SB8 system regardless of the presence or absence of

Characteristics of radiophotoluminescent glass dosimeters

Energy Technology Data Exchange (ETDEWEB)

Ito, Masashi; Shiraishi, Akemi; Murakami, Hiroyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-07-01

In Japan Atomic Energy Research Institute, a film badge is recently replaced by a new type radiophotoluminescent (RPL) glass dosimeter for external personal monitoring. Some fundamental characteristics of this dosimeter, such as dose dependence linearity, energy dependence, angular dependence, dose evaluation accuracy at mixed irradiation conditions, fading, etc., were examined at the Facility of Radiation Standard (FRS), JAERI. The results have proved that the RPL glass dosimeter has sufficient characteristics for practical use as a personal dosimeter for all of the items examined. (author)

Preliminary analysis of projected construction employment effects of building the defense waste processing facility at the Savannah River Plant

International Nuclear Information System (INIS)

Garey, R.B.; Blair, L.M.; Craig, R.L.; Stevenson, W.

1981-09-01

This study estimates the probable effects of constructing the DWPF on the surrounding labor markets. Analyses are based on data from the local and regional labor markets, information from experts on local construction activities, information on the labor requirements of the Vogtle Power Plant (two nuclear reactors) being built by Georgia Power Company in Burke County, Georgia, and an econometric model of the construction labor market, based on several surveys of workers building three Tennessee Valley Authority nuclear power plants. The results of this study are reported in three parts. In Part I, completed in May 1980, we describe the 1979 (base year) employment levels within the local and regional labor markets surrounding SRP, from which most DWPF construction workers are likely to be drawn. In Part II, completed in June 1980, we define the four local sources of construction employment that will compete for craftsmen when DWPF is built. Also in Part II, most of the projected impacts of the DWPF reference immobilization alternative (one of several alternatives that may be chosen) are reported. Several construction schedules and labor demand scenarios for the reference alternative are considered. In Part III, completed in January 1981, most of the estimated impacts of the DWPF alternative referred to as the staged process alternative are reported. Several construction schedules and labor demand scenarios for this alternative are considered

Low expansion and high gain Nd laser glasses

International Nuclear Information System (INIS)

Izumitani, Tetsuro; Peng, B.

1995-01-01

Due to the relationship between Judd-Ofelt intensity parameter and covalency, new laser glasses have been developed which have low expansion coefficients (85--91 x 10 -7 /cm C, 0--70 C) and high emission cross sections. They have good chemical properties, high Young's modulus and high thermal conductivities. These glasses are suitable for the National Ignition Facility

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

PNL vitrification technology development project glass formulation strategy for LLW vitrification

International Nuclear Information System (INIS)

Kim, D.; Hrma, P.R.; Westsik, J.H. Jr.

1996-03-01

This Glass Formulation Strategy describes development approaches to optimize glass compositions for Hanford's low-level waste vitrification between now and the projected low-level waste facility start-up in 2005. The objectives of the glass formulation task are to develop optimized glass compositions with satisfactory long-term durability, acceptable processing characteristics, adequate flexibility to handle waste variations, maximize waste loading to practical limits, and to develop methodology to respond to further waste variations

ANL technical support program for DOE Office of Environmental Management. Annual report, October 1994--September 1995

International Nuclear Information System (INIS)

Bates, J.K.; Buck, E.C.; Dietz, N.L.; DiSanto, T.; Ebert, W.L.

