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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  2. 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

  3. 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

  4. Translating DWPF design criteria into an engineered facility design

    International Nuclear Information System (INIS)

    Kemp, J.B.

    1986-01-01

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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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.

  11. 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.

  12. 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

  13. 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

  14. 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.

  15. 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

  16. 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

  17. Glass sampling program during DWPF Integrated Cold Runs

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  19. 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

  20. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-18

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

  2. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-30

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  5. 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

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

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1992-01-01

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

  7. 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

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-06-01

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

  15. 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

  16. 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

  17. 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)

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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.

  4. 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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  6. Characterization of the Defense Waste Processing Facility (DWPF) Environmental Assessment (EA) glass Standard Reference Material. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Crawford, C.L.; Pickett, M.A.

    1993-06-01

    Liquid high-level nuclear waste at the Savannah River Site (SRS) will be immobilized by vitrification in borosilicate glass. The glass will be produced and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). Other waste form producers, such as West Valley Nuclear Services (WVNS) and the Hanford Waste Vitrification Project (HWVP), will also immobilize high-level radioactive waste in borosilicate glass. The canistered waste will be stored temporarily at each facility for eventual permanent disposal in a geologic repository. The Department of Energy has defined a set of requirements for the canistered waste forms, the Waste Acceptance Product Specifications (WAPS). The current Waste Acceptance Primary Specification (WAPS) 1.3, the product consistency specification, requires the waste form producers to demonstrate control of the consistency of the final waste form using a crushed glass durability test, the Product Consistency Test (PCI). In order to be acceptable, a waste glass must be more durable during PCT analysis than the waste glass identified in the DWPF Environmental Assessment (EA). In order to supply all the waste form producers with the same standard benchmark glass, 1000 pounds of the EA glass was fabricated. The chemical analyses and characterization of the benchmark EA glass are reported. This material is now available to act as a durability and/or redox Standard Reference Material (SRM) for all waste form producers.

  7. 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

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-11

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  11. 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

  12. 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

  13. DWPF waste glass Product Composition Control System

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  14. 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

  15. 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

  16. 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

  17. 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

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

    International Nuclear Information System (INIS)

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

    1992-05-01

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

  19. 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.

  20. 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

  1. 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

  2. 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

  3. Remotely replaceable jumpers and embedded wiring for the DWPF

    International Nuclear Information System (INIS)

    Heckendorn, F.M. II.

    1984-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-08

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

  5. 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.

  6. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-30

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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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

  15. 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

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

    International Nuclear Information System (INIS)

    Landon, L.F.

    1992-01-01

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

  17. 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

  18. 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.

  19. The corrosion behavior of DWPF glasses

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  20. 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

  1. 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

  2. DWPF Flowsheet Studies with Simulants to Determine Modular Caustic Side Solvent Extraction Unit Solvent Partitioning and Verify Actinide Removal Process Incorporation Strategy

    International Nuclear Information System (INIS)

    Herman, C

    2006-01-01

    The Actinide Removal Process (ARP) facility and the Modular Caustic Side Solvent Extraction Unit (MCU) are scheduled to begin processing salt waste in fiscal year 2007. A portion of the streams generated in the salt processing facilities will be transferred to the Defense Waste Processing Facility (DWPF) to be incorporated in the glass matrix. Before the streams are introduced, a combination of impact analyses and research and development studies must be performed to quantify the impacts on DWPF processing. The Process Science and Engineering (PS and E) section of the Savannah River National Laboratory (SRNL) was requested via Technical Task Request (TTR) HLW/DWPF/TTR-2004-0031 to evaluate the impacts on DWPF processing. Simulant Chemical Process Cell (CPC) flowsheet studies have been performed using previous composition and projected volume estimates for the ARP sludge/monosodium titanate (MST) stream. Due to changes in the flammability control strategy for DWPF for salt processing, the incorporation strategy for ARP has changed and additional ARP flowsheet tests were necessary to validate the new processing strategy. The last round of ARP testing included the incorporation of the MCU stream and identified potential processing issues with the MCU solvent. The identified issues included the potential carry-over and accumulation of the MCU solvent components in the CPC condensers and in the recycle stream to the Tank Farm. Therefore, DWPF requested SRNL to perform additional MCU flowsheet studies to better quantify the organic distribution in the CPC vessels. The previous MCU testing used a Sludge Batch 4 (SB4) simulant since it was anticipated that both of these facilities would begin salt processing during SB4 processing. The same sludge simulant recipe was used in this round of ARP and MCU testing to minimize the number of changes between the two phases of testing so a better comparison could be made. ARP and MCU stream simulants were made for this phase of

  3. 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

  4. 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.

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  20. Liquidus Temperature Data for DWPF Glass

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  1. 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

  2. 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

  3. 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

  4. 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

  5. Phase II of a Six sigma Initiative to Study DWPF SME Analytical Turnaround Times: SRNL's Evaluation of Carbonate-Based Dissolution Methods

    International Nuclear Information System (INIS)

    Edwards, Thomas

    2005-01-01

    The Analytical Development Section (ADS) and the Statistical Consulting Section (SCS) of the Savannah River National Laboratory (SRNL) are participating in a Six Sigma initiative to improve the Defense Waste Processing Facility (DWPF) Laboratory. The Six Sigma initiative has focused on reducing the analytical turnaround time of samples from the Slurry Mix Evaporator (SME) by developing streamlined sampling and analytical methods [1]. The objective of Phase I was to evaluate the sub-sampling of a larger sample bottle and the performance of a cesium carbonate (Cs 2 CO 3 ) digestion method. Successful implementation of the Cs 2 CO 3 fusion method in the DWPF would have important time savings and convenience benefits because this single digestion would replace the dual digestion scheme now used. A single digestion scheme would result in more efficient operations in both the DWPF shielded cells and the inductively coupled plasma--atomic emission spectroscopy (ICP-AES) laboratory. By taking a small aliquot of SME slurry from a large sample bottle and dissolving the vitrified SME sample with carbonate fusion methods, an analytical turnaround time reduction from 27 hours to 9 hours could be realized in the DWPF. This analytical scheme has the potential for not only dramatically reducing turnaround times, but also streamlining operations to minimize wear and tear on critical shielded cell components that are prone to fail, including the Hydragard(trademark) sampling valves and manipulators. Favorable results from the Phase I tests [2] led to the recommendation for a Phase II effort as outlined in the DWPF Technical Task Request (TTR) [3]. There were three major tasks outlined in the TTR, and SRNL issued a Task Technical and QA Plan [4] with a corresponding set of three major task activities: (1) Compare weight percent (wt%) total solids measurements of large volume samples versus peanut vial samples. (2) Evaluate Cs 2 CO 3 and K 2 CO 3 fusion methods using DWPF simulated

  6. 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

  7. 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.

  8. 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

  9. 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

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

    International Nuclear Information System (INIS)

    Stone, M.

    2010-01-01

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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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)

  16. 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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  18. IMPACTS OF SMALL COLUMN ION EXCHANGE STREAMS ON DWPF GLASS FORMULATION: KT01, KT02, KT03, AND KT04-SERIES GLASS COMPOSITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K.; Edwards, T.

    2010-11-01

    Four series of glass compositions were selected, fabricated, and characterized as part of a study to determine the impacts of the addition of Crystalline Silicotitanate (CST) and Monosodium Titanate (MST) from the Small Column Ion Exchange (SCIX) process on the Defense Waste Processing Facility (DWPF) glass waste form and the applicability of the DWPF process control models. The KT01 and KT02-series of glasses were chosen to allow for the identification of the influence of the concentrations of major components of the glass on the retention of TiO{sub 2}. The KT03 series of glasses was chosen to allow for the identification of these influences when higher Nb{sub 2}O{sub 5} and ZrO{sub 2} concentrations are included along with TiO2. The KT04 series of glasses was chosen to investigate the properties and performance of glasses based on the best available projections of actual compositions to be processed at the DWPF (i.e., future sludge batches including the SCIX streams).

  19. 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

  20. 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).

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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)

  11. 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.

  12. 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

  13. 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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  15. 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

  16. 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

  17. 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

  18. 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)

  19. 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

  20. 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

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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)

  11. 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

  12. 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

  13. 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).

  14. 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.

  15. DWPF GLASS BEADS AND GLASS FRIT TRANSPORT DEMONSTRATION

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, D; Bradley Pickenheim, B

    2008-11-24

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

  16. DWPF GLASS BEADS AND GLASS FRIT TRANSPORT DEMONSTRATION

    International Nuclear Information System (INIS)

    Adamson, D.; Pickenheim, Bradley

    2008-01-01

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

  17. 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

  18. 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

  19. 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)

  20. 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

  1. 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.

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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)

  10. 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

  11. 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).

  12. 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

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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).

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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.

  11. 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

  12. 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

  13. 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.

  14. 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

  15. 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.

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. Miscibility Evaluation Of The Next Generation Solvent With Polymers Currently Used At DWPF, MCU, And Saltstone

    Energy Technology Data Exchange (ETDEWEB)

    Fondeur, F. F.

    2013-04-17

    The Office of Waste Processing, within the Office of Technology Innovation and Development, funded the development of an enhanced Caustic-Side Solvent Extraction (CSSX) solvent for deployment at the Savannah River Site for removal of cesium from High Level Waste. This effort lead to the development of the Next Generation Solvent (NGS) with Tris (3,7-dimethyl octyl) guanidine (TiDG). The first deployment target for the NGS solvent is within the Modular CSSX Unit (MCU). Deployment of a new chemical within an existing facility requires verification that the new chemical components are compatible with the installed equipment. In the instance of a new organic solvent, the primary focus is on compatibility of the solvent with organic polymers used in the affected facility. This report provides the calculated data from exposing these polymers to the Next Generation Solvent. An assessment of the dimensional stability of polymers known to be used or present in the MCU, Defense Waste Processing Facility (DWPF), and Saltstone facilities that will be exposed to the NGS showed that TiDG could selectively affect the elastomers and some thermoplastics to varying extents, but the typical use of these polymers in a confined geometry will likely prevent the NGS from impacting component performance. The polymers identified as of primary concern include Grafoil® (flexible graphite), Tefzel®, Isolast®, ethylene-propylene-diene monomer (EPDM) rubber, nitrile-butadiene rubber (NBR), styrene-butadiene rubber (SBR), ultra high molecular weight polyethylene (UHMWPE), and fluorocarbon rubber (FKM). Certain polymers like NBR and EPDM were found to interact mildly with NGS but their calculated swelling and the confined geometry will impede interaction with NGS. In addition, it was found that Vellumoid (cellulose fibers-reinforced glycerin and protein) may leach protein and Polyvinyl Chloride (PVC) may leach plasticizer (such as Bis-Ethylhexyl-Phthalates) into the NGS solvent. Either case

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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.

  18. 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.

  19. 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

  20. 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

  1. 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.

  2. 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

  3. 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.

  4. 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

  5. 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

  6. 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].

  7. 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.

  8. 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

  9. How nuclear power began

    International Nuclear Information System (INIS)

    Gowing, M.

    1987-01-01

    Many of the features of the story of nuclear power, both in nuclear weapons and nuclear power stations, derive from their timing. Usually, in the history of science the precise timing of discovery does not make much difference, but in the case of nuclear fission there was the coincidence that crucial discoveries were made and openly published in the same year, 1939, as the outbreak of the Second World War. It is these events of the 1930s and the early post-war era that are mainly discussed. However, the story began a lot earlier and even in the early 1900s the potential power within the atom had been foreseen by Soddy and Rutherford. In the 1930s Enrico Fermi and his team saw the technological importance of their discoveries and took out a patent on their process to produce artificial radioactivity from slow neutron beams. The need for secrecy because of the war, and the personal trusts and mistrusts run through the story of nuclear power. (UK)

  10. 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).

  11. 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)

  12. 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

  13. Successful start for new CLIC test facility

    CERN Document Server

    2004-01-01

    A new test facility is being built to study key feasibility issues for a possible future linear collider called CLIC. Commissioning of the first part of the facility began in June 2003 and nominal beam parameters have been achieved already.

  14. Stainless Steel RSM Beneficial Reuse technical feasibility to business reality

    International Nuclear Information System (INIS)

    Boettinger, W.L.; Mishra, G.

    1997-08-01

    The Stainless Steel Beneficial Reuse Program began in 1994 as a demonstration funded by the DOE Office of Science and Technology. The purpose was to assess the practicality of stainless steel radioactive scrap metal (RSM) recycle. Technical feasibility has been demonstrated through the production of a number of products made from recycled RSM. A solid business foundation is yet to be achieved. However, a business environment is beginning to develop as multiple markets and applications for RSM are surfacing around the Complex. The criteria for a successful business reality includes: - affordable programs, - a continuing production base from which to expand, - real products needs, -adequate RSM supply, and - a multi-year program This program currently sponsored by SRS and DOE-ORO to fabricate Defense Waste Processing Facility (DWPF) canisters from RSM provides an activity that satisfies these criteria. The program status is discussed. A comparison of the cost of DWPF canisters fabricated from recycled RSM and virgin metal is presented. The comparison is a function of several factors: disposal costs, the fabrication cost of virgin metal canisters, the fabrication cost of recycled RSM canisters, free release decontamination costs, and the cost to accumulate the RSM. These variables are analyzed and the relationship established to show the break-even point for various values of each parameter

  15. 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

  16. 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

  17. 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

  18. 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.'

  19. 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.'

  20. 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.

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. Ever Since the World Began: A Reading & Interview with Masha Tupitsyn

    OpenAIRE

    Masha Tupitsyn

    2013-01-01

    Writer and cultural critic Masha Tupitsyn is interviewed on her audio recording of her reading Ever Since This World Began, produced specially for this issue of continent. and adapted from her recently published Love Dog (Success and Failure) out from Penny-Ante Editions.

  20. 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

  1. Ever Since the World Began: A Reading & Interview with Masha Tupitsyn

    Directory of Open Access Journals (Sweden)

    Masha Tupitsyn

    2013-07-01

    Full Text Available Writer and cultural critic Masha Tupitsyn is interviewed on her audio recording of her reading Ever Since This World Began, produced specially for this issue of continent. and adapted from her recently published Love Dog (Success and Failure out from Penny-Ante Editions.

  2. 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

  3. 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

  4. 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)

  5. Large Scale Leach Test Facility: Development of equipment and methods, and comparison to MCC-1 leach tests

    International Nuclear Information System (INIS)

    Pellarin, D.J.; Bickford, D.F.

    1985-01-01

    This report describes the test equipment and methods, and documents the results of the first large-scale MCC-1 experiments in the Large Scale Leach Test Facility (LSLTF). Two experiments were performed using 1-ft-long samples sectioned from the middle of canister MS-11. The leachant used in the experiments was ultrapure deionized water - an aggressive and well characterized leachant providing high sensitivity for liquid sample analyses. All the original test plan objectives have been successfully met. Equipment and procedures have been developed for large-sample-size leach testing. The statistical reliability of the method has been determined, and ''bench mark'' data developed to relate small scale leach testing to full size waste forms. The facility is unique, and provides sampling reliability and flexibility not possible in smaller laboratory scale tests. Future use of this facility should simplify and accelerate the development of leaching models and repository specific data. The factor of less than 3 for leachability, corresponding to a 200,000/1 increase in sample volume, enhances the credibility of small scale test data which precedes this work, and supports the ability of the DWPF waste form to meet repository criteria

  6. Implementation of flowsheet change to minimize hydrogen and ammonia generation during chemical processing of high level waste in the Defense Waste Processing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, Dan P. [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); Williams, Matthew S. [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); Luther, Michelle C. [Auburn Univ., AL (United States); Brandenburg, Clayton H. [Univ.of South Carolina, Columbia, SC (United States)

    2016-09-27

    Testing was completed to develop a chemical processing flowsheet for the Defense Waste Processing Facility (DWPF), designed to vitrify and stabilize high level radioactive waste. DWPF processing uses a reducing acid (formic acid) and an oxidizing acid (nitric acid) to rheologically thin the slurry and complete the necessary acid base and reduction reactions (primarily mercury and manganese). Formic acid reduces mercuric oxide to elemental mercury, allowing the mercury to be removed during the boiling phase of processing through steam stripping. In runs with active catalysts, formic acid can decompose to hydrogen and nitrate can be reduced to ammonia, both flammable gases, due to rhodium and ruthenium catalysis. Replacement of formic acid with glycolic acid eliminates the generation of rhodium- and ruthenium-catalyzed hydrogen and ammonia. In addition, mercury reduction is still effective with glycolic acid. Hydrogen, ammonia and mercury are discussed in the body of the report. Ten abbreviated tests were completed to develop the operating window for implementation of the flowsheet and determine the impact of changes in acid stoichiometry and the blend of nitric and glycolic acid as it impacts various processing variables over a wide processing region. Three full-length 4-L lab-scale simulations demonstrated the viability of the flowsheet under planned operating conditions. The flowsheet is planned for implementation in early 2017.