1996-07-01

A program was established for the DOE Office of Environmental Management (EM) to evaluate factors that are anticipated to affect waste glass reaction during repository disposal, especially in an unsaturated environment typical of what may be expected for the proposed Yucca Mountain repository site. This report covers progress in FY 1995 on the following tasks: (1) Tests are ongoing to evaluate and compare the reactivity of fully radioactive glasses with that of glasses having the same compositions except for the absence of radionuclides under conditions representative of a high-level waste repository environment. Data from these tests will be used to evaluate the effect of radionuclides on the glass corrosion behavior and to determine the disposition of the radionuclides as the glass corrodes. Static dissolution tests and unsaturated tests are being conducted with several Defense Waste Processing Facility (DWPF) and West Valley Demonstration Project (WVDP) glasses. (2) A series of static dissolution tests is being performed to compare the corrosion behavior of nuclear waste glasses made with SRL 131 and SRL 202 frits at different S/V ratios. The S/V ratio affects the extent to which dissolved glass species are diluted; the solution chemistry then affects continued glass dissolution. The solutions generated in tests at high S/V ratios are conducive to the formation of alteration phases that may be deleterious to the glass. After long time periods, the glass dissolution rates of both glasses increase coincidentally with the formation of analcime and other alteration phases. However, the release of radionuclides from the glasses into solution is controlled by their individual solubilities

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

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

Actual waste demonstration of the nitric-glycolic flowsheet for sludge batch 9 qualification

Energy Technology Data Exchange (ETDEWEB)

Newell, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, J. [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); Reboul, S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Coleman, 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); Johnson, F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-03-09

For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) performs qualification testing to demonstrate that the sludge batch is processable. Based on the results of this actual-waste qualification and previous simulant studies, SRNL recommends implementation of the nitric-glycolic acid flowsheet in DWPF. Other recommendations resulting from this demonstration are reported in section 5.0.

Ventilation criteria for IDMS facility

International Nuclear Information System (INIS)

Lambert, D.P.

1996-01-01

Both Facility Evaluation Board (FEB) reviews of the Integrated DWPF Melter System (IDMS) have identified the inconsistency of the current IDMS Process Hazards Review (PHR) versus actual IDMS practice as regards the criteria to contain air borne pollutants that may be present in the Process Room (e.g. benzene and mercury). The PHR states that a 1.0 in. wc pressure differential be maintained between the IDMS Process Room and Building 672-T. In addition, the PHR further specifies that the linear velocity through openings into the Process Room (e.g. open doors) be equal to or greater than 150 fpm. Finally, the PHR recommended that mercury vapor and benzene monitors be installed in the Process Room ventilation exhaust to alert personnel to the presence of vapors of benzene and/or mercury before entering the Process Room. This report summarizes the results of reassessment of these criteria and the specific recommendation for permanent installation of mercury and benzene vapor monitors in the vapor exhaust of the Process Room

Prediction of waste glass melt rates

International Nuclear Information System (INIS)

Lee, L.

1987-01-01

Under contract to the Department of Energy, the Du Pont Company has begun construction of a Defense Waste Processing Facility to immobilize radioactive wastes now stored as liquids at the Department of Energy's Savannah River Plant. The immobilization process solidifies waste sludge by vitrification into a leach-resistant borosilicate glass. Development of this process has been the responsibility of the Savannah River Laboratory. As part of the development, a simple model was developed to predict the melt rates for the waste glass melter. This model is based on an energy balance for the cold cap and gives very good agreement with melt rate data obtained from experimental campaigns in smaller scale waste glass melters

High-level waste borosilicate glass a compendium of corrosion characteristics. Volume 1

International Nuclear Information System (INIS)

Cunnane, J.C.

1994-03-01

Current plans call for the United States Department of Energy (DOE) to start up facilities for vitrification of high-level radioactive waste (HLW) stored in tanks at the Savannah River Site, Aiken, South Carolina, in 1995; West Valley Demonstration Project, West Valley, New York, in 1996; and at the Hanford Site, Richland, Washington, after the year 2000. The product from these facilities will be canistered HLW borosilicate glass, which will be stored, transported, and eventually disposed of in a geologic repository. The behavior of this glass waste product, under the range of likely service conditions, is the subject of considerable scientific and public interest. Over the past few decades, a large body of scientific information on borosilicate waste glass has been generated worldwide. The intent of this document is to consolidate information pertaining to our current understanding of waste glass corrosion behavior and radionuclide release. The objective, scope, and organization of the document are discussed in Section 1.1, and an overview of borosilicate glass corrosion is provided in Section 1.2. The history of glass as a waste form and the international experience with waste glass are summarized in Sections 1.3 and 1.4, respectively