  7. 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

  8. 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

  9. Antifoam degradation testing

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, D. P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Ecology Lab. (SREL); Zamecnik, J. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Ecology Lab. (SREL); Newell, D. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Ecology Lab. (SREL); Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Ecology Lab. (SREL)

    2015-08-20

    This report describes the results of testing to quantify the degradation products resulting from the dilution and storage of Antifoam 747. Antifoam degradation is of concern to the Defense Waste Processing Facility (DWPF) due to flammable decomposition products in the vapor phase of the Chemical Process Cell vessels, as well as the collection of flammable and organic species in the offgas condensate. The discovery that hexamethyldisiloxane is formed from the antifoam decomposition was the basis for a Potential Inadequacy in the Safety Analysis declaration by the DWPF.

  10. DEVELOPMENT OF REMOTE HANFORD CONNECTOR GASKET REPLACEMENT TOOLING FOR THE SAVANNAH RIVER SITE'S DEFENSE WASTE PROCESSING FACILITY

    International Nuclear Information System (INIS)

    Krementz, D

    2007-01-01

    The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF) requested development of tooling for remote replacement of gaskets in mechanical Hanford connectors. The facility has compressed air supply, two master-slave manipulators (MSM's) and a lightweight robotic arm for operation of the remote tools. The Savannah River National Laboratory (SRNL) developed and tested multiple tools to perform the gasket replacement tasks. Separate pneumatic snap-ring removal tools that use the connector skirt as a reaction surface were developed for removal of the snap ring and spent gasket on both vertical and horizontal Hanford connectors. A pneumatic tool that clamps and centers on the jumper pipe ID was developed to simultaneously install the new gasket and snap ring. A pneumatic snap-ring-loading tool was developed that compresses the snap ring and places it in a groove in the installation tool. 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 vents the tools as needed. The entire system has been successfully tested using MSM's to manipulate the various tools. Deployment of the entire system is expected during FY08. The Hanford connector gasket replacement tooling has been successfully tested using MSM's to manipulate the various tools. Nitric acid is used in many of the decontamination processes performed in the REDC, where the tooling will be deployed. Although most of the tool components were fabricated/purchased with nitric acid and radioactive service in mind, some of the prototype parts must be replaced with parts that are more compatible with nitric acid/radioactive service. Several modifications to the various tools are needed to facilitate maintenance and replacement of failed components. Development of installation tools for replacement of 1-inch, 2-inch and multi-hole gaskets is being considered. Deployment of the existing system in the DWPF REDC is expected during FY

  11. Hydrogen generation in SRAT with nitric acid and late washing flowsheets

    International Nuclear Information System (INIS)

    Hsu, C.W.

    1992-01-01

    Melter feed preparation processes, incorporating a final wash of the precipitate slurry feed to Defense Waste Processing Facility (DWPF) and a partial substitution of the SRAT formic acid requirement with nitric acid, should not produce peak hydrogen generation rates during Cold Chemical Runs (CCR's) and radioactive operation greater than their current, respective hydrogen design bases of 0.024 lb/hr and 1.5 lb/hr. A single SRAT bench-scale process simulation for CCR-s produced a DWPF equivalent peak hydrogen generation rate of 0.004 lb/hr. During radioactive operation, the peak hydrogen generation rate will be dependent on the extent DWPF deviates from the nominal precipitate hydrolysis and melter feed preparation process operating parameters. Two actual radioactive sludges were treated according to the new flowsheets. The peak hydrogen evolution rates were equivalent to 0.038 and 0.20 lb/hr (DWPF scale) respectively. Compared to the formic acid -- HAN hydrolysis flowsheets, these peak rates were reduced by a factor of 2.5 and 3.4 for Tank 15 and Tank 11 sludges, respectively

  12. Confinement Physics Research Facility/ZTH: A progress report

    International Nuclear Information System (INIS)

    Hammer, C.F.; Thullen, P.

    1989-01-01

    In October 1985 the Los Alamos National Laboratory's Controlled Thermonuclear Research (CTR) Division began the design and construction of the Confinement Physics Research Facility (CPRF) and the ZTH toroidal, reversed-field-pinch (RFP), plasma physics experiment. The CPRF is a facility which will provide the buildings, utilities, pulsed power system, control system and diagnostics needed to operate a magnetically confined fusion experiment, and ZTH will be the first experiment operated in the facility. The construction of CPRF/ZTH is scheduled for completion in the first quarter of 1993. 5 figs

  13. Effect of canister size on costs of disposal of SRP high-level wastes

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1982-01-01

    The current plan for managing the high-level nuclear wastes at the Savannah River Plant (SRP) calls for processing them into solid forms contained in stainless steel canisters for eventual disposal in a federal geologic repository. A new SRP facility called the Defense Waste Processing Facility (DWPF) is being designed for the onsite waste processing operations. Preliminary evaluations indicate that costs of the overall disposal operation will depend significantly on the size of the canisters, which determines the number of waste forms to be processed. The objective of this study was to evaluate the effects of canister size on costs of DWPF process operations, including canister procurement, waste solidification, and interim storage, on offsite transport, and on repository costs of disposal, including provision of suitable waste packages

  14. Evaluation Of A Turbidity Meter For Use At The Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Mahannah, R. N.; Edwards, T. B.

    2013-01-01

    Savannah River Remediation's (SRR's) Defense Waste Processing Facility (DWPF) Laboratory currently tests for sludge carry-over into the Recycle Collection Tank (RCT) by evaluating the iron concentration in the Slurry Mix Evaporator Condensate Tank (SMECT) and relating this iron concentration to the amount of sludge solids present. A new method was proposed for detecting the amount of sludge in the SMECT that involves the use of an Optek turbidity sensor. Waste Services Laboratory (WSL) personnel conducted testing on two of these units following a test plan developed by Waste Solidification Engineering (WSE). Both Optek units (SN64217 and SN65164) use sensor model AF16-N and signal converter model series C4000. The sensor body of each unit was modified to hold a standard DWPF 12 cc sample vial, also known as a ''peanut'' vial. The purpose of this testing was to evaluate the use of this model of turbidity sensor, or meter, to provide a measurement of the sludge solids present in the SMECT based upon samples from that tank. During discussions of the results from this study by WSE, WSL, and Savannah River National Laboratory (SRNL) personnel, an upper limit on the acceptable level of solids in SMECT samples was set at 0.14 wt%. A ''go/no-go'' decision criterion was to be developed for the critical turbidity response, which is expressed in concentration units (CUs), for each Optek unit based upon the 0.14 wt% solids value. An acceptable or a ''go'' decision for the SMECT should reflect the situation that there is an identified risk (e.g. 5%) for a CU response from the Optek unit to be less than the critical CU value when the solids content of the SMECT is actually 0.14 wt% or greater, while a ''no-go'' determination (i.e., an Optek CU response above the critical CU value, a conservative decision relative to risk) would lead to additional evaluations of the SMECT to better quantify the possible solids content of the tank. A sludge simulant was used to develop standards

  15. Evaluation Of A Turbidity Meter For Use At The Defense Waste Processing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Mahannah, R. N.; Edwards, T. B.

    2013-01-15

    Savannah River Remediation's (SRR's) Defense Waste Processing Facility (DWPF) Laboratory currently tests for sludge carry-over into the Recycle Collection Tank (RCT) by evaluating the iron concentration in the Slurry Mix Evaporator Condensate Tank (SMECT) and relating this iron concentration to the amount of sludge solids present. A new method was proposed for detecting the amount of sludge in the SMECT that involves the use of an Optek turbidity sensor. Waste Services Laboratory (WSL) personnel conducted testing on two of these units following a test plan developed by Waste Solidification Engineering (WSE). Both Optek units (SN64217 and SN65164) use sensor model AF16-N and signal converter model series C4000. The sensor body of each unit was modified to hold a standard DWPF 12 cc sample vial, also known as a ''peanut'' vial. The purpose of this testing was to evaluate the use of this model of turbidity sensor, or meter, to provide a measurement of the sludge solids present in the SMECT based upon samples from that tank. During discussions of the results from this study by WSE, WSL, and Savannah River National Laboratory (SRNL) personnel, an upper limit on the acceptable level of solids in SMECT samples was set at 0.14 wt%. A ''go/no-go'' decision criterion was to be developed for the critical turbidity response, which is expressed in concentration units (CUs), for each Optek unit based upon the 0.14 wt% solids value. An acceptable or a ''go'' decision for the SMECT should reflect the situation that there is an identified risk (e.g. 5%) for a CU response from the Optek unit to be less than the critical CU value when the solids content of the SMECT is actually 0.14 wt% or greater, while a ''no-go'' determination (i.e., an Optek CU response above the critical CU value, a conservative decision relative to risk) would lead to additional evaluations of the SMECT to better quantify the possible solids content of the tank. A sludge simulant was used to develop standards

  16. Phase 2 Report--Mercury Behavior In The Defense Waste Processing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bannochie, C. [Savannah River Site (SRS), Aiken, SC (United States); Fellinger, T. [Savannah River Site (SRS), Aiken, SC (United States)

    2016-07-27

    The purpose of this report is to provide a summary of the DWPF processing history in regards to mercury, document the mercury results obtained on the product and condensate samples, and provide further recommendations based on the data obtained.

  17. 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.

  18. History of the small cylindrical melter

    International Nuclear Information System (INIS)

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

    1985-08-01

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

  19. 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

  20. 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.

  1. Nitric-glycolic flowsheet testing for maximum hydrogen generation rate

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site is developing for implementation a flowsheet with a new reductant to replace formic acid. Glycolic acid has been tested over the past several years and found to effectively replace the function of formic acid in the DWPF chemical process. The nitric-glycolic flowsheet reduces mercury, significantly lowers the chemical generation of hydrogen and ammonia, allows purge reduction in the Sludge Receipt and Adjustment Tank (SRAT), stabilizes the pH and chemistry in the SRAT and the Slurry Mix Evaporator (SME), allows for effective adjustment of the SRAT/SME rheology, and is favorable with respect to melter flammability. The objective of this work was to perform DWPF Chemical Process Cell (CPC) testing at conditions that would bound the catalytic hydrogen production for the nitric-glycolic flowsheet.

  2. 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.)

  3. 2727-S Nonradioactive Dangerous Waste Storage Facility Closure Plan

    International Nuclear Information System (INIS)

    Wilczek, T.A.; Laws, J.R.; Izatt, R.D.

    1992-01-01

    This closure plan describes the activities for final closure of the 2727-S Nonradioactive Dangerous Waste Storage (NRDWS) Facility at the Hanford Site. The 2727-S NRDWS Facility provided container storage for nonradioactive dangerous and extremely hazardous wastes generated in the research and development laboratories, process operations, and maintenance and transportation functions throughout the Hanford Site. Storage operations began at the 2727-S NRDWS Facility March 14, 1983, and continued until December 30, 1986, when the last shipment of materials from the facility took place. These storage operations have been moved to the new 616 NRDWS Facility, which is an interim status unit located between the 200 East and 200 West Areas of the Hanford Site

  4. Sample vial inserts: A better approach for sampling heterogeneous slurry samples in the SRS Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Coleman, C.J.; Goode, S.R.

    1996-01-01

    A convenient and effective new approach for analyzing DWPF samples involves the use of inserts with volumes of 1.5--3 ml placed in the neck of 14 ml sample vials. The inserts have rims that conform to the rim of the vials so that they sit straight and stable in the vial. The DWPF tank sampling system fills the pre-weighed insert rather than the entire vial, so the vial functions only as the insert holder. The shielded cell operator then removes the vial cap and decants the insert containing the sample into a plastic bottle, crucible, etc., for analysis. Inert materials such as Teflon, plastic, and zirconium are used for the insert so it is unnecessary to separate the insert from the sample for most analyses. The key technique advantage of using inserts to take DWPF samples versus filling sample vials is that it provides a convenient and almost foolproof way of obtaining and handling small volumes of slurry samples in a shielded cell without corrupting the sample. Since the insert allows the entire sample to be analyzed, this approach eliminates the errors inherent with subsampling heterogeneous slurries that comprise DWPF samples. Slurry samples can then be analyzed with confidence. Analysis times are dramatically reduced by eliminating the drying and vitrification steps normally used to produce a homogeneous solid sample. Direct dissolution and elemental analysis of slurry samples are achieved in 8 hours or less compared with 40 hours for analysis of vitrified slurry samples. Comparison of samples taken in inserts versus full vials indicate that the insert does not significantly affect sample composition

  5. 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

  6. 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.

  7. Savannah River waste plant takes another broadside

    International Nuclear Information System (INIS)

    Setzer, S.W.

    1992-01-01

    This article is a discussion of Government Accounting Office findings related to the high-level waste disposal facilities, and in particular the Defense Waste Processing Facility, at Savannah River. Cost and schedule problems are noted, and the report concluded that ineffective management, both by DOE personnel and M ampersand AO contractor personnel, was a principal factor contributing to these problems at the DWPF and supporting facilities

  8. Rheology of Savannah River Site Tank 42 radioactive sludges. Revision 1

    International Nuclear Information System (INIS)

    Ha, B.C.; Bibler, N.E.

    1995-01-01

    Knowledge of the rheology of the radioactive sludge slurries at the Savannah River Site (SRS) is necessary in order to ensure that they can be retrieved from waste tanks and processed for final disposal. At Savannah River Site (SRS), Tank 42 sludge represents one of the first HLW radioactive sludges to be vitrified in the Defense Waste Processing Facility (DWPF). The rheological properties of unwashed Tank 42 sludge slurries at various solids concentrations were measured remotely in the Shielded Cells at the Savannah River Technology Center (SRTC) using a modified Haake Rotovisco viscometer. Rheological properties of Tank 42 radioactive sludge were measured as a function of weight percent total solids to ensure that the first DWPF radioactive sludge batch can be pumped and processed in the DWPF with the current design bases. The yield stress and consistency of the sludge slurries were determined by assuming a Bingham plastic fluid model

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

    International Nuclear Information System (INIS)

    Clemons, J.S.

    1993-01-01

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

  10. 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.

  11. 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

  12. 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

  13. Analyses and Comparison of Bulk and Coil Surface Samples from the DWPF Slurry Mix Evaporator

    International Nuclear Information System (INIS)

    Hay, M.; Nash, C.; Stone, M.

    2012-01-01

    Sludge samples from the DWPF Slurry Mix Evaporator (SME) heating coil frame and coil surface were characterized to identify differences that might help identify heat transfer fouling materials. The SME steam coils have seen increased fouling leading to lower boil-up rates. Samples of the sludge were taken from the coil frame somewhat distant from the coil (bulk tank material) and from the coil surface (coil surface sample). The results of the analysis indicate the composition of the two SME samples are very similar with the exception that the coil surface sample shows ∼5-10X higher mercury concentration than the bulk tank sample. Elemental analyses and x-ray diffraction results did not indicate notable differences between the two samples. The ICP-MS and Cs-137 data indicate no significant differences in the radionuclide composition of the two SME samples. Semi-volatile organic analysis revealed numerous organic molecules, these likely result from antifoaming additives. The compositions of the two SME samples also match well with the analyzed composition of the SME batch with the exception of significantly higher silicon, lithium, and boron content in the batch sample indicating the coil samples are deficient in frit relative to the SME batch composition.

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

    International Nuclear Information System (INIS)

    Shadday, M.A. Jr.

    1994-08-01

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

  15. Development of the Holifield Radioactive Ion Beam Facility

    International Nuclear Information System (INIS)

    Tatum, B.A.

    1997-01-01

    The Holifield Radioactive Ion Beam Facility (HRIBF) construction project has been completed and the first radioactive ion beam has been successfully accelerated. The project, which began in 1992, has involved numerous facility modifications. The Oak Ridge Isochronous Cyclotron has been converted from an energy booster for heavy ion beams to a light ion accelerator with internal ion source. A target-ion source and mass analysis system have been commissioned as key components of the facility's radioactive ion beam injector to the 25MV tandem electrostatic accelerator. Beam transport lines have been completed, and new diagnostics for very low intensity beams have been developed. Work continues on a unified control system. Development of research quality radioactive beams for the nuclear structure and nuclear astrophysics communities continues. This paper details facility development to date

  16. 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

  17. EXPERIENCE AND PLANS OF THE JLAB FEL FACILITY AS A USER FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Michelle D. Shinn

    2007-08-26

    Jefferson Lab's IR Upgrade FEL building was planned from the beginning to be a user facility, and includes an associated 600 m2 area containing seven laboratories. The high average power capability (multikilowatt-level) in the near-infrared (1-3 microns), and many hundreds of watts at longer wavelengths, along with an ultrafast (~ 1 ps) high PRF (10's MHz) temporal structure makes this laser a unique source for both applied and basic research. In addition to the FEL, we have a dedicated laboratory capable of delivering high power (many tens of watts) of broadband THz light. After commissioning the IR Upgrade, we once again began delivering beam to users in 2005. In this presentation, I will give an overview of the FEL facility and its current performance, lessons learned over the last two years, and a synopsis of current and future experiments.

  18. Status of the SXFEL Facility

    Directory of Open Access Journals (Sweden)

    Zhentang Zhao

    2017-06-01

    Full Text Available The Shanghai soft X-ray Free-Electron Laser facility (SXFEL is being developed in two steps; the SXFEL test facility (SXFEL-TF, and the SXFEL user facility (SXFEL-UF. The SXFEL-TF is a critical development step towards the construction a soft X-ray FEL user facility in China, and is under commissioning at the Shanghai Synchrotron Radiation Facility (SSRF campus. The test facility is going to generate 8.8 nm FEL radiation using an 840 MeV electron linac passing through the two-stage cascaded HGHG-HGHG or EEHG-HGHG (high-gain harmonic generation, echo-enabled harmonic generation scheme. The construction of the SXFEL-TF started at the end of 2014. Its accelerator tunnel and klystron gallery were ready for equipment installation in April 2016, and the installation of the SXFEL-TF linac and radiator undulators were completed by the end of 2016. In the meantime, the SXFEL-UF, with a designated wavelength in the water window region, began construction in November 2016. This was based on upgrading the linac energy to 1.5 GeV, and the building of a second undulator line and five experimental end-stations. Construction status and the future plans of the SXFEL are reported in this paper.

  19. An Assessment of Using Vibrational Compaction of Calcined HLW and LLW in DWPF Canisters

    International Nuclear Information System (INIS)

    Yi, Yun-Bo; Amme, Robert C.; Shayer, Zeev

    2008-01-01

    Since 1963, the INEL has calcined almost 8 million gallons of liquid mixed waste and liquid high-level waste, converting it to some 1.1 million gallons of dry calcine (about 4275.0 m3), which consists of alumina-and zirconia-based calcine and zirconia-sodium blend calcine. In addition, if all existing and projected future liquid wastes are solidified, approximately 2,000 m3 of additional calcine will be produced primarily from sodium-bearing waste. Calcine is a more desirable material to store than liquid radioactive waste because it reduces volume, is much less corrosive, less chemically reactive, less mobile under most conditions, easier to monitor and more protective of human health and the environment. This paper describes the technical issue involved in the development of a feasible solution for further volume reduction of calcined nuclear waste for transportation and long term storage, using a standard DWPF canister. This will be accomplished by developing a process wherein the canisters are transported into a vibrational machine, for further volume reduction by about 35%. The random compaction experiments show that this volume reduction is achievable. The main goal of this paper is to demonstrate through computer modeling that it is feasible to use volume reduction vibrational machine without developing stress/strain forces that will weaken the canister integrity. Specifically, the paper presents preliminary results of the stress/strain analysis of the DWPF canister as a function of granular calcined height during the compaction and verifying that the integrity of the canister is not compromised. This preliminary study will lead to the development of better technology for safe compactions of nuclear waste that will have significant economical impact on nuclear waste storage and treatment. The preliminary results will guide us to find better solutions to the following questions: 1) What are the optimum locations and directions (vertical versus horizontal or

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

    International Nuclear Information System (INIS)

    Colven, W.P.

    1983-01-01

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

  1. 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

  2. Introducing COSS: A new and unique oil spill research facility

    International Nuclear Information System (INIS)

    Kitchen, R. B.; Bonner, J. S.; Autenrieth, R. L.; Donnelly, K. C.; Ernest, A. N. S.

    1997-01-01

    A new oil spill research facility in Corpus Christi, Texas began operation in April 1997 to address the appropriate use, application and effectiveness of chemical, physical and biological oil spill response agents. The Coastal Oil Spill Simulation (COSS) facility consists of nine meso scale wave tanks and will offer to science and industry a unique opportunity to spill oil in a controlled environment and to study fate, transport and remediation of oil releases in simulated coastal, intertidal, lagunal, channel and porous media. 1 ref

  3. Decommissioning the UHTREX Reactor Facility at Los Alamos, New Mexico

    International Nuclear Information System (INIS)

    Salazar, M.; Elder, J.

    1992-08-01

    The Ultra-High Temperature Reactor Experiment (UHTREX) facility was constructed in the late 1960s to advance high-temperature and gas-cooled reactor technology. The 3-MW reactor was graphite moderated and helium cooled and used 93% enriched uranium as its fuel. The reactor was run for approximately one year and was shut down in February 1970. The decommissioning of the facility involved removing the reactor and its associated components. This document details planning for the decommissioning operations which included characterizing the facility, estimating the costs of decommissioning, preparing environmental documentation, establishing a system to track costs and work progress, and preplanning to correct health and safety concerns in the facility. Work to decommission the facility began in 1988 and was completed in September 1990 at a cost of $2.9 million. The facility was released to Department of Energy for other uses in its Los Alamos program

  4. Startup of Large Coil Test Facility

    International Nuclear Information System (INIS)

    Haubenreich, P.N.; Bohanan, R.E.; Fietz, W.A.; Luton, J.N.; May, J.R.

    1985-01-01

    The Large Coil Test Facility (LCTF) is being used to test superconducting toroidal field coils about one-third the size of those for INTOR. Eventually, six different coils from four countries will be tested. Operations began in 1983 with acceptance testing of the helium refrigerator/liquefier system. Comprehensive shakedown of the facility and tests with the first three coils (from Japan, the United States, and Switzerland) were successfully accomplished in the summer of 1984. Currents up to 10,200 A and fields up to 6.4 T were reached. Data were obtained on performance of refrigerator, helium distribution, power supplies, controls, and data acquisition systems and on the acoustic emission, voltages, currents, and mechanical strains during charging and discharging the coils

  5. Startup of Large Coil Test Facility

    International Nuclear Information System (INIS)

    Haubenreich, P.N.; Bohanan, R.E.; Fietz, W.A.; Luton, J.N.; May, J.R.

    1984-01-01

    The Large Coil Test Facility (LCTF) is being used to test superconducting toroidal field coils about one-third the size of those for INTOR. Eventually, six different coils from four countries will be tested. Operations began in 1983 with acceptance testing of the helium refrigerator/liquefier system. Comprehensive shakedown of the facility and tests with the first three coils (from Japan, the United States, and Switzerland) were successfully accomplished in the summer of 1984. Currents up to 10,200 A and fields up to 6.4 T were reached. Data were obtained on performance of refrigerator, helium distribution, power supplies, controls, and data acquisition systems and on the acoustic emission, voltages, currents, and mechanical strains during charging and discharging the coils

  6. SRS ES and H Standards Compliance Program Implementation Plan

    International Nuclear Information System (INIS)

    Hearn, W.H.

    1993-01-01

    On March 8, 1990, the Defense Nuclear Facilities Safety Board (DNFSB) issued Recommendation 90-2 to the Secretary of Energy. This recommendation, based upon the DNFSB's initial review and evaluation of the content and implementation of standards relating to the design, construction, operations, and decommissioning of defense nuclear facilities of the U.S. Department of Energy (DOE), called for three actions: identification of specific standards that apply to design, construction, operation and decommissioning of DOE facilities; assessment of the adequacy of those standards for protecting public health and safety; and determination of the extent to which they have and are being implemented. The purpose of this Implementation Plan is to define the single program for all sitewide and facility 90-2 ES and H Standards Compliance efforts, which will satisfy the HQ Implementation Plan, avoid duplicate efforts, be as simple and achievable as possible, include cost-saving innovations, use a graded approach based on facility hazards and future needs of facilities, and support configuration control for facility requirements. The Defense Waste Processing Facility (DWPF) has been designated a pilot facility for the 90-2 program and has progressed with their facility program ahead of the site-level program. The DWPF, and other Government-Owned Contractor-Operated (GOCO) facilities that progress on an enhanced schedule, will serve as pilot facilities for the site-level program. The lessons learned with their requirement identifications, and their assessments of the adequacy of and their compliance with these requirements will be used to improve the efficiency of the site-level and subsequent programs

  7. Experimental Facilities Division. Progress report 1996-97

    International Nuclear Information System (INIS)

    1997-04-01

    This progress report summarizes the activities of the Experimental Facilities Division (XFD) in support of the users of the Advanced Photon Source (APS), primarily focusing on the past year of operations. In September 1996, the APS began operations as a national user facility serving the US community of x-ray researchers from private industry, academic institutions, and other research organizations. The start of operations was about three months ahead of the baseline date established in 1988. This report is divided into the following sections: (1) overview; (2) user operations; (3) user administration and technical support; (4) R and D in support of view operations; (5) collaborative research; and (6) long-term strategic plans for XFD

  8. Experimental Facilities Division progress report 1996--97

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

    This progress report summarizes the activities of the Experimental Facilities Division (XFD) in support of the users of the Advanced Photon Source (APS), primarily focusing on the past year of operations. In September 1996, the APS began operations as a national user facility serving the US community of x-ray researchers from private industry, academic institutions, and other research organizations. The start of operations was about three months ahead of the baseline date established in 1988. This report is divided into the following sections: (1) overview; (2) user operations; (3) user administration and technical support; (4) R and D in support of view operations; (5) collaborative research; and (6) long-term strategic plans for XFD.

  9. Study on eliminating fire dampers to maintain process confinement

    International Nuclear Information System (INIS)

    Walling, R.C.; Patel, J.B.; Strunk, A.J.

    1991-01-01

    The DOE General Design Criteria for the Defense Waste Processing Facility (DWPF) at the Westinghouse Savannah River Site (WSRS) requires the NFPA National Fire Codes to be incorporated into the design and simultaneously maintain process confinement integrity to prevent the release of radioactivity. Although the NFPA Standard for the Installation of Air Conditioning and Ventilation Systems, NFPA 90, requires fire dampers (FD) in HVAC duct penetrations of two hour rated fire barriers, closure of fire dampers at DWPF may compromise the integrity of the process confinement system. This leads to the need for an overall risk assessment to determine the value of 39 fire dampers that are identified later in the study as capable of a confinement system upset

  10. 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

  11. Evaluation of a turbidity meter for use at the Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Mahannah, R.N; Edwards, T.B.

    2013-01-01

    Savannah River RemediationÆs (SRRÆs) Defense Waste Processing Facility (DWPF) Laboratory currently tests for sludge carry-over into the Recycle Collection Tank (RCT) by evaluating the iron concentration in the Slurry Mix Evaporator Condensate Tank (SMECT) and relating this iron concentration to the amount of sludge solids present. A new method was proposed for detecting the amount of sludge in the SMECT that involves the use of an Optek turbidity sensor. Waste Services Laboratory (WSL) personnel conducted testing on two of these units following a test plan developed by Waste Solidification Engineering (WSE). Both Optek units (SN64217 and SN65164) use sensor model AF16-N and signal converter model series C4000. The sensor body of each unit was modified to hold a standard DWPF 12 cc sample vial, also known as a ôpeanutö vial. The purpose of this testing was to evaluate the use of this model of turbidity sensor, or meter, to provide a measurement of the sludge solids present in the SMECT based upon samples from that tank. During discussions of the results from this study by WSE, WSL, and Savannah River National Laboratory (SRNL) personnel, an upper limit on the acceptable level of solids in SMECT samples was set at 0.14 weight percent (wt%). A ''go/no-go'' decision criterion was to be developed for the critical turbidity response, which is expressed in concentration units (CUs), for each Optek unit based upon the 0.14 wt% solids value. An acceptable or a ''go'' decision for the SMECT should reflect the situation that there is an identified risk (e.g. 5%) for a CU response from the Optek unit to be less than the critical CU value when the solids content of the SMECT is actually 0.14 wt% or greater, while a ''no-go'' determination (i.e., an Optek CU response above the critical CU value, a conservative decision relative to risk) would lead to additional evaluations of the SMECT to better quantify the possible solids content of the tank. Subsequent to the issuance of

  12. EVALUATION OF A TURBIDITY METER FOR USE AT THE DEFENSE WASTE PROCESSING FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Mahannah, R.; Edwards, T.

    2013-06-04

    Savannah River Remediation’s (SRR’s) Defense Waste Processing Facility (DWPF) Laboratory currently tests for sludge carry-over into the Recycle Collection Tank (RCT) by evaluating the iron concentration in the Slurry Mix Evaporator Condensate Tank (SMECT) and relating this iron concentration to the amount of sludge solids present. A new method was proposed for detecting the amount of sludge in the SMECT that involves the use of an Optek turbidity sensor. Waste Services Laboratory (WSL) personnel conducted testing on two of these units following a test plan developed by Waste Solidification Engineering (WSE). Both Optek units (SN64217 and SN65164) use sensor model AF16-N and signal converter model series C4000. The sensor body of each unit was modified to hold a standard DWPF 12 cc sample vial, also known as a “peanut” vial. The purpose of this testing was to evaluate the use of this model of turbidity sensor, or meter, to provide a measurement of the sludge solids present in the SMECT based upon samples from that tank. During discussions of the results from this study by WSE, WSL, and Savannah River National Laboratory (SRNL) personnel, an upper limit on the acceptable level of solids in SMECT samples was set at 0.14 weight percent (wt%). A “go/no-go” decision criterion was to be developed for the critical turbidity response, which is expressed in concentration units (CUs), for each Optek unit based upon the 0.14 wt% solids value. An acceptable or a “go” decision for the SMECT should reflect the situation that there is an identified risk (e.g. 5%) for a CU response from the Optek unit to be less than the critical CU value when the solids content of the SMECT is actually 0.14 wt% or greater, while a “no-go” determination (i.e., an Optek CU response above the critical CU value, a conservative decision relative to risk) would lead to additional evaluations of the SMECT to better quantify the possible solids content of the tank. Subsequent to the

  13. Quality Assurance Program description, Defense Waste Processing Facility (DWPF)

    International Nuclear Information System (INIS)

    Maslar, S.R.

    1992-01-01

    This document describes the Westinghouse Savannah River Company's (WSRC) Quality Assurance Program for Defense Waste Processing at the Savannah River Site (SRS). WSRC is the operating contractor for the US Department of Energy (DOE) at the SRS. The following objectives are achieved through developing and implementing the Quality Assurance Program: (1) Ensure that the attainment of quality (in accomplishing defense high-level waste processing objectives at the SRS) is at a level commensurate with the government's responsibility for protecting public health and safety, the environment, the public investment, and for efficiently and effectively using national resources. (2) Ensure that high-level waste from qualification and production activities conform to requirements defined by OCRWM. These activities include production processes, equipment, and services; and products that are planned, designed, procured, fabricated, installed, tested, operated, maintained, modified, or produced

  14. Before time began the Big Bang and the emerging universe

    CERN Document Server

    Satz, Helmut

    2017-01-01

    What is the origin of the universe? What was there before the universe appeared? We are currently witnessing a second Copernican revolution: neither our Earth and Sun, nor our galaxy, nor even our universe, are the end of all things. Beyond our world, in an endless multiverse, are innumerable other universes, coming and going, like ours or different. Fourteen billion years ago, one of the many bubbles constantly appearing and vanishing in the multiverse exploded to form our universe. The energy liberated in the explosion provided the basis for all the matter our universe now contains. But how could this hot, primordial plasma eventually produce the complex structure of our present world? Does not order eventually always lead to disorder, to an increase of entropy? Modern cosmology is beginning to find out how it all came about and where it all might lead. Before Time Began tells that story.