Office of River Protection Advanced Low-Activity Waste Glass Research and Development Plan

Energy Technology Data Exchange (ETDEWEB)

Peeler, David K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kim, Dong-Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vienna, John D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schweiger, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Piepel, Gregory F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-11-01

The U.S. Department of Energy Office of River Protection (ORP) has initiated and leads an integrated Advanced Waste Glass (AWG) program to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product performance requirements. The integrated ORP program is focused on providing a technical, science-based foundation for making key decisions regarding the successful operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) facilities in the context of an optimized River Protection Project (RPP) flowsheet. The fundamental data stemming from this program will support development of advanced glass formulations, key product performance and process control models, and tactical processing strategies to ensure safe and successful operations for both the low-activity waste (LAW) and high-level waste vitrification facilities. These activities will be conducted with the objective of improving the overall RPP mission by enhancing flexibility and reducing cost and schedule. The purpose of this advanced LAW glass research and development plan is to identify the near-term, mid-term, and longer-term research and development activities required to develop and validate advanced LAW glasses, property-composition models and their uncertainties, and an advanced glass algorithm to support WTP facility operations, including both Direct Feed LAW and full pretreatment flowsheets. Data are needed to develop, validate, and implement 1) new glass property-composition models and 2) a new glass formulation algorithm. Hence, this plan integrates specific studies associated with increasing the Na2O and SO3/halide concentrations in glass, because these components will ultimately dictate waste loadings for LAW vitrification. Of equal importance is the development of an efficient and economic strategy for 99Tc management. Specific and detailed studies are being implemented to understand the fate of Tc throughout

Release of ammonia from HAN-type PHA

International Nuclear Information System (INIS)

Zamecnik, J.R.

1992-01-01

A preliminary design basis for ammonia scrubbers in the DWPF has been issued. This design basis is based on a theoretical model of ammonia evolution from the SRAT, SME and RCT. It is desirable to acquire actual process data on ammonia evolution prior to performing detailed design of scrubbers for DWPF. The evolution of ammonia from the SRAT and SME in the Integrated DWPF Melter System (IDMS) was investigated during the HM4 run. In this run, Precipitate Hydrolysis Aqueous (PHA), which was made in the Precipitate Hydrolysis Experimental Facility (PHEF) using the HAN (hydroxylamine nitrate) process was used, thus resulting in PHA with a high concentration of ammonium ion

Liquid level measurement in high level nuclear waste slurries

International Nuclear Information System (INIS)

Weeks, G.E.; Heckendorn, F.M.; Postles, R.L.

1990-01-01

Accurate liquid level measurement has been a difficult problem to solve for the Defense Waste Processing Facility (DWPF). The nuclear waste sludge tends to plug or degrade most commercially available liquid-level measurement sensors. A liquid-level measurement system that meets demanding accuracy requirements for the DWPF has been developed. The system uses a pneumatic 1:1 pressure repeater as a sensor and a computerized error correction system. 2 figs

ILAW Glass Testing for Disposal at IDF: Phase 1 Testing

Energy Technology Data Exchange (ETDEWEB)

Papathanassiu, Adonia [The Catholic Univ. of America, Washington, DC (United States). Virteous State Lab.; Muller, Isabelle S. [The Catholic Univ. of America, Washington, DC (United States). Virteous State Lab.; Brandys, Marek [The Catholic Univ. of America, Washington, DC (United States). Virteous State Lab.; Gilbo, Konstantin [The Catholic Univ. of America, Washington, DC (United States). Virteous State Lab.; Barkatt, Aaron [The Catholic Univ. of America, Washington, DC (United States). Virteous State Lab.; Joseph, Innocent [EnergySolutions Federal EPC, Inc., Columbia, MD (United States); The Catholic Univ. of America, Washington, DC (United States). Virteous State Lab.; Pegg, Ian L. [The Catholic Univ. of America, Washington, DC (United States). Virteous State Lab.; Brown, Elvie E. [Washington River Protection Solutions, LLC, Richland, WA (United States); Swanberg, David J. [Washington River Protection Solutions, LLC, Richland, WA (United States)

2011-04-11

This document reports the results of the testing of phase 1 ORP LAW (low activity waste) glasses, also identified as enhanced LAW glasses. Testing involved are SPFT (Single Pass Flow Through), VHT (Vapor Hydration Test), and PCT (Product Consistency Test), along with the analytical tests (XRD and SEM-EDS). This report contains the data of the high waste loading ORP LAW glasses that will be used for the performance assessment of the IDF (Integrated Disposal Facility).