  15. Refurbishment and Automation of Thermal Vacuum Facilities at NASA/GSFC

    Science.gov (United States)

    Dunn, Jamie; Gomez, Carlos; Donohue, John; Johnson, Chris; Palmer, John; Sushon, Janet

    1999-01-01

    The thermal vacuum facilities located at the Goddard Space Flight Center (GSFC) have supported both manned and unmanned space flight since the 1960s. Of the eleven facilities, currently ten of the systems are scheduled for refurbishment or replacement as part of a five-year implementation. Expected return on investment includes the reduction in test schedules, improvements in safety of facility operations, and reduction in the personnel support required for a test. Additionally, GSFC will become a global resource renowned for expertise in thermal engineering, mechanical engineering, and for the automation of thermal vacuum facilities and tests. Automation of the thermal vacuum facilities includes the utilization of Programmable Logic Controllers (PLCs), the use of Supervisory Control and Data Acquisition (SCADA) systems, and the development of a centralized Test Data Management System. These components allow the computer control and automation of mechanical components such as valves and pumps. The project of refurbishment and automation began in 1996 and has resulted in complete computer control of one facility (Facility 281), and the integration of electronically controlled devices and PLCs in multiple others.

  16. 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

  17. 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.

  18. Status of Thomas Jefferson National Accelerator Facility (Jefferson Lab)

    International Nuclear Information System (INIS)

    H.A. Grunder

    1997-01-01

    When first beam was delivered on target in July 1994, the Continuous Electron Beam Accelerator Facility (CEBAF), in Newport News, Virginia realized the return on years of planning and work to create a laboratory devoted to exploration of matter that interacts through the strong force, which holds the quarks inside the proton and binds protons and neutrons into the nucleus. Dedicated this year as the Thomas Jefferson National Accelerator Facility (Jefferson Lab), the completion of construction and beginning of its experimental program has culminated a process that began more than a decade ago with the convening of the Bromley Panel to look at research possibilities for such an electron accelerator

  19. Late Wash/Nitric Acid flowsheet hydrogen generation bases for simulation of a deflagration/detonation in the DWPF CPC

    International Nuclear Information System (INIS)

    Ritter, J.A.

    1993-01-01

    Hydrogen generation data obtained from IDMS runs PX4 and PX5 will be used to determine a bases for a deflagration/detonation simulation in the DWPF CPC. This simulation is necessary due to the new chemistry associated with the Late Wash/ Nitric Acid flowsheet and process modifications associated with the presence of H 2 in the offgas. The simulation will be performed by Professor Van Brunt from the University of South Carolina. The scenario which leads up to the deflagration/detonation simulation will be chosen such that the following conditions apply. The SRAT is filled to its maximum operating level with 9,600 gal of sludge, which corresponds to the minimum vapor space above the sludge. The SRAT is at the boiling point, producing H 2 at a very low rate (about 10 % of the peak) and 15 scfm of air inleakage is entering the SRAT. Then, the H 2 generation rate will be allowed to increase exponentially (catalyst activation) until it readies the peak H 2 generation rate of the IDMS run, after which the H 2 generation rate will be allowed to decay exponentially (catalyst deactivation) until the total amount of H2 produced is between 85 and 100% of that produced during the IDMS run

  20. 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.

  1. 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

  2. Automatically controlled facilities for irradiation of silicon crystals at the Rossendorf Research Reactor

    International Nuclear Information System (INIS)

    Ross, R.

    1988-01-01

    This report describes the facilities for neutron transmutation doping of silicon in GDR. The irradiation of silicon single crystals began at Rossendorf in 1978 with simple equipment. Only a small amount of silicon could be irradiated in it. The fast increasing need of NTD-silicon made it necessary to design and construct new and better facilities. The new facilities are capable of irradiating silicon from 2'' to 3'' in diameter. The irradiation process takes place automatically with the assistance of a computer. Material produced has an axial homogeneity of ± 7%. Irradiation riggs, techniques, irradiation control and quality control are discussed. (author). 4 figs

  3. 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

  4. 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.

  5. 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

  6. Ecological studies related to the construction of the Defense Waste Processing Facility on the Savannah River Site. Annual report, FY-1994 and FY-1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-01

    The Savannah River Ecology Laboratory initiated ecological studies related to the construction of the DWPF on the SRS in FY-1979. Two areas have been used for biological surveys and long-term monitoring: the DWPF construction site (S-Area and Z-Area), and two control sites (Rainbow Bay and Tinker Creek). The Rainbow Bay study area and S-Area are located within 5 km of each other on the SRS, and both once contained Carolina bays which were very similar ecologically. One goal of the SREL`s faunal studies is to compare the natural variation in amphibian populations at the Rainbow Bay control site to the variation observed at the human-altered site (Sun Bay, formerly on the DWPF construction site). Pre-construction biological surveys included data on vegetation, birds, mammals, amphibians, reptiles, fish and several invertebrate groups. No species on the Federal Endangered or Threatened lists were found on either site, but several plants and animals of threatened or special-concern status in South Carolina were present and the gopher frog (Rana areolata) currently is being considered for federal listing. Continuing studies are directed towards assessing construction impacts on the biota and towares modeling the effects of alteration of wetland hydroperiod on the biota. Primary emphasis is being paced on evaluation the effectiveness of mitigation measures undertaken by DOE.

  7. 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

  8. Plan for reevaluation of NRC policy on decommissioning of nuclear facilities

    International Nuclear Information System (INIS)

    1978-03-01

    Recognizing that the current generation of large commercial reactors and supporting nuclear facilities would substantially increase future decommissioning needs, the NRC staff began an in-depth review and re-evaluation of NRC's regulatory approach to decommissioning in 1975. Major technical studies on decommissioning have been initiated at Battelle Pacific Northwest Laboratory in order to provide a firm information base on the engineering methodology, radiation risks, and estimated costs of decommissioning light water reactors and associated fuel cycle facilities. The Nuclear Regulatory Commission is now considering development of a more explicit overall policy for nuclear facility decommissioning and amending its regulations in 10 CFR Parts 30, 40, 50, and 70 to include more specific guidance on decommissioning criteria for production and utilization facility licensees and byproduct, source, and special nuclear material licensees. The report sets forth in detail the NRC staff plan for the development of an overall NRC policy on decommissioning of nuclear facilities

  9. 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

  10. Carbon-14 in sludge

    International Nuclear Information System (INIS)

    Fowler, J.R.; Coleman, C.J.

    1983-01-01

    The level of C-14 in high-level waste is needed to establish the amount of C-14 that will be released to the environment either as off-gas from the Defense Waste Processing Facility (DWPF) or as a component of saltstone. Available experimental data confirmed a low level of C-14 in soluble waste, but no data was available for sludge. Based on the processes used in each area, Purex LAW sludge in F-area and HM HAW sludge in H-area will contain the bulk of any sludge produced by the cladding. Accordingly, samples from Tank 8F containing Purex LAW and Tank 15H containing HM HAW were obtained and analyzed for C-14. These two waste types constitute approximately 70% of the total sludge inventory now stored in the waste tanks. Results from analyses of these two sludge types show: the total C-14 inventory in sludge now stored in the waste tanks is 6.8 Ci; C-14 releases to the atmosphere from the DWPF will average approximately 0.6 Ci annually at the projected sludge processing rate in the DWPF. 4 references, 2 tables

  11. 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

  12. 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

  13. Characterization Of The As-Received Sludge Batch 9 Qualification Sample (Htf-51-15-81)

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-30

    Savannah River National Laboratory (SRNL) personnel have been requested to qualify the next sludge batch (Sludge Batch 9 – SB9) for processing at the Defense Waste Processing Facility (DWPF). To accomplish this task, Savannah River Remediation (SRR) has sent SRNL a 3-L slurried sample of Tank 51H (HTF-51-15-81) to be characterized, washed, and then used in a lab-scale demonstration of the DWPF flowsheet (potentially after combining with Tank 40H sludge). This report documents the first steps of the qualification process – characterization of the as-received Tank 51H qualification sample. These results will be used to support a reprojection of SB9 by SRR from which final Tank 51H washing, frit development, and Chemical Processing Cell (CPC) activities will be based.

  14. 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

  15. 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

  16. 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

  17. Westinghouse Savannah River Company (WSRC) approach to nuclear facility maintenance

    International Nuclear Information System (INIS)

    Harrison, D.W.

    1991-01-01

    The Savannah River Site (SRS) in South Carolina is a 300+ square mile facility owned by the US Department of Energy (DOE) and operated by Westinghouse Savannah River Company (WSRC), the prime contractor; Bechtel Savannah River, Incorporated (BSRI) is a major subcontractor. The site has used all of the five nuclear reactors and it has the necessary nuclear materials processing facilities, as well as waste management and research facilities. The site has produced materials for the US nuclear arsenal and various isotopes for use in space research and nuclear medicine for more than 30 years. In 1989, WSRC took over as prime contractor, replacing E.I. du Pont de Nemours and Company. At this time, a concentrated effort began to more closely align the operating standards of this site with those accepted by the commercial nuclear industry of the United States. Generally, this meant acceptance of standards of the Institute of Nuclear Power Operations (INPO) for nuclear-related facilities at the site. The subject of this paper is maintenance of nuclear facilities and, therefore, excludes discussion of the maintenance of non-nuclear facilities and equipment

  18. Decommissioning and environmental restoration of nuclear facilities in China

    International Nuclear Information System (INIS)

    Pan Ziqiang

    2000-01-01

    In the beginning of the 1980s, the Scientific and Technological Commission (STC) began the study on the environmental impact of the nuclear industry in China. At the end of the 1980s, the STC initiated the study on the decommissioning of nuclear facilities and environmental restoration. In 1989 the STC completed the project entitled ''Radiological and Environmental Quality Assessment of the Nuclear Industry in China Over the Past Thirty Years''. The status of the environmental pollution of various nuclear facility sites was subsequently analysed. In 1994, the decommissioning and environmental restoration of the first research and manufacture complex for nuclear weapons was completed. The complex is now accessible to the public without restriction and the site has become a town. Some nuclear related facilities, such as uranium mines, are currently being decommissioned. Although uranium mining and milling has a more serious impact on the environment, the technology for decommissioning and environmental restoration in mining and milling installations is not much more complicated than that used for reactor and reprocessing facilities: much has been achieved in the area of mining and milling. (author)

  19. And so it all began: A personal tribute to the man behind the scientist

    Science.gov (United States)

    Aliotta, Marialuisa

    2018-01-01

    At the time I began my scientific career as a PhD student under the supervision of Claudio Spitaleri, the Trojan Horse Method was still in its infancy. Like with any new-born idea, it took time and passion and effort to plant the early seeds that would eventually develop into a now well-established method in nuclear astrophysics research. Here, I offer my own recollection of those early years as a personal homage to Claudio's unique mix of human traits that shaped our professional relationship for many years since.

  20. The Geneva conference - How it began

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-08-15

    The First International Conference on the Peaceful Uses of Atomic Energy had its origin in President Eisenhower's initiative of the early nineteen-fifties, when he proposed a concerted international effort to divert the power of the atom from warlike purposes into the service of peace. To the United Nations General Assembly in December 1953, he pledged the determination of the United States 'to help solve the fearful atomic dilemma - to devote its entire heart and mind to finding the way by which the miraculous inventiveness of man shall not be dedicated to his death, but consecrated to his life'. The UN General Assembly in plenary session, in December 1954, unanimously and enthusiastically adopted a resolution which provided for the establishment of an International Atomic Energy Agency, and for the holding of an international technical conference of governments under the auspices of the United Nations. To prepare the way, an Advisory Committee was set up, consisting of representatives of Brazil, Canada, France, India, USSR, United Kingdom and USA. The result was the largest meeting that had been convened under the auspices of the United Nations; it was held from 8 to 25 August 1955 in the Palais des Nations, Geneva, where the necessary facilities were available for such a large multilingual conference. Thirty-eight governments submitted 1067 papers and 1428 participants attended. The conference was wide in scope, embracing all major aspects of the peaceful applications of atomic energy.

  1. The Geneva conference - How it began

    International Nuclear Information System (INIS)

    1964-01-01

    The First International Conference on the Peaceful Uses of Atomic Energy had its origin in President Eisenhower's initiative of the early nineteen-fifties, when he proposed a concerted international effort to divert the power of the atom from warlike purposes into the service of peace. To the United Nations General Assembly in December 1953, he pledged the determination of the United States 'to help solve the fearful atomic dilemma - to devote its entire heart and mind to finding the way by which the miraculous inventiveness of man shall not be dedicated to his death, but consecrated to his life'. The UN General Assembly in plenary session, in December 1954, unanimously and enthusiastically adopted a resolution which provided for the establishment of an International Atomic Energy Agency, and for the holding of an international technical conference of governments under the auspices of the United Nations. To prepare the way, an Advisory Committee was set up, consisting of representatives of Brazil, Canada, France, India, USSR, United Kingdom and USA. The result was the largest meeting that had been convened under the auspices of the United Nations; it was held from 8 to 25 August 1955 in the Palais des Nations, Geneva, where the necessary facilities were available for such a large multilingual conference. Thirty-eight governments submitted 1067 papers and 1428 participants attended. The conference was wide in scope, embracing all major aspects of the peaceful applications of atomic energy.

  2. Statistical evaluation of effluent monitoring data for the 200 Area Treated Effluent Disposal Facility

    International Nuclear Information System (INIS)

    Chou, C.J.; Johnson, V.G.

    2000-01-01

    The 200 Area Treated Effluent Disposal Facility (TEDF) consists of a pair of infiltration basins that receive wastewater originating from the 200 West and 200 East Areas of the Hanford Site. TEDF has been in operation since 1995 and is regulated by State Waste Discharge Permit ST 4502 (Ecology 1995) under the authority of Chapter 90.48 Revised Code of Washington (RCW) and Washington Administrative Code (WAC) Chapter 173-216. The permit stipulates monitoring requirements for effluent (or end-of-pipe) discharges and groundwater monitoring for TEDF. Groundwater monitoring began in 1992 prior to TEDF construction. Routine effluent monitoring in accordance with the permit requirements began in late April 1995 when the facility began operations. The State Waste Discharge Permit ST 4502 included a special permit condition (S.6). This condition specified a statistical study of the variability of permitted constituents in the effluent from TEDF during its first year of operation. The study was designed to (1) demonstrate compliance with the waste discharge permit; (2) determine the variability of all constituents in the effluent that have enforcement limits, early warning values, and monitoring requirements (WHC 1995); and (3) determine if concentrations of permitted constituents vary with season. Additional and more frequent sampling was conducted for the effluent variability study. Statistical evaluation results were provided in Chou and Johnson (1996). Parts of the original first year sampling and analysis plan (WHC 1995) were continued with routine monitoring required up to the present time

  3. Research facility access & science education

    Energy Technology Data Exchange (ETDEWEB)

    Rosen, S.P. [Univ. of Texas, Arlington, TX (United States); Teplitz, V.L. [Southern Methodist Univ., Dallas, TX (United States). Physics Dept.