ILAW Glass Testing for Disposal at IDF: Phase 1 Testing

International Nuclear Information System (INIS)

Papathanassiu, Adonia; Swanberg, David J.

2011-01-01

This document reports the results of the testing of phase 1 ORP LAW (low activity waste) glasses, also identified as enhanced LAW glasses. Testing involved are SPFT (Single Pass Flow Through), VHT (Vapor Hydration Test), and PCT (Product Consistency Test), along with the analytical tests (XRD and SEM-EDS). This report contains the data of the high waste loading ORP LAW glasses that will be used for the performance assessment of the IDF (Integrated Disposal Facility).

Product consistency leach tests of Savannah River Site radioactive waste glasses

International Nuclear Information System (INIS)

Bibler, N.E.; Bates, J.K.

1990-01-01

The product consistency test (PCT) is a glass leach test developed at the Savannah River Site (SRS) to confirm the durability of radioactive nuclear waste glasses that will be produced in the Defense Waste Processing Facility. The PCT is a seven day, crushed glass leach test in deionized water at 90C. Final leachates are filtered and acidified prior to analysis. To demonstrate the reproducibility of the PCT when performed remotely, SRS and Argonne National Laboratory have performed the PCT on samples of two radioactive glasses. The tests were also performed to compare the releases of the radionuclides with the major nonradioactive glass components and to determine if radiation from the glass was affecting the results of the PCT. The test was performed in triplicate at each laboratory. For the major soluble elements, B, Li, Na, and Si, in the glass, each investigator obtained relative precisions in the range 2-5% in the triplicate tests. This range indicates good precision for the PCT when performed remotely with master slave manipulators in a shielded cell environment. When the results of the two laboratories were compared to each other, the agreement was within 20%. Normalized concentrations for the nonradioactive and radioactive elements in the PCT leachates measured at both facilities indicated that the radionuclides were released from the glass slower than the major soluble elements in the glass. For both laboratories, the normalized releases for both glasses were in the general order Li ∼ B ∼ Na > Si > Cs - 137 > Sb - 125 < Sr - 90. The normalized releases for the major soluble elements and the final pH values in the tests with radioactive glass are consistent with those for nonradioactive glasses with similar compositions. This indicates that there was no significant effect of radiation on the results of the PCT

Glass packages in interim storage

International Nuclear Information System (INIS)

Jacquet-Francillon, N.

1994-10-01

This report summarize the current state of knowledge concerning the behavior of type C waste packages consisting of vitrified high-level solutions produced by reprocessing spent fuel. The composition and the physical and chemical properties of the feed solutions are reviewed, and the vitrification process is described. Sodium alumino-borosilicate glass compositions are generally employed - the glass used at la Hague for LWR fuel solutions, for example, contains 45 % SiO 2 . The major physical, chemical, mechanical and thermal properties of the glass are reviewed. In order to allow their thermal power to diminish, the 3630 glass packages produced (as of January 1993) in the vitrification facilities at Marcoule and La Hague are placed in interim storage for several decades. The actual interim storage period has not been defined, as it is closely related to the concept and organization selected for the final destination of the packages: a geological repository. The glass behavior under irradiation is described. Considerable basic and applied research has been conducted to assess the aqueous leaching behavior of nuclear containment glass. The effects of various repository parameters (temperature, flow rate, nature of the environmental materials) have been investigated. The experimental findings have been used to specify a model describing the kinetics of aqueous corrosion of the glass. More generally all the ''source term'' models developed in France by the CEA or by ANDRA are summarized. (author). 152 refs., 33 figs

Development and characterization of single gap glass RPC

Energy Technology Data Exchange (ETDEWEB)

Manisha, E-mail: manisha@pu.ac.in; Bhatnagar, V.; Shahi, J.S.; Singh, J.B.