    1994-10-01

    As Congress voted to terminate the Superconducting Super Collider (SSC) Laboratory in October of 1993, the Department of Energy was encouraged to maximize the benefits to the nation of approximately $2 billion which had already been expended to date on its evolution. Having been recruited to Texas from other intellectually challenging enclaves around the world, many regional scientists, especially physicists, of course, also began to look for viable ways to preserve some of the potentially short-lived gains made by Texas higher education in anticipation of {open_quotes}the SSC era.{close_quotes} In fact, by November, 1993, approximately 150 physicists and engineers from thirteen Texas universities and the SSC itself, had gathered on the SMU campus to discuss possible re-uses of the SSC assets. Participants at that meeting drew up a petition addressed to the state and federal governments requesting the creation of a joint Texas Facility for Science Education and Research. The idea was to create a facility, open to universities and industry alike, which would preserve the research and development infrastructure and continue the educational mission of the SSC.

  4. Overview of Corrosion, Erosion, and Synergistic Effects of Erosion and Corrosion in the WTP Pre-treatment Facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-27

    Corrosion is an extremely complex process that is affected by numerous factors. Addition of a flowing multi-phase solution further complicates the analysis. The synergistic effects of the multiple corrosive species as well as the flow-induced synergistic effects from erosion and corrosion must be thoroughly evaluated in order to predict material degradation responses. Public domain data can help guide the analysis, but cannot reliably provide the design basis especially when the process is one-of-a-kind, designed for 40 plus years of service, and has no viable means for repair or replacement. Testing in representative simulants and environmental conditions with prototypic components will provide a stronger technical basis for design. This philosophy was exemplified by the Defense Waste Processing Facility (DWPF) at the Savannah River Site and only after 15 plus years of successful operation has it been validated. There have been “hiccups”, some identified during the cold commissioning phase and some during radioactive operations, but they were minor and overcome. In addition, the system is robust enough to tolerate most flowsheet changes and the DWPF design allows minor modifications and replacements – approaches not available with the Hanford Waste Treatment Plant (WTP) “Black Cell” design methodology. Based on the available data, the synergistic effect between erosion and corrosion is a credible – virtually certain – degradation mechanism and must be considered for the design of the WTP process systems. Testing is recommended due to the number of variables (e.g., material properties, process parameters, and component design) that can affect synergy between erosion and corrosion and because the available literature is of limited applicability for the complex process chemistries anticipated in the WTP. Applicable testing will provide a reasonable and defensible path forward for design of the WTP Black Cell and Hard-to-Reach process equipment. These

  5. Considerations in the development and implementation of a maintenance robot for nuclear power facilities

    International Nuclear Information System (INIS)

    Rohrabacher, A.; Carlton, R.; Gelhaus, F.

    1987-01-01

    The Nuclear Power Division of the Electric Power Research Institute (EPRI) began its automated nuclear plant maintenance research (research project 2232) in 1982 with surveys of domestic and Japanese ongoing and potential uses of robotic systems in nuclear power facilities. After studying these potential research areas, EPRI initiated a three-pronged effort. All surveys concluded that a mobile system for general environmental, visual, and auto surveillance would be beneficial, and EPRI contracted for the development and evaluation of the Surveyor robot. Work also began to catalog existing systems and to quantify the methodology that a utility could use to optimally complete the plant-mission, robot-system selection process. The conceptualizing phase of the effort defined the necessary physical attributes for the machine to perform a variety of maintenance tasks. Characteristics such as mobility, strength, conditions of service, materials of construction, tool functional requirements, facility interfaces, and operator interface were defined to the extent practical. Substantial versatility is required for the variability of the targeted maintenance tasks and for the variability of facility layouts and designs found throughout the nuclear industry. A key contribution to this development effort comes from the project's Utility Advisors Group. The contract's host utility provides membership to lead this group, and that company will train in the use of the machine for 6 months prior to utilizing it in their power plant for 2 yr

  6. 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

  7. Hot Gas Cleanup Test Facility for gasification and pressurized combustion. Quarterly report, October--December 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-02-01

    The objective of this project is to evaluate hot gas particle control technologies using coal-derived gas streams. This will entail the design, construction, installation, and use of a flexible test facility which can operate under realistic gasification and combustion conditions. The major particulate control device issues to be addressed include the integration of the particulate control devices into coal utilization systems, on-line cleaning techniques, chemical and thermal degradation of components, fatigue or structural failures, blinding, collection efficiency as a function of particle size, and scale-up of particulate control systems to commercial size. The conceptual design of the facility was extended to include a within scope, phased expansion of the existing Hot Gas Cleanup Test Facility Cooperative Agreement to also address systems integration issues of hot particulate removal in advanced coal-based power generation systems. This expansion included the consideration of the following modules at the test facility in addition to the original Transport Reactor gas source and Hot Gas Cleanup Units: carbonizer/pressurized circulating fluidized bed gas source; hot gas cleanup units to mate to all gas streams; combustion gas turbine; and fuel cell and associated gas treatment. The major emphasis during this reporting period was continuing the detailed design of the facility and integrating the particulate control devices (PCDs) into structural and process designs. Substantial progress in underground construction activities was achieved during the quarter. Delivery and construction of coal handling and process structural steel began during the quarter. Delivery and construction of coal handling and process structural steel began during the quarter. MWK equipment at the grade level and the first tier are being set in the structure.

  8. E-government Facilities Analysis for Public Services in Higher Education

    Science.gov (United States)

    Astawa, I. P. M.; Dewi, K. C.

    2018-01-01

    E-Government in higher education can be utilized in order to provide public services to stakeholders both internal and external. The research objectives is to analyze the e-government facilities for public services in higher education. The research began by reviewing the concept of public services and e-government, then continued by analysing e-government facilities based on the E-Government Maturity Level developed by Wirtz and Piehler. The research subject was the e-government website of three universities that ranked the top three of webometrics version (Indonesia country rank), while the research object was e-government facilities for public services. Data collection was done by observing e-government sites via online browsing. The research’s results indicated that all three e-government sites have met four e-government business model and provided e-government services in line with the fourth stage on the e-government development stage. It can concluded that the three universities have achieved e-government maturity at the fourth level.

  9. Developing standardized facility contingency plans

    International Nuclear Information System (INIS)

    Davidson, D.A.

    1993-01-01

    Texaco consists of several operating departments that are, in effect, independent companies. Each of these departments is responsible for complying with all environmental laws and regulations. This includes the preparation by each facility to respond to an oil spill at that location. For larger spills, however, management of the response will rest with corporate regional response teams. Personnel from all departments make up the regional teams. In 1990, Congress passed the Oil Pollution Act. In 1991, the US Coast Guard began developing oil spill response contingency plan regulations, which they are still working on. Meanwhile, four of the five west coast states have also passed laws requiring contingency plans. (Only Hawaii has chosen to wait and see what the federal regulations will entail). Three of the states have already adopted regulations. Given these laws and regulations, along with its corporate structure, Texaco addressed the need to standardize local facility plans as well as its response organization. This paper discusses how, by working together, the Texaco corporate international oil spill response staff and the Texaco western region on-scene commander developed: A standard contingency plan format crossing corporate boundaries and meeting federal and state requirements. A response organization applicable to any size facility or spill. A strategy to sell the standard contingency plan and response organization to the operating units

  10. BEHAVIOR OF MERCURY DURING DWPF CHEMICAL PROCESS CELL PROCESSING

    Energy Technology Data Exchange (ETDEWEB)

    Zamecnik, J.; Koopman, D.

    2012-04-09

    The Defense Waste Processing Facility has experienced significant issues with the stripping and recovery of mercury in the Chemical Processing Cell (CPC). The stripping rate has been inconsistent, often resulting in extended processing times to remove mercury to the required endpoint concentration. The recovery of mercury in the Mercury Water Wash Tank has never been high, and has decreased significantly since the Mercury Water Wash Tank was replaced after the seventh batch of Sludge Batch 5. Since this time, essentially no recovery of mercury has been seen. Pertinent literature was reviewed, previous lab-scale data on mercury stripping and recovery was examined, and new lab-scale CPC Sludge Receipt and Adjustment Tank (SRAT) runs were conducted. For previous lab-scale data, many of the runs with sufficient mercury recovery data were examined to determine what factors affect the stripping and recovery of mercury and to improve closure of the mercury material balance. Ten new lab-scale SRAT runs (HG runs) were performed to examine the effects of acid stoichiometry, sludge solids concentration, antifoam concentration, form of mercury added to simulant, presence of a SRAT heel, operation of the SRAT condenser at higher than prototypic temperature, varying noble metals from none to very high concentrations, and higher agitation rate. Data from simulant runs from SB6, SB7a, glycolic/formic, and the HG tests showed that a significant amount of Hg metal was found on the vessel bottom at the end of tests. Material balance closure improved from 12-71% to 48-93% when this segregated Hg was considered. The amount of Hg segregated as elemental Hg on the vessel bottom was 4-77% of the amount added. The highest recovery of mercury in the offgas system generally correlated with the highest retention of Hg in the slurry. Low retention in the slurry (high segregation on the vessel bottom) resulted in low recovery in the offgas system. High agitation rates appear to result in lower

  11. Results for the second quarter 2014 tank 50 WAC slurry sample chemical and radionuclide contaminants

    International Nuclear Information System (INIS)

    Bannochie, C.

    2014-01-01

    This report details the chemical and radionuclide contaminant results for the characterization of the 2014 Second Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC) in effect at that time. Information from this characterization will be used by DWPF & Saltstone Facility Engineering (DSFE) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System

  12. Results for the Third Quarter 2014 Tank 50 WAC slurry sample: Chemical and radionuclide contaminants

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, Charles L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-01-08

    This report details the chemical and radionuclide contaminant results for the characterization of the 2014 Third Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC) in effect at that time.1 Information from this characterization will be used by DWPF & Saltstone Facility Engineering (DSFE) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System.

  13. Results For The Fourth Quarter 2014 Tank 50 WAC Slurry Sample: Chemical And Radionuclide Contaminants

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-30

    This report details the chemical and radionuclide contaminant results for the characterization of the Calendar Year (CY) 2014 Fourth Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC) in effect at that time. Information from this characterization will be used by DWPF & Saltstone Facility Engineering (DSFE) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System.

  14. Results For The Third Quarter 2013 Tank 50 WAC Slurry Sample

    Energy Technology Data Exchange (ETDEWEB)

    Bannochie, Christopher J.

    2013-11-26

    This report details the chemical and radionuclide contaminant results for the characterization of the 2013 Third Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC) in effect at that time. Information from this characterization will be used by DWPF & Saltstone Facility Engineering (DSFE) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System.

  15. Where Radiobiology Began in Russia: A Physician’s Perspective

    Science.gov (United States)

    2010-09-01

    White Archipelago”) (Gubarev, 2004): “Visiting a number of production facilities that worked with plutonium and polonium - 210 , I was struck by the...its owners. The wide array of food on the table was all vegetarian fare. I found out for the first time that both of them were Orthodox Christians...was ordinary food poisoning. After several days’ treatment, the workers returned to the area. Later, after changes in their skin and blood

  16. Commercial regional incinerator facility for treatment of low-level radioactive waste

    International Nuclear Information System (INIS)

    Sauer, R.E.

    1984-01-01

    In 1981, US Ecology, Inc. began studies on the feasibility of constructing and operating a regional radioactive waste incinerator facility. In December, 1982, US Ecology requested turnkey quotations from several vendors for engineering, procurement, and construction of the new facility. After technical and commercial evaluations, a contract was awarded to Associated Technologies, Inc., of Charlotte, North Carolina, in June, 1983. In June, 1984, US Ecology made a public announcement that they were studying two sites in North Carolina for location of the facility. This same month, they submitted their permit application for a radioactive material license to the North Carolina Department of Human Resources. The facility will accept wastes from power reactors, medical and research institutions and other industrial users, and will incinerate dry solid waste, pathological waste, scintillation fluids, and turbine oils. The incinerator will be a dual chamber controlled air design, rated at 600 lbs/hr, with a venturi scrubber, packed column, HEPA, and charcoal filters for pollution control. The stack will have a continuous monitor

  17. Commercial regional incinerator facility for treatment of low-level radioactive waste

    International Nuclear Information System (INIS)

    Sauer, R.E.; Jessop, D.; Associated Technologies, Inc., Charlotte, NC)

    1985-01-01

    In 1981, US Ecology, Inc. began studies on the feasibility of constructing and operating a regional radioactive waste incinerator facility. In December, 1982, US Ecology requested turnkey quotations from several vendors for engineering, procurement, and construction of the new facility. After technical and commercial evaluations, a contract was awarded to Associated Technologies, Inc., of Charlotte, North Carolina, in June, 1983. In June, 1984, US Ecology made a public announcement that they were studying two sites in North Carolina for location of the facility. This same month, they submitted their permit application for a radioactive material license to the North Carolina Department of Human Resources. The facility will accept wastes from power reactors, medical and research institutions and other industrial users, and will incinerate dry solid waste, pathological waste, scintillation fluids, and turbine oils. The incinerator will be a dual chamber controlled air design, rated at 600 lbs/h, with a venturi scrubber, packed column, HEPA, and charcoal filters for pollution control. The stack will have a continuous monitor. 4 figs

  18. Commercial regional incinerator facility for treatment of low-level radioactive waste

    International Nuclear Information System (INIS)

    Sauer, R.E.; Jessop, D.

    1985-01-01

    In 1981, US Ecology, Inc. began studies on the feasibility of constructing and operating a regional radioactive waste incinerator facility. In December, 1982, US Ecology requested turnkey quotations from several vendors for engineering, procurement, and construction of the new facility. After technical and commercial evaluations, a contract was awarded to Associated Technologies, Inc., of Charlotte, North Carolina, in June, 1983. In June, 1984, US Ecology made a public announcement that they were studying two sites in North Carolina for location of the facility. This same month, they submitted their permit application for a radioactive material license to the North Carolina Department of Human Resources. The facility will accept wastes from power reactors, medical and research institutions and other industrial users, and will incinerate dry solid waste, pathological waste, scintillation fluids, and turbine oils. The incinerator will be a dual chamber controlled air design, rated at 600 lbs/h, with a venturi scrubber, packed column, HEPA, and charcoal filters for pollution control. The stack will have a continuous monitor

  19. 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

  20. Learning and education on environmental radioactivity by residents of Rokkasho Site for the spent fuel recycling facilities

    International Nuclear Information System (INIS)

    Kawauchi, Kiye; Itoh, Natsuko; Ishikawa, Tomiye; Nihonyanagi, Haruko; Aratani, Michi

    2005-01-01

    The neutron criticality accident at the JCO, a private company for nuclear fuel processing facilities in Tokai has drastically changed minds and attitudes of residents toward environmental radioactivity. The accident happened on September 30, 1999. Before the accident the residents of the Rokkasho Site were not anxious about environmental radioactivity, because they thought the facilities were safe enough concerning containment policy of the radioactivity inside the facilities. Residents, however, had not been taught on a neutron. It is an unfamiliar radiation for them. So, they promptly learnt on neutrons, and some of them began the fixed point measurement of neutrons at the nearest site of the Spent Fuel Recycling Facilities of Rokkasho by the help of Prof. Kazuhisa. Komura, Kanazawa University. Members of the Reading Cicle, Rokkasho Culture Society, mainly women, learnt measurements of environmental radioactivity using simplified counters for alpha-, beta-, and gamma-ray from natural radioactive elements and prepared various kinds of environmental samples. After learning of environmental radioactivity, they began educational activities on the environmental radioactivity for boys and girls in the region. Monitoring of environmental radioactivity is performed by different institutions and with their purposes. Here is reported learning of environmental radioactivity by the residents and education of environmental radioactivity toward the young. Even with the simplest counters, we think that the monitoring of environmental radioactivity by the residents themselves is the royal road to the safety of the regional society. (author)

  1. Facility Design and Health Management Program at the Sinnhuber Aquatic Research Laboratory.

    Science.gov (United States)

    Barton, Carrie L; Johnson, Eric W; Tanguay, Robert L

    2016-07-01

    The number of researchers and institutions moving to the utilization of zebrafish for biomedical research continues to increase because of the recognized advantages of this model. Numerous factors should be considered before building a new or retooling an existing facility. Design decisions will directly impact the management and maintenance costs. We and others have advocated for more rigorous approaches to zebrafish health management to support and protect an increasingly diverse portfolio of important research. The Sinnhuber Aquatic Research Laboratory (SARL) is located ∼3 miles from the main Oregon State University campus in Corvallis, Oregon. This facility supports several research programs that depend heavily on the use of adult, larval, and embryonic zebrafish. The new zebrafish facility of the SARL began operation in 2007 with a commitment to build and manage an efficient facility that diligently protects human and fish health. An important goal was to ensure that the facility was free of Pseudoloma neurophilia (Microsporidia), which is very common in zebrafish research facilities. We recognize that there are certain limitations in space, resources, and financial support that are institution dependent, but in this article, we describe the steps taken to build and manage an efficient specific pathogen-free facility.

  2. A state-of-the art on decommissioning of nuclear facilities in Japan

    International Nuclear Information System (INIS)

    Park, Seung Kook; Kim, Hee Reyoung; Chung, Un Soo; Jung, Ki Jung

    2002-05-01

    While proceeding the KRR-1 and 2 decommissioning project, we are carried out study for the state of the art on decommissioning of nuclear facilities in Japan. Also, we are studied for the research reactors and commercial power plant that has the object of decommissioning, and for the government and the organization related on decommissioning operation. We are investigated for decommissioning activities of nuclear facilities achieved by JAERI, and collected the information and data for decommissioning techniques and computational system through the JPDR(Japan Power Demonstration Reactor) decommissioning activities. Such techniques are applying for Tokai Power Station began the decommissioning project from last year, and for Fugen Nuclear Power Station to be planned the decommissioning from 2003. Recent techniques for decommissioning was acquired by direct contact. The status of the treatment for decommissioning waste and the disposal facility for the very low-level radioactive concrete wastes was grasped

  3. A commercial regional incinerator facility for treatment of low-level radioactive waste

    International Nuclear Information System (INIS)

    Sauer, R.E.; Jessop, D.T.

    1986-01-01

    In 1981, US Ecology, Inc. began studies on the feasibility of constructing and operating a regional radioactive waste incinerator facility. In December, 1982, US Ecology requested turnkey quotations from several vendors for engineering, procurement, and construction of the new facility. After technical and commercial evaluations, a contract was awarded to Associated Technologies, Inc., of Charlotte, North Carolina, in June, 1983. In June, 1984, US Ecology made a public announcement that they were studying two sites in North Carolina for location of the facility. This same month, they submitted their permit application fro a radioactive material license to the North Carolina Department of Human Resources. The facility will accept wastes from power reactors, medical and research institutions and other industrial users, and will incinerate dry solid waste, pathological waste, scintillation fluids, and turbine oils. The incinerator will be a dual chamber controlled air design, rated at 600 lbs/hr, with a venturi scrubber, packed column, HEPA, and charcoal filters for pollution control. The stack will have a continuous monitor

  4. Sodium removal from the grapples of the fuel handling facility of Joyo

    Energy Technology Data Exchange (ETDEWEB)

    Mukaibo, R; Matsuno, Y; Sato, I; Yoneda, Y; Sato, H [O-arai Engineering Centre, PNC, Ibaraki-ken, Tokio (Japan)

    1978-08-01

    Sodium removal from the grapples of the fuel handling facility of 'JOYO' is done in alcohol. The operations of the cleaning facility started as the functional tests of the fuel handling facility began. Since then, criticality test and low power tests had been done and during this period, sodium removal from the grapples, after a certain amount of time in use, were done. In order to lessen the time for the cleaning process for the grapples of the machines inside the containment vessel, demineralized water concentration in the alcohol was gained to as much as 10% and good results were obtained. On the other hand, there were very small amounts of sodium on the grapples of the machine used outside the containment vessel and direct charging of demineralized water into the cleaning pot was done experimentally, also with good results. In this report, the sodium removal experience of the grapples before power up tests and some remarks on the improvements of the facility for the future are presented. (author)

  5. Lessons Learned From The 200 West Pump And Treatment Facility Construction Project At The US DOE Hanford Site - A Leadership For Energy And Environmental Design (LEED) Gold-Certified Facility

    International Nuclear Information System (INIS)

    Dorr, Kent A.; Ostrom, Michael J.; Freeman-Pollard, Jhivaun R.

    2012-01-01

    CH2M Hill Plateau Remediation Company (CHPRC) designed, constructed, commissioned, and began operation of the largest groundwater pump and treatment facility in the U.S. Department of Energy's (DOE) nationwide complex. This one-of-a-kind groundwater pump and treatment facility, located at the Hanford Nuclear Reservation Site (Hanford Site) in Washington State, was built in an accelerated manner with American Recovery and Reinvestment Act (ARRA) funds and has attained Leadership in Energy and Environmental Design (LEED) GOLD certification, which makes it the first non-administrative building in the DOE Office of Environmental Management complex to earn such an award. There were many contractual, technical, configuration management, quality, safety, and LEED challenges associated with the design, procurement, construction, and commissioning of this $95 million, 52,000 ft groundwater pump and treatment facility. This paper will present the Project and LEED accomplishments, as well as Lessons Learned by CHPRC when additional ARRA funds were used to accelerate design, procurement, construction, and commissioning of the 200 West Groundwater Pump and Treatment (2W PandT) Facility to meet DOE's mission of treating contaminated groundwater at the Hanford Site with a new facility by June 28, 2012

  6. Lessons Learned From The 200 West Pump And Treatment Facility Construction Project At The US DOE Hanford Site - A Leadership For Energy And Environmental Design (LEED) Gold-Certified Facility

    Energy Technology Data Exchange (ETDEWEB)

    Dorr, Kent A. [CH2M HILL Plateau Remediation Company, Richland, WA (United States); Ostrom, Michael J. [CH2M HILL Plateau Remediation Company, Richland, WA (United States); Freeman-Pollard, Jhivaun R. [CH2M HILL Plateau Remediation Company, Richland, WA (United States)

    2012-11-14

    CH2M Hill Plateau Remediation Company (CHPRC) designed, constructed, commissioned, and began operation of the largest groundwater pump and treatment facility in the U.S. Department of Energy's (DOE) nationwide complex. This one-of-a-kind groundwater pump and treatment facility, located at the Hanford Nuclear Reservation Site (Hanford Site) in Washington State, was built in an accelerated manner with American Recovery and Reinvestment Act (ARRA) funds and has attained Leadership in Energy and Environmental Design (LEED) GOLD certification, which makes it the first non-administrative building in the DOE Office of Environmental Management complex to earn such an award. There were many contractual, technical, configuration management, quality, safety, and LEED challenges associated with the design, procurement, construction, and commissioning of this $95 million, 52,000 ft groundwater pump and treatment facility. This paper will present the Project and LEED accomplishments, as well as Lessons Learned by CHPRC when additional ARRA funds were used to accelerate design, procurement, construction, and commissioning of the 200 West Groundwater Pump and Treatment (2W P&T) Facility to meet DOE's mission of treating contaminated groundwater at the Hanford Site with a new facility by June 28, 2012.

  7. Lessons Learned from the 200 West Pump and Treatment Facility Construction Project at the US DOE Hanford Site - A Leadership for Energy and Environmental Design (LEED) Gold-Certified Facility

    Energy Technology Data Exchange (ETDEWEB)

    Dorr, Kent A.; Ostrom, Michael J.; Freeman-Pollard, Jhivaun R.

    2013-01-11

    CH2M Hill Plateau Remediation Company (CHPRC) designed, constructed, commissioned, and began operation of the largest groundwater pump and treatment facility in the U.S. Department of Energy’s (DOE) nationwide complex. This one-of-a-kind groundwater pump and treatment facility, located at the Hanford Nuclear Reservation Site (Hanford Site) in Washington State, was built to an accelerated schedule with American Recovery and Reinvestment Act (ARRA) funds. There were many contractual, technical, configuration management, quality, safety, and Leadership in Energy and Environmental Design (LEED) challenges associated with the design, procurement, construction, and commissioning of this $95 million, 52,000 ft groundwater pump and treatment facility to meet DOE’s mission objective of treating contaminated groundwater at the Hanford Site with a new facility by June 28, 2012. The project team’s successful integration of the project’s core values and green energy technology throughout design, procurement, construction, and start-up of this complex, first-of-its-kind Bio Process facility resulted in successful achievement of DOE’s mission objective, as well as attainment of LEED GOLD certification, which makes this Bio Process facility the first non-administrative building in the DOE Office of Environmental Management complex to earn such an award.

  8. The trail to Leduc began in the North

    Energy Technology Data Exchange (ETDEWEB)

    McKenzie-Brown, P.

    2000-06-01

    A milestone event in the history of the petroleum industry in Canada occurred in 1914 with the expedition down the Mackenzie River by Dr. T. O. Bosworth, a British geologist, which rivals in importance two better known events, namely the Dingman No. 1 discovery which began disgorging wet gas at Turner Valley in 1914, and the beginning of the modern era with the discovery of oil at Leduc, Alberta in 1947. The Bosworth expedition was commissioned by two Calgary businessmen to investigate the petroleum potential of northern Alberta and beyond, and to stake the most promising claims. That Dr. Bosworth did not disappoint his sponsors is evident from his 70-page long report which he produced upon his return, indicating excellent exploration prospects in three general regions, namely the Mackenzie River between Old Fort Good Hope and Fort Norman, the Tar Springs District on the Great Slave Lake, and the Tar Sand District on the Athabasca River. Exploration was postponed by the exigencies of World War I, but Imperial Oil drilled on one of Bosworth's claims after the war ended and oil was found in 1920. Because of lack of infrastructure to get the oil to major markets, development was lagging until after the Japanese attack on Pearl Harbor in 1941, when Norman Wells was developed with American help, primarily to supply oil to the Pacific Fleet. Imperial drilled, while construction crews built a 1,000-km oil pipeline over the Mackenzie Mountains to a newly constructed refiner in Whitehorse. Despite a total cost to the U.S. taxpayers of $134 million, the Canol project (as it was known) contributed little to the war effort. Total production was 1.98 million barrels of oil, of which 46,000 barrels were spilled along the poorly constructed pipeline. Refined petroleum product output was just 866,670 barrels. The appalling disposal and clean up practices eventually led to having the Whitehorse site declared an environmentally contaminated site in 1998. The Maxwell Tar Pit is

  9. Start-up of commercial high level waste vitrification facilities at La Hague

    International Nuclear Information System (INIS)

    Sombret, C.; Jouan, A.; Fournier, W.; Alexandre, D.; Leroy, L.

    1991-01-01

    The paper describes industial experience gained in France for vitrification of fission products generated by spent fuel reprocessing. The continuous vitrification process developed by CEA, SGN and COGEMA is outlined and Marcoule Vitrification Facility (AVM), with output results since start-up of hot operation in June 1978, briefly presented. Vitrification of high-level liquid waste has now entered an industrial phase at La Hague with R7 and T7 facilities. R7 and T7 have each been designed to process FP solutions generated by reprocessing LWR fuel with an initial enrichment of 3.5% and a discharge burn-up of 33,000 MWd/t. R7 active operations began on June, 1989. This facility is now vitrifying the backlog of fission products resulting from the existing UP2 reprocessing plant, which is being currently extended. Scheduled to start early in 1992, T7 will vitrify the fission products, dissolution fines and sodium-rich solutions issuing from UP3 plant

  10. Lessons Learned from the 200 West Pump and Treatment Facility Construction Project at the US DOE Hanford Site - A Leadership for Energy and Environmental Design (LEED) Gold-Certified Facility - 13113

    Energy Technology Data Exchange (ETDEWEB)

    Dorr, Kent A.; Freeman-Pollard, Jhivaun R.; Ostrom, Michael J. [CH2M HILL Plateau Remediation Company, P.O. Box 1600, MSIN R4-41, 99352 (United States)

    2013-07-01

    CH2M Hill Plateau Remediation Company (CHPRC) designed, constructed, commissioned, and began operation of the largest groundwater pump and treatment facility in the U.S. Department of Energy's (DOE) nationwide complex. This one-of-a-kind groundwater pump and treatment facility, located at the Hanford Nuclear Reservation Site (Hanford Site) in Washington State, was built to an accelerated schedule with American Recovery and Reinvestment Act (ARRA) funds. There were many contractual, technical, configuration management, quality, safety, and Leadership in Energy and Environmental Design (LEED) challenges associated with the design, procurement, construction, and commissioning of this $95 million, 52,000 ft groundwater pump and treatment facility to meet DOE's mission objective of treating contaminated groundwater at the Hanford Site with a new facility by June 28, 2012. The project team's successful integration of the project's core values and green energy technology throughout design, procurement, construction, and start-up of this complex, first-of-its-kind Bio Process facility resulted in successful achievement of DOE's mission objective, as well as attainment of LEED GOLD certification (Figure 1), which makes this Bio Process facility the first non-administrative building in the DOE Office of Environmental Management complex to earn such an award. (authors)

  11. Lessons Learned from the 200 West Pump and Treatment Facility Construction Project at the US DOE Hanford Site - A Leadership for Energy and Environmental Design (LEED) Gold-Certified Facility - 13113

    International Nuclear Information System (INIS)

    Dorr, Kent A.; Freeman-Pollard, Jhivaun R.; Ostrom, Michael J.

    2013-01-01

    CH2M Hill Plateau Remediation Company (CHPRC) designed, constructed, commissioned, and began operation of the largest groundwater pump and treatment facility in the U.S. Department of Energy's (DOE) nationwide complex. This one-of-a-kind groundwater pump and treatment facility, located at the Hanford Nuclear Reservation Site (Hanford Site) in Washington State, was built to an accelerated schedule with American Recovery and Reinvestment Act (ARRA) funds. There were many contractual, technical, configuration management, quality, safety, and Leadership in Energy and Environmental Design (LEED) challenges associated with the design, procurement, construction, and commissioning of this $95 million, 52,000 ft groundwater pump and treatment facility to meet DOE's mission objective of treating contaminated groundwater at the Hanford Site with a new facility by June 28, 2012. The project team's successful integration of the project's core values and green energy technology throughout design, procurement, construction, and start-up of this complex, first-of-its-kind Bio Process facility resulted in successful achievement of DOE's mission objective, as well as attainment of LEED GOLD certification (Figure 1), which makes this Bio Process facility the first non-administrative building in the DOE Office of Environmental Management complex to earn such an award. (authors)

  12. 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

  13. Characterization and decant of Tank 42H sludge sample ESP-200

    International Nuclear Information System (INIS)

    Hay, M.S.

    2000-01-01

    DWPF Engineering requested that the Savannah River Technology Center (SRTC) provide a demonstration of the DWPF flowsheet on sludge from Tank 42H in the Shielded Cell facility. A 5 liter sample of the Tank 42H sludge (ESP-200), obtained with the tank contents fully mixed, arrived at SRTC on January 20, 1998. This report details receipt of the 5 liter sample at SRTC, the decant of the sample, and the characterization of the pre- and post-decant Tank 42H sludge. Evaluation of the measured composition of the supernate indicates Sample ESP-200 became diluted approximately 20 percent by volume prior to receipt. This dilution complicates the relationship of the characterization of Post-Decant ESP-200 to the current contents of Tank 42H. For the purposes of modeling the current tank contents of Tank 42H, this report provides an estimated composition based on analytical data of recent samples from Tank 42H

  14. Characterization and decant of Tank 42H sludge sample ESP-200

    Energy Technology Data Exchange (ETDEWEB)

    Hay, M.S.

    2000-04-25

    DWPF Engineering requested that the Savannah River Technology Center (SRTC) provide a demonstration of the DWPF flowsheet on sludge from Tank 42H in the Shielded Cell facility. A 5 liter sample of the Tank 42H sludge (ESP-200), obtained with the tank contents fully mixed, arrived at SRTC on January 20, 1998. This report details receipt of the 5 liter sample at SRTC, the decant of the sample, and the characterization of the pre- and post-decant Tank 42H sludge. Evaluation of the measured composition of the supernate indicates Sample ESP-200 became diluted approximately 20 percent by volume prior to receipt. This dilution complicates the relationship of the characterization of Post-Decant ESP-200 to the current contents of Tank 42H. For the purposes of modeling the current tank contents of Tank 42H, this report provides an estimated composition based on analytical data of recent samples from Tank 42H.

  15. Behavior of mercury in the formic acid vent condenser. Interim report

    International Nuclear Information System (INIS)

    Zamecnik, J.R.