2016-12-21

India-based Neutrino Observatory (INO) facility is going to have a 50 kton magnetized Iron CALorimeter (ICAL) detector for precision measurements of neutrino oscillations using atmospheric neutrinos. The proposed ICAL detector will be a stack of magnetized iron plates (acting as target material) interleaved with glass Resistive Plate Chambers (RPCs) as the active detector elements. An RPC is a gaseous detector made up of two parallel electrode plates having high bulk resistivity like that of a float glass and bakelite. For the ICAL detector, glass is preferred over bakelite as it does not need any kind of surface treatment to achieve better surface uniformity and also the cost of associated electronics is reduced. Under the detector R&D efforts for the proposed glass RPC detector, a few glass RPCs of 1 m × 1 m dimension are fabricated procuring glass of ∼2 mm thickness from one of the Indian glass manufacturers (Asahi). In the present paper, we report the characterization of RPC based on leakage current, muon detection efficiency and noise rate studies with varying gas compositions.

BNFL Report Glass Formers Characterization

Energy Technology Data Exchange (ETDEWEB)

Schumacher, R.F.

2000-07-27

The objective of this task was to obtain powder property data on candidate glass former materials, sufficient to guide conceptual design and estimate the cost of glass former handling facilities as requested under Part B1 of BNFL Technical and Development Support. Twenty-nine glass forming materials were selected and obtained from vendors for the characterization of their physical properties, durability in caustic solution, and powder flow characteristics. A glass former was selected based on the characterization for each of the ten oxide classes required for Envelope A, B, and C mixtures. Three blends (A, B, and C) were prepared based on formulations provided by Vitreous State Laboratory and evaluated with the same methods employed for the glass formers. The properties obtained are presented in a series of attached Tables. It was determined that five of the ten glass formers, (kyanite, iron oxide, titania, zircon, and zinc oxide) have the potential to cause some level of solids f low problems. In addition, all of the blends may require consideration for their handling. A number of engineering considerations and recommendations were prepared based on the experimental findings, experience, and other process considerations. Recommendations for future testing are included. In conjunction with future work, it is recommended that a professional consultant be engaged to guide and assist with testing and design input.

BNFL Report Glass Formers Characterization

International Nuclear Information System (INIS)

Schumacher, R.F.

2000-01-01

The objective of this task was to obtain powder property data on candidate glass former materials, sufficient to guide conceptual design and estimate the cost of glass former handling facilities as requested under Part B1 of BNFL Technical and Development Support. Twenty-nine glass forming materials were selected and obtained from vendors for the characterization of their physical properties, durability in caustic solution, and powder flow characteristics. A glass former was selected based on the characterization for each of the ten oxide classes required for Envelope A, B, and C mixtures. Three blends (A, B, and C) were prepared based on formulations provided by Vitreous State Laboratory and evaluated with the same methods employed for the glass formers. The properties obtained are presented in a series of attached Tables. It was determined that five of the ten glass formers, (kyanite, iron oxide, titania, zircon, and zinc oxide) have the potential to cause some level of solids f low problems. In addition, all of the blends may require consideration for their handling. A number of engineering considerations and recommendations were prepared based on the experimental findings, experience, and other process considerations. Recommendations for future testing are included. In conjunction with future work, it is recommended that a professional consultant be engaged to guide and assist with testing and design input

Letter report: Evaluation of LFCM off-gas system technologies for the HWVP

Energy Technology Data Exchange (ETDEWEB)

Goles, R.W.; Mishima, J.; Schmidt, A.J.