    1994-01-01

    (This report relates to the Defense Waste Processing Facility.) The concentrations of mercury at the FAFC inlet and exit were measured during the BL1 and PX6 runs of the Integrated DWPF Melter System (IDMS) with the HEME bypassed and without the ammonia scrubber. The results show that mercury concentrations of approximately 2.6-12.7 (mean = 6.2) times saturation occur at the FAFC exit. The concentration of mercury at the SRAT condenser exit was found to be 10 times the saturation value. FAVC exit mercury concentrations of 6.2 times saturation would result in DWPF emitting up to 438 lb/yr of mercury at 100 percent attainments, which is in excess of the permit limit of 175 lb/yr. However, operation of the FAVC with the HEME should reduce the mercury emissions. The addition of the ammonia scrubbers should also reduce the mercury emissions since the nitric acid used to scrub ammonia should also scrub mercury

  16. 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

  17. 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

  18. Implementation of the DYMAC system at the new Los Alamos Plutonium Processing Facility. Phase II report

    Energy Technology Data Exchange (ETDEWEB)

    Malanify, J.J.; Amsden, D.C.

    1982-08-01

    The DYnamic Materials ACcountability System - called DYMAC - performs accountability functions at the new Los Alamos Plutonium Processing Facility where it began operation when the facility opened in January 1978. A demonstration program, DYMAC was designed to collect and assess inventory information for safeguards purposes. It accomplishes 75% of its design goals. DYMAC collects information about the physical inventory through deployment of nondestructive assay instrumentation and video terminals throughout the facility. The information resides in a minicomputer where it can be immediately sorted and displayed on the video terminals or produced in printed form. Although the capability now exists to assess the collected data, this portion of the program is not yet implemented. DYMAC in its present form is an excellent tool for process and quality control. The facility operator relies on it exclusively for keeping track of the inventory and for complying with accountability requirements of the US Department of Energy.

  19. Implementation of the DYMAC system at the new Los Alamos Plutonium Processing Facility. Phase II report

    International Nuclear Information System (INIS)

    Malanify, J.J.; Amsden, D.C.

    1982-08-01

    The DYnamic Materials ACcountability System - called DYMAC - performs accountability functions at the new Los Alamos Plutonium Processing Facility where it began operation when the facility opened in January 1978. A demonstration program, DYMAC was designed to collect and assess inventory information for safeguards purposes. It accomplishes 75% of its design goals. DYMAC collects information about the physical inventory through deployment of nondestructive assay instrumentation and video terminals throughout the facility. The information resides in a minicomputer where it can be immediately sorted and displayed on the video terminals or produced in printed form. Although the capability now exists to assess the collected data, this portion of the program is not yet implemented. DYMAC in its present form is an excellent tool for process and quality control. The facility operator relies on it exclusively for keeping track of the inventory and for complying with accountability requirements of the US Department of Energy

  20. 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

  1. 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

  2. Description and Operational Experiences of the Engineering Test Facility - Helium Technology (ETF-HT)

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Yang Mingde; Bo Hanliang; Duan Riqqiang; Zhu Hongye

    2014-01-01

    This paper presents the configuration of the Engineering Test Facility - Helium Technology (ETF-HT) and the information of its key components and subsystems, which is located in the Changping campus of Tsinghua University. The ETF-HT facility began to be constructed in Jan. 2009. The main objective of the facility is to test and verify the thermo-hydraulic performance of one full-sized modular unit of HTR-PM helically coiled SG assembly. In the ETF-HT facility, electricity energy is used to heat the loop helium, centrifugal blower is used to circulate the helium medium, and the heat sink is one would-tested SG module. Up to now, except for the tested SG module, preheater and hot gas duct under way of construction, the other components has been installed in situ. Via the temporary connection of the installed components, the preliminary operation of the loop has been carried out to test its performances as can be done, which include the loop leak tightness, blower pneumatic performance and electrical heater at partial thermal load. (author)

  3. Results for the First, Second, and Third Quarter Calendar Year 2015 Tank 50H WAC slurry samples chemical and radionuclide contaminants

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-18

    This report details the chemical and radionuclide contaminant results for the characterization of the Calendar Year (CY) 2015 First, Second, and Third Quarter sampling of Tank 50H for the Saltstone Waste Acceptance Criteria (WAC) in effect at that time. Information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering (D&S-FE) to support the transfer of low-level aqueous waste from Tank 50H to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50H Waste Characterization System. Previous memoranda documenting the WAC analyses results have been issued for these three samples.

  4. The reactor and cold neutron research facility at NIST

    Energy Technology Data Exchange (ETDEWEB)

    Prask, H J; Rowe, J M [Reactor Radiation Division, National Institute of Standards and Technology, Gaithersburg, MD (United States)

    1992-07-01

    The NIST Reactor (NBSR) is a 20 MW research reactor located at the Gaithersburg, MD site, and has been in operation since 1969. It services 26 thermal neutron facilities which are used for materials science, chemical analysis, nondestructive evaluation, neutron standards work, and irradiations. In 1987 the Department of Commerce and NIST began development of the CNRF - a $30M National Facility for cold neutron research -which will provide fifteen new experimental stations with capabilities currently unavailable in this country. As of May 1992, four of the planned seven guides and a cold port were installed, eight cold neutron experimental stations were operational, and the Call for Proposals for the second cycle of formally-reviewed guest-researcher experiments had been sent out. Some details of the performance of instrumentation are described, along with the proposed design of the new hydrogen cold source which will replace the present D{sub 2}O/H{sub 2}O ice cold source. (author)

  5. The reactor and cold neutron research facility at NIST

    International Nuclear Information System (INIS)

    Prask, H.J.; Rowe, J.M.

    1992-01-01

    The NIST Reactor (NBSR) is a 20 MW research reactor located at the Gaithersburg, MD site, and has been in operation since 1969. It services 26 thermal neutron facilities which are used for materials science, chemical analysis, nondestructive evaluation, neutron standards work, and irradiations. In 1987 the Department of Commerce and NIST began development of the CNRF - a $30M National Facility for cold neutron research -which will provide fifteen new experimental stations with capabilities currently unavailable in this country. As of May 1992, four of the planned seven guides and a cold port were installed, eight cold neutron experimental stations were operational, and the Call for Proposals for the second cycle of formally-reviewed guest-researcher experiments had been sent out. Some details of the performance of instrumentation are described, along with the proposed design of the new hydrogen cold source which will replace the present D 2 O/H 2 O ice cold source. (author)

  6. IFMIF-KEP. International fusion materials irradiation facility key element technology phase report

    International Nuclear Information System (INIS)

    2003-03-01

    The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D-Li neutron source designed to produce an intense neutron field that will simulate the neutron environment of a D-T fusion reactor. IFMIF will provide a neutron flux equivalent to 2 MW/m 2 , 20 dpa/y in Fe, in a volume of 500 cm 3 and will be used in the development and qualification of materials for fusion systems. The design activities of IFMIF are performed under an IEA collaboration which began in 1995. In 2000, a three-year Key Element Technology Phase (KEP) of IFMIF was undertaken to reduce the key technology risk factors. This KEP report describes the results of the three-year KEP activities in the major project areas of accelerator, target, test facilities and design integration. (author)

  7. Directions in low-level radioactive waste management. Low level-radioactive waste disposal: currently operating commercial facilities

    International Nuclear Information System (INIS)

    1983-09-01

    This publication discusses three commercial facilities that receive and dispose of low-level radioactive waste. The facilities are located in Barnwell, South Carolina; Beatty, Nevada; and Richland, Washington. All three facilities initiated operations in the 1960s. The three facilities have operated without such major problems as those which led to the closure of three other commercial disposal facilities located in the United States. The Beatty site could be closed in 1983 as a result of a Nevada Board of Health ruling that renewal of the site license would be inimical to public health and safety. The site remains open pending federal and state court hearings, which began in January 1983, to resolve the Board of Health ruling. The three sites may also be affected by NRC's 10 CFR Part 61 regulations, but the impact of those regulations, issued in December 1982, has not yet been assessed. This document provides detailed information on the history and current status of each facility. This information is intended, primarily, to assist state officials - executive, legislative, and agency - in planning for, establishing, and managing low-level waste disposal facilities. 12 references

  8. 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)

  9. NENIMF: Northeast National Ion Microprobe Facility - A Multi-User Facility for SIMS Microanalysis

    Science.gov (United States)

    Layne, G. D.; Shimizu, N.

    2002-12-01

    The MIT-Brown-Harvard Regional Ion Microprobe Facility was one of the earliest multi-user facilities enabled by Dan Weill's Instrumentation and Facilities Program - and began with the delivery of a Cameca IMS 3f ion microprobe to MIT in 1978. The Northeast National Ion Microprobe Facility (NENIMF) is the direct descendant of this original facility. Now housed at WHOI, the facility incorporates both the original IMS 3f, and a new generation, high transmission-high resolution instrument - the Cameca IMS 1270. Purchased with support from NSF, and from a consortium of academic institutions in the Northeast (The American Museum of Natural History, Brown University, The Lamont-Doherty Earth Observatory, MIT, Rensselaer Polytechnic Institute, WHOI) - this latest instrument was delivered and installed during 1996. NENIMF continues to be supported by NSF EAR I&F as a multi-user facility for geochemical research. Work at NENIMF has extended the original design strength of the IMS 1270 for microanalytical U-Pb zircon geochronology to a wide variety of novel and improved techniques for geochemical research. Isotope microanalysis for studies in volcanology and petrology is currently the largest single component of facility activity. This includes the direct measurement of Pb isotopes in melt inclusions, an application developed at NENIMF, which is making an increasingly significant contribution to our understanding of basalt petrogenesis. This same technique has also been extended to the determination of Pb isotopes in detrital feldspar grains, for the study of sedimentary provenance and tectonics of the Himalayas and other terrains. The determination of δ11B in volcanic melt inclusions has also proven to be a powerful tool in the modeling of subduction-related magmatism. The recent development of δ34S and δ37Cl determination in glasses is being applied to studies of the behavior of these volatile elements in both natural and experimental systems. Other recent undertakings

  10. The National Ignition Facility

    International Nuclear Information System (INIS)

    Hogan, W.J.; Moses, E.; Warner, B.; Sorem, M.; Soures, J.M.

    2001-01-01

    The National Ignition Facility (NIF) is the largest construction project ever undertaken at Lawrence Livermore National Laboratory (LLNL). NIF consists of 192 forty-centimeter-square laser beams and a 10-m-diameter target chamber. NIF is being designed and built by an LLNL-led team from Los Alamos National Laboratory, Sandia National Laboratories, the University of Rochester, and LLNL. Physical construction began in 1997. The Laser and Target Area Building and the Optics Assembly Building were the first major construction activities, and despite several unforeseen obstacles, the buildings are now 92% complete and have been done on time and within cost. Prototype component development and testing has proceeded in parallel. Optics vendors have installed full-scale production lines and have done prototype production runs. The assembly and integration of the beampath infrastructure has been reconsidered and a new approach has been developed. This paper will discuss the status of the NIF project and the plans for completion. (author)

  11. Potential Benefits to the Philippines of a Nuclear Facility

    International Nuclear Information System (INIS)

    Asuncion-Astronomo, A.; Romallosa, K.M.D.; Olivares, R.U.

    2015-01-01

    During the late 1950’s, the Philippines was one of the many countries which began the pursuit of the beneficial applications of atomic energy. With the commissioning of the first Philippine Research Reactor (PRR-1) which attained its first criticality in 1963, our country had the capability for radioisotope production, activation analysis of materials, irradiation studies and various opportunities for basic and applied nuclear science research. The Nuclear Power Plant (PNNP-1) in training plant operators and regulators for the first Philippine Nuclear Power Plant (PNPP-1) in Bataan, which was eventually mothballed in 1986. It is thus unfortunate that the only operating nuclear facility in the country, the PRR-1 encountered technical problems during an upgrade and was shut down in 1988. The problem was not resolved and eventually led to the decommissioning of the PRR-1 in 2005. Without an operating nuclear facility available in the country, the number of personnel knowledgeable and skilled in reactor and nuclear science and engineering has greatly declined and lagged behind our counterparts. This has been the situation for more than two decaded and can only be addressed if the country decides to put up a new nuclear facility. It is acknowledged that putting up a nuclear facility is a major undertaking which requires careful planning, preparation and investment. Thus, a decision by any country to embark on this poster, we will provide an overview of the many potential benefits as well as challenges of establishing a new research reactor and/or accelerator facility in the country. The global distribution, comparisons, capabilities and the different application of these facilities will presented as well.(author)

  12. 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.

  13. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Zamecnik, J.; Choi, A.

    2010-08-18

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

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

    International Nuclear Information System (INIS)

    Zamecnik, J.; Choi, A.

    2009-01-01

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

  16. 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

  17. Sludge batch 9 simulant runs using the nitric-glycolic acid flowsheet

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, D. P. [Savannah River Site (SRS), Aiken, SC (United States); Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States); Brandenburg, C. H. [Savannah River Site (SRS), Aiken, SC (United States); Luther, M. C. [Savannah River Site (SRS), Aiken, SC (United States); Newell, J. D. [Savannah River Site (SRS), Aiken, SC (United States); Woodham, W. H. [Savannah River Site (SRS), Aiken, SC (United States)

    2016-11-01

    Testing was completed to develop a Sludge Batch 9 (SB9) nitric-glycolic acid chemical process flowsheet for the Defense Waste Processing Facility’s (DWPF) Chemical Process Cell (CPC). CPC simulations were completed using SB9 sludge simulant, Strip Effluent Feed Tank (SEFT) simulant and Precipitate Reactor Feed Tank (PRFT) simulant. Ten sludge-only Sludge Receipt and Adjustment Tank (SRAT) cycles and four SRAT/Slurry Mix Evaporator (SME) cycles, and one actual SB9 sludge (SRAT/SME cycle) were completed. As has been demonstrated in over 100 simulations, the replacement of formic acid with glycolic acid virtually eliminates the CPC’s largest flammability hazards, hydrogen and ammonia. Recommended processing conditions are summarized in section 3.5.1. Testing demonstrated that the interim chemistry and Reduction/Oxidation (REDOX) equations are sufficient to predict the composition of DWPF SRAT product and SME product. Additional reports will finalize the chemistry and REDOX equations. Additional testing developed an antifoam strategy to minimize the hexamethyldisiloxane (HMDSO) peak at boiling, while controlling foam based on testing with simulant and actual waste. Implementation of the nitric-glycolic acid flowsheet in DWPF is recommended. This flowsheet not only eliminates the hydrogen and ammonia hazards but will lead to shorter processing times, higher elemental mercury recovery, and more concentrated SRAT and SME products. The steady pH profile is expected to provide flexibility in processing the high volume of strip effluent expected once the Salt Waste Processing Facility starts up.

  18. 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

  19. IFMIF-KEP. International fusion materials irradiation facility key element technology phase report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-03-01

    The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D-Li neutron source designed to produce an intense neutron field that will simulate the neutron environment of a D-T fusion reactor. IFMIF will provide a neutron flux equivalent to 2 MW/m{sup 2}, 20 dpa/y in Fe, in a volume of 500 cm{sup 3} and will be used in the development and qualification of materials for fusion systems. The design activities of IFMIF are performed under an IEA collaboration which began in 1995. In 2000, a three-year Key Element Technology Phase (KEP) of IFMIF was undertaken to reduce the key technology risk factors. This KEP report describes the results of the three-year KEP activities in the major project areas of accelerator, target, test facilities and design integration. (author)

  20. Air pollution control at a DOE facility

    International Nuclear Information System (INIS)

    Curn, B.L.