1996-03-01

Radioactive high-level liquid waste (HLLW), a byproduct of defense nuclear fuel reprocessing activities, is currently being stored in underground tanks at several US sites. Because its mobility poses significant environmental risks, HLLW is not a suitable waste form for long-term storage. Thus, high-temperature processes for solidifying and isolating the radioactive components of HLLW have been developed and demonstrated by the US Department of Energy (DOE) and its contractors. Vitrification using liquidfed ceramic melters (LFCMs) is the reference process for converting US HLLW into a borosilicate glass. Two vitrification plants are currently under construction in the United States: the West Valley Demonstration Plant (WVDP) being built at the former West Valley Nuclear Fuels Services site in West Valley, New York; and the Defense Waste Processing Facility (DWPF), which is currently 85% complete at DOE`s Savannah River Plant (SRP). A third facility, the Hanford Waste Vitrification Plant (HWVP), is being designed at DOE`s Hanford Site.

Letter report: Evaluation of LFCM off-gas system technologies for the HWVP

International Nuclear Information System (INIS)

Goles, R.W.; Mishima, J.; Schmidt, A.J.

1996-03-01

Radioactive high-level liquid waste (HLLW), a byproduct of defense nuclear fuel reprocessing activities, is currently being stored in underground tanks at several US sites. Because its mobility poses significant environmental risks, HLLW is not a suitable waste form for long-term storage. Thus, high-temperature processes for solidifying and isolating the radioactive components of HLLW have been developed and demonstrated by the US Department of Energy (DOE) and its contractors. Vitrification using liquidfed ceramic melters (LFCMs) is the reference process for converting US HLLW into a borosilicate glass. Two vitrification plants are currently under construction in the United States: the West Valley Demonstration Plant (WVDP) being built at the former West Valley Nuclear Fuels Services site in West Valley, New York; and the Defense Waste Processing Facility (DWPF), which is currently 85% complete at DOE's Savannah River Plant (SRP). A third facility, the Hanford Waste Vitrification Plant (HWVP), is being designed at DOE's Hanford Site

Disposition of excess plutonium using ''off-spec'' MOX pellets as a sintered ceramic waste form

International Nuclear Information System (INIS)

Armantrout, G.A.; Jardine, L.J.

1996-02-01

The authors describe a potential strategy for the disposition of excess weapons plutonium in a way that minimizes (1) technological risks, (2) implementation costs and completion schedules, and (3) requirements for constructing and operating new or duplicative Pu disposition facilities. This is accomplished by an optimized combination of (1) using existing nuclear power reactors to ''burn'' relatively pure excess Pu inventories as mixed oxide (MOX) fuel and (2) using the same MOX fuel fabrication facilities to fabricate contaminated or impure excess Pu inventories into an ''off-spec'' MOX solid ceramic waste form for geologic disposition. Diversion protection for the SCWF to meet the ''spent fuel standard'' introduced by the National Academy of Sciences can be achieved in at least three ways. (1) One can utilize the radiation field from defense high-level nuclear waste by first packaging the SCWF pellets in 2- to 4-L cans that are subsequently encapsulated in radioactive glass in the Defense Waste Processing Facility (DWPF) glass canisters (a ''can-in-canister'' approach). (2) One can add 137 Cs (recovered from defense wastes at Hanford and currently stored as CsCl in capsules) to an encapsulating matrix such as cement for the SCWF pellets in a small hot-cell facility and thus fabricate large monolithic forms. (3) The SCWF can be fabricated into reactor fuel-like pellets and placed in tubes similar to fuel assemblies, which can then be mixed in sealed repository containers with irradiated spent nuclear fuel for geologic disposition

Anomalies in the Thermophysical Properties of Undercooled Glass-Forming Alloys

Science.gov (United States)

Hyers, Robert W.; Rogers, Jan R.; Kelton, Kenneth F.; Gangopadhyay, Anup

2008-01-01

The surface tension, viscosity, and density of several bulk metallic glass-forming alloys have been measured using noncontact techniques in the electrostatic levitation facility (ESL) at NASA Marshall Space Flight Center. All three properties show unexpected behavior in the undercooled regime. Similar deviations were previously observed in titanium-based quasicrystal-forming alloys,but the deviations in the properties of the glass-forming alloys are much more pronounced. New results for anomalous thermophysical properties in undercooled glass-forming alloys will be presented and discussed.

Literature Review On Impact Of Glycolate On The 2H Evaporator And The Effluent Treatment Facility

International Nuclear Information System (INIS)

Adu-Wusu, K.

2012-01-01

Glycolic acid (GA) is being studied as an alternate reductant in the Defense Waste Processing Facility (DWPF) feed preparation process. It will either be a total or partial replacement for the formic acid that is currently used. A literature review has been conducted on the impact of glycolate on two post-DWPF downstream systems - the 2H Evaporator system and the Effluent Treatment Facility (ETF). The DWPF recycle stream serves as a portion of the feed to the 2H Evaporator. Glycolate enters the evaporator system from the glycolate in the recycle stream. The overhead (i.e., condensed phase) from the 2H Evaporator serves as a portion of the feed to the ETF. The literature search revealed that virtually no impact is anticipated for the 2H Evaporator. Glycolate may help reduce scale formation in the evaporator due to its high complexing ability. The drawback of the solubilizing ability is the potential impact on the criticality analysis of the 2H Evaporator system. It is recommended that at least a theoretical evaluation to confirm the finding that no self-propagating violent reactions with nitrate/nitrites will occur should be performed. Similarly, identification of sources of ignition relevant to glycolate and/or update of the composite flammability analysis to reflect the effects from the glycolate additions for the 2H Evaporator system are in order. An evaluation of the 2H Evaporator criticality analysis is also needed. A determination of the amount or fraction of the glycolate in the evaporator overhead is critical to more accurately assess its impact on the ETF. Hence, use of predictive models like OLI Environmental Simulation Package Software (OLI/ESP) and/or testing are recommended for the determination of the glycolate concentration in the overhead. The impact on the ETF depends on the concentration of glycolate in the ETF feed. The impact is classified as minor for feed glycolate concentrations (le) 33 mg/L or 0.44 mM. The ETF unit operations that will have

LITERATURE REVIEW ON IMPACT OF GLYCOLATE ON THE 2H EVAPORATOR AND THE EFFLUENT TREATMENT FACILITY

Energy Technology Data Exchange (ETDEWEB)

Adu-Wusu, K.

2012-05-10

Glycolic acid (GA) is being studied as an alternate reductant in the Defense Waste Processing Facility (DWPF) feed preparation process. It will either be a total or partial replacement for the formic acid that is currently used. A literature review has been conducted on the impact of glycolate on two post-DWPF downstream systems - the 2H Evaporator system and the Effluent Treatment Facility (ETF). The DWPF recycle stream serves as a portion of the feed to the 2H Evaporator. Glycolate enters the evaporator system from the glycolate in the recycle stream. The overhead (i.e., condensed phase) from the 2H Evaporator serves as a portion of the feed to the ETF. The literature search revealed that virtually no impact is anticipated for the 2H Evaporator. Glycolate may help reduce scale formation in the evaporator due to its high complexing ability. The drawback of the solubilizing ability is the potential impact on the criticality analysis of the 2H Evaporator system. It is recommended that at least a theoretical evaluation to confirm the finding that no self-propagating violent reactions with nitrate/nitrites will occur should be performed. Similarly, identification of sources of ignition relevant to glycolate and/or update of the composite flammability analysis to reflect the effects from the glycolate additions for the 2H Evaporator system are in order. An evaluation of the 2H Evaporator criticality analysis is also needed. A determination of the amount or fraction of the glycolate in the evaporator overhead is critical to more accurately assess its impact on the ETF. Hence, use of predictive models like OLI Environmental Simulation Package Software (OLI/ESP) and/or testing are recommended for the determination of the glycolate concentration in the overhead. The impact on the ETF depends on the concentration of glycolate in the ETF feed. The impact is classified as minor for feed glycolate concentrations {le} 33 mg/L or 0.44 mM. The ETF unit operations that will have

Annual report to the Laser Facility Committee 1984

International Nuclear Information System (INIS)

1984-01-01

The report describes the work carried out at, or in association with, the Central Laser Facility (CLF), during the year ending March 1984. The CLF programme is divided into three main sections. The first, the glass laser scientific programme, is concerned with applications of the high power Nd glass laser. The second, the ultra violet radiation facility scientific programme, involves the excimer pumped frequency tunable lasers. The last, high power KrF laser development, describes Research and development work on this laser. (U.K.)

Enhanced LAW Glass Correlation - Phase 1

Energy Technology Data Exchange (ETDEWEB)

Muller, Isabelle S. [The Catholic Univ. of America, Washington, DC (United States). Vitreous State Lab.; Matlack, Keith S. [The Catholic Univ. of America, Washington, DC (United States). Vitreous State Lab.; Pegg, Ian L. [The Catholic Univ. of America, Washington, DC (United States). Vitreous State Lab.; Joseph, Innocent [Atkins Energy Federal EPC, Inc., Columbia, MD (United States)

2016-12-01

About 50 million gallons of high-level mixed waste is currently stored in underground tanks at the United States Department of Energy’s (DOE’s) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE’s Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal. The tank waste will be separated into low- and high-activity waste fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High Level Waste (IHLW) products. The ILAW product will be disposed in an engineered facility on the Hanford site while the IHLW product is designed for acceptance into a national deep geological disposal facility for high-level nuclear waste. The ILAW and IHLW products must meet a variety of requirements with respect to protection of the environment before they can be accepted for disposal. Acceptable glass formulations for vitrification of Hanford low activity waste (LAW) must meet a variety of product quality, processability, and waste loading requirements. To this end, The Vitreous State Laboratory (VSL) at The Catholic University of America (CUA) developed and tested a number of glass formulations during Part A, Part B1 and Part B2 of the WTP development program. The testing resulted in the selection of target glass compositions for the processing of eight of the Phase I LAW tanks. The selected glass compositions were tested at the crucible scale to confirm their compliance with ILAW performance requirements. Duramelter 100 (DM100) and LAW Pilot Melter tests were then conducted to demonstrate the viability of these glass compositions for LAW vitrification at high processing rates.

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

HLW immobilization in glass

International Nuclear Information System (INIS)

Leroy, P.; Jacquet-Francillon, N.; Runge, S.

1992-01-01

The immobilization of High Level Waste in glass in France is a long history which started as early as in the 1950's. More than 30 years of Research and Development have been invested in that field. Two industrial facilities are operating (AVM and R7) and a third one (T7), under cold testing, is planned to start active operation in the mid-92. While vitrification has been demonstrated to be an industrially mastered process, the question of the quality of the final waste product, i.e. the HLW glass, must be addressed. The scope of the present paper is to focus on the latter point from both standpoints of the R and D and of the industrial reality

A Method for Cobalt and Cesium Leaching from Glass Fiber in HEPA Filter

International Nuclear Information System (INIS)

Kim, Gye Nam; Lee, Suk Chol; Yang, Hee Chul; Yoon, In Ho; Choi, Wang Kyu; Moon, Jei Kwon

2011-01-01

A great amount of radioactive waste has been generated during the operation of nuclear facilities. Recently, the storage space of a radioactive waste storage facility in the Korea Atomic Energy Research Institute (KAERI) was almost saturated with many radioactive wastes. So, the present is a point of time that a volume reduction of the wastes in a radioactive waste storage facility needs. There are spent HEPA filter wastes of about 2,226 sets in the radioactive waste storage facility in KAERI. All these spent filter wastes have been stored in accordance with their original form without any treatment. Up to now a compression treatment of these spent HEPA filters has been carried out to repack the compressed spent HEPA filters into a 200 liter drum for their volume reduction. Frame and separator are contaminated with a low concentration of nuclide, while the glass fiber is contaminated with a high concentration of nuclide. So, for the disposal of the glass filter to the environment, the glass fiber should be leached to lower its radioactive concentration first and then must be stabilized by solidification and so on. Therefore, it is necessary to develop a leaching process of glass fiber in a HEPA filter. Leaching is a separation technology, which is often used to remove a metal or a nuclide from a solid mixture with the help of a liquid solvent

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