    1995-11-01

    The Department of Energy (DOE) plutonium production program Produced some of the greatest scientific and engineering accomplishments of all time. It is remarkable to consider the accomplishments of the Manhattan Project. The Reactor on the Hanford Site, the first production reactor in the world, began operation only 13 months after the start of construction. The DOE nuclear production program was also instrumental in pioneering other fields such as health physics an radiation monitoring. The safety record of these installations is remarkable considering that virtually every significant accomplishment was on the technological threshold of the time. One other area that the DOE Facilities pioneered was the control of radioactive particles and gases emitted to the atmosphere. The high efficiency particulate air filter (HEPA) was a development that provided high collection efficiencies of particulates to protect workers and the public. The halogen and noble gases also were of particular concern. Radioactive iodine is captured by adsorption on activated carbon or synthetic zeolites. Besides controlling radioncuclide air pollution, DOE facilities are concerned with other criteria pollutants and hazardous air pollutant emissions. The Hanford Site encompasses all those air pollution challenges

  1. Independent Technical Review of In-Tank Precipitation (ITP) at the Savannah River Site

    International Nuclear Information System (INIS)

    1993-06-01

    An Independent Technical Review of In-Tank Precipitation (ITP) and Extended Sludge Processing (ESP) at the Savannah River Site (SRS) was carried out in March, 1993. The review focused on ITP/ESP equipment and chemical processes, integration of ITP/ESP within the High Level Waste (HLW) and Defense Waste Processing Facility (DWPF) systems, and management and regulatory concerns. Following the ITR executive summary, this report includes: Chapter I--summary assessment; Chapter II--recommendations; and Chapter III--technical evaluations

  2. Design of agitation systems in Bingham slurries by pilot simulation

    International Nuclear Information System (INIS)

    Nielsen, M.G.

    1987-01-01

    A method was required to determine the optimum agitator speed needed to produce overall motion of the Defense Waste Processing Facility (DWPF) high-level waste slurries in remote process cell vessels. Project schedule and limited process space required an accurate determination of agitator horsepower and size without the benefit of full-scale testing. The small scale testing of unique clear rheologically similar fluid is described along with tests and scale-up procedures. 2 refs., 3 figs

  3. Virological failure of staggered and simultaneous treatment interruption in HIV patients who began Efavirenz-based regimens after allergic reactions to nevirapine

    Directory of Open Access Journals (Sweden)

    Siripassorn Krittaecho

    2013-01-01

    Full Text Available Abstract Objective The objective of this work was to study the virological outcomes associated with two different types of treatment interruption strategies in patients with allergic reactions to nevirapine (NVP. We compared the virological outcomes of (1 HIV-1-infected patients who discontinued an initial NVP-based regimen because of cutaneous allergic reactions to NVP; different types of interruption strategies were used, and second-line regimen was based on efavirenz (EFV; and (2 HIV-1-infected patients who began an EFV-based regimen as a first-line therapy (controls. Methods This retrospective cohort included patients who began an EFV-based regimen, between January 2002 and December 2008, as either an initial regimen or as a subsequent regimen after resolving a cutaneous allergic reaction against an initial NVP-based regimen. The study ended in March 2010. The primary outcome was virological failure, which was defined as either (a two consecutive plasma HIV-1 RNA levels >400 copies/mL or (b a plasma HIV-1 RNA level >1,000 copies/mL plus any genotypic resistance mutation. Results A total of 559 patients were stratified into three groups: (a Simultaneous Interruption, in which the subjects simultaneously discontinued all the drugs in an NVP-based regimen following an allergic reaction (n=161; (b Staggered Interruption, in which the subjects discontinued NVP treatment while continuing nucleoside reverse transcriptase inhibitor (NRTI backbone therapy for a median of 7 days (n=82; and (c Control, in which the subjects were naïve to antiretroviral therapy (n=316. The overall median follow-up time was 43 months. Incidence of virological failure in Simultaneous Interruption was 12.9 cases per 1,000 person-years, which trended toward being higher than the incidences in Staggered Interruption (5.4 and Control (6.6. However, differences were not statistically significant. Conclusions Among the patients who had an acute allergic reaction to first

  4. DEMOLITION OF HANFORD'S 232-Z WASTE INCINERATION FACILITY

    International Nuclear Information System (INIS)

    LLOYD, E.R.

    2006-01-01

    The 232-Z Plutonium Incinerator Facility was a small, highly alpha-contaminated, building situated between three active buildings located in an operating nuclear complex. Approximately 500 personnel worked within 250 meters (800 ft) of the structure and expectations were that the project would neither impact plant operations nor result in any restrictions when demolition was complete. Precision demolition and tight controls best describe the project. The team used standard open-air demolition techniques to take the facility to slab-on-grade. Several techniques were key to controlling contamination and confining it to the demolition area: spraying fixatives before demolition began; using misting systems, frequently applying fixatives, and using a methodical demolition sequence and debris load-out process. Detailed air modeling was done before demolition to determine necessary facility source-term levels, establish radiological boundaries, and confirm the adequacy of the proposed demolition approach. By only removing the major source term in equipment, HEPA filters, gloveboxes, and the like, and leaving fixed contamination on the walls, ceilings and floors, the project showed considerable savings and reduced worker hazards and exposure. The ability to perform this demolition safely and without the spread of contamination provides confidence that similar operations can be performed successfully. By removing the major source terms, fixing the remaining contamination in the building, and using controlled demolition and contamination control techniques, similar structures can be demolished cost effectively and safely

  5. Regional-interstate site-review procedure: low-level radioactive waste disposal facility

    International Nuclear Information System (INIS)

    1983-09-01

    The attributes of the Southern States Energy Board (SSEB) enable it to view federal/state interface problem areas from a perspective that can be uniquely constructive. The board is sensitive to the interests of both federal and state levels of government since it is composed of member states with common regional interests and confirmed by federal legislative action. It has been most effective when exercising a leadership role in finding procedures and practices that use the resources of both levels of government that are mutually supportive and nonduplicative. SSEB began an NRC-funded effort in that direction related to nuclear power plant siting in June 1975, entitled Regional-Interstate Nuclear Facility Siting Procedure Demonstration Project. SSEB approached the problem by working with interested states to analyze various elements of the licensing process, in particular with NEPA review procedures for interstate coordination where potential impacts extend beyond a single state and where the facility serves an interstate or regional need. SSEB also served as a catalyst in the development of a region-wide nuclear facility siting procedure that could improve the effectiveness and timeliness of the regulatory process

  6. Westinghouse-DOE integration: Meeting the challenge

    International Nuclear Information System (INIS)

    Price, S.V.

    1992-01-01

    The Westinghouse Electric Corporation (WEC) is in a unique position to affect national environmental management policy approaching the 21st Century. Westinghouse companies are management and operating contractors (MOC,s) at several environmentally pivotal government-owned, contractor operated (GOCO) facilities within the Department of Energy's (DOE's) nuclear defense complex. One way the WEC brings its companies together is by activating teams to solve specific DOE site problems. For example, one challenging issue at DOE facilities involves the environmentally responsible, final disposal of transuranic and high-level nuclear wastes (TRUs and HLWS). To address these disposal issues, the DOE supports two Westinghouse-based task forces: The TRU Waste Acceptance Criteria Certification Committee (WACCC) and the HLW Vitrification Committee. The WACCC is developing methods to characterize an estimated 176,287 cubic meters of retrievably stored TRUs generated at DOE production sites. Once characterized, TRUs could be safely deposited in the WIPP repository. The Westinghouse HLW Vitrification Committee is dedicated to assess appropriate methods to process an estimated 380,702 cubic meters of HLWs currently stored in underground storage tanks (USTs). As planned, this processing will involve segregating, and appropriately treating, low level waste (LLW) and HLW tank constituents for eventual disposal. The first unit designed to process these nuclear wastes is the SRS Defense Waste Processing Facility (DWPF). Initiated in 1973, the DWPF project is scheduled to begin operations in 1991 or 1992. Westinghouse companies are also working together to achieve appropriate environmental site restoration at DOE sites via the GOCO Environmental Restoration Committee

  7. Facility status and progress of the INEL's WERF MLLW and LLW incinerator

    International Nuclear Information System (INIS)

    Conley, D.; Corrigan, S.

    1996-01-01

    The Idaho National Engineering Laboratory's (INEL) Waste Experimental Reduction Facility (WERF) incinerator began processing beta/gamma- emitting low-level waste (LLW) in September 1984. A Resource Conservation and Recovery Act (RCRA) trial burn for the WERF incinerator was conducted in 1986, and in 1989 WERF began processing (hazardous and low-level radioactive) waste known as mixed low-level waste (MLLW). On February 14, 1991 WERF operations were suspended to improve operating procedures and configuration management. On July 12, 1995, WERF initiated incineration of LLW; and on September 20, 1995 WERF resumed its primary mission of incinerating MLLW. MLLW incineration is proceeding under RCRA interim status. State of Idaho issuance of the Part B permit is one of the State's highest permitting priorities. The State of Idaho's Division of Environmental Quality is reviewing the permit application along with a revised trial burn plan that was also submitted with the application. The trial burn has been proposed to be performed in 1996 to demonstrate compliance with the current incinerator guidance. This paper describes the experiences and problems associated with WERF's operations, incineration of MLLW, and the RCRA Part B Permit Application. Some of the challenges that have been overcome include waste characterization, waste repackaging, repackaged waste storage, and implementation of RCRA interim status requirements. A number of challenges remain. They include revision of the RCRA Part B Permit Application and the Trial Burn Plan in response to comments from the state permit application reviewers as well as facility and equipment upgrades required to meet RCRA Permitted Status

  8. 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

  9. Development of a mixed waste management facility at the Nevada Test Site

    International Nuclear Information System (INIS)

    Dodge, R.L.; Brich, R.F.

    1988-01-01

    The U.S. Department of Energy (DOE) produces radioactive low-level wastes (LLW) which contain hazardous components as identified by 40 Code of Federal Regulations (CFR) 261. Management of those mixed wastes (MW) requires compliance with U.S.Environmental Protection Agency (EPA) regulations for hazardous wastes and DOE regulations for LLW. In 1988, DOE's Nevada Operations Office (NV) began disposing of MW at the Nevada Test Site (NTS) under interim status as authorized by the state of Nevada. MW disposal is limited to Pit 3 while operating under interim status. This paper discusses how preparations for operation of a separate mixed waste management facility (MWMF) are underway. Those preparations include revising the NTS Part B Permit application, developing a MW certification program, developing and operating a vadose zone monitoring system, preparing an Environmental Assessment (EA), developing protocols for analysis of MW, and facility design and construction

  10. PROJECTIZING AN OPERATING NUCLEAR FACILITY

    International Nuclear Information System (INIS)

    Adams, N

    2007-01-01

    This paper will discuss the evolution of an operations-based organization to a project-based organization to facilitate successful deactivation of a major nuclear facility. It will describe the plan used for scope definition, staff reorganization, method estimation, baseline schedule development, project management training, and results of this transformation. It is a story of leadership and teamwork, pride and success. Workers at the Savannah River Site's (SRS) F Canyon Complex (FCC) started with a challenge--take all the hazardous byproducts from nearly 50 years of operations in a major, first-of-its-kind nuclear complex and safely get rid of them, leaving the facility cold, dark, dry and ready for whatever end state is ultimately determined by the United States Department of Energy (DOE). And do it in four years, with a constantly changing workforce and steadily declining funding. The goal was to reduce the overall operating staff by 93% and budget by 94%. The facilities, F Canyon and its adjoined sister, FB Line, are located at SRS, a 310-square-mile nuclear reservation near Aiken, S.C., owned by DOE and managed by Washington Group International subsidiary Washington Savannah River Company (WSRC). These facilities were supported by more than 50 surrounding buildings, whose purpose was to provide support services during operations. The radiological, chemical and industrial hazards inventory in the old buildings was significant. The historical mission at F Canyon was to extract plutonium-239 and uranium-238 from irradiated spent nuclear fuel through chemical processing. FB Line's mission included conversion of plutonium solutions into metal, characterization, stabilization and packaging, and storage of both metal and oxide forms. The plutonium metal was sent to another DOE site for use in weapons. Deactivation in F Canyon began when chemical separations activities were completed in 2002, and a cross-functional project team concept was implemented to successfully

  11. Evaluation of replacement tritium facility (RTF) compliance with DOE safety goals using probabilistic consequence assessment methodology

    International Nuclear Information System (INIS)

    O'Kula, K.R.; East, J.M.; Moore, M.L.

    1993-01-01

    The Savannah River Site (SRS), operated by the Westinghouse Savannah River Company (WSRC) for the US Department of Energy (DOE), is a major center for the processing of nuclear materials for national defense, deep-space exploration, and medical treatment applications in the United States. As an integral part of the DOE's effort to modernize facilities, implement improved handling and processing technology, and reduce operational risk to the general public and onsite workers, transition of tritium processing at SRS from the Consolidated Tritium Facility to the Replacement Tritium Facility (RTF) began in 1993. To ensure that operation of new DOE facilities such as RTF present minimum involuntary and voluntary risks to the neighboring public and workers, indices of risk have been established to serve as target levels or safety goals of performance for assessing nuclear safety. These goals are discussed from a historical perspective in the initial part of this paper. Secondly, methodologies to quantify risk indices are briefly described. Lastly, accident, abnormal event, and normal operation source terms from RTF are evaluated for consequence assessment purposes relative to the safety targets

  12. 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

  13. 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

  14. Waste and Simulant Precipitation Issues

    International Nuclear Information System (INIS)

    Steele, W.V.

    2000-01-01

    As Savannah River Site (SRS) personnel have studied methods of preparing high-level waste for vitrification in the Defense Waste Processing Facility (DWPF), questions have arisen with regard to the formation of insoluble waste precipitates at inopportune times. One option for decontamination of the SRS waste streams employs the use of an engineered form of crystalline silicotitanate (CST). Testing of the process during FY 1999 identified problems associated with the formation of precipitates during cesium sorption tests using CST. These precipitates may, under some circumstances, obstruct the pores of the CST particles and, hence, interfere with the sorption process. In addition, earlier results from the DWPF recycle stream compatibility testing have shown that leaching occurs from the CST when it is stored at 80 C in a high-pH environment. Evidence was established that some level of components of the CST, such as silica, was leached from the CST. This report describes the results of equilibrium modeling and precipitation studies associated with the overall stability of the waste streams, CST component leaching, and the presence of minor components in the waste streams

  15. Plan for reevaluation of NRC policy on decommissioning of nuclear facilities

    International Nuclear Information System (INIS)

    1978-12-01

    The present decommissioning regulations contained in Sections 50.33(f) and 50.82 of 10 CFR part 50 require applicants for power reactor operating licenses to demonstrate that they can obtain the funds needed to meet both operating costs and estimated costs of shutdown and decommissioning. The development of detailed, specific decommissioning plans for nuclear power plants is not currently required until the licensee seeks to terminate his operating license. Recognizing that the current generation of large commercial reactors and supporting nuclear facilities would substantially increase the need for future decommissionings, the NRC staff began an in-depth review and reevaluation of NRC's regulatory approach to decommissioning in 1975. The Nuclear Regulatory Commission is now considering development of a more explicit overall policy for nuclear facility decommissioning and amending its regulations in 10 CFR Parts 30, 40, 50, and 70 to include more specific guidance on decommissioning criteria for production and utilization facility licensees and byproduct, source, and special nuclear material licensees. In response to comments from the public and states, and to information gained during the initial stage of execution of the plan, several modifications of the plan are now required. The revised overall report sets forth in detail the current NRC staff plan for the development of an overall NRC policy on decommissioning of nuclear facilities

  16. 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

  17. 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

  18. Refurbishment and Automation of the Thermal/Vacuum Facilities at the Goddard Space Flight Center

    Science.gov (United States)

    Donohue, John T.; Johnson, Chris; Ogden, Rick; Sushon, Janet

    1998-01-01

    The thermal/vacuum facilities located at the Goddard Space Flight Center (GSFC) have supported both manned and unmanned space flight since the 1960s. Of the 11 facilities, currently 10 of the systems are scheduled for refurbishment and/or replacement as part of a 5-year implementation. Expected return on investment includes the reduction in test schedules, improvements in the safety of facility operations, reduction in the complexity of a test and the reduction in personnel support required for a test. Additionally, GSFC will become a global resource renowned for expertise in thermal engineering, mechanical engineering and for the automation of thermal/vacuum facilities and thermal/vacuum tests. Automation of the thermal/vacuum facilities includes the utilization of Programmable Logic Controllers (PLCs) and the use of Supervisory Control and Data Acquisition (SCADA) systems. These components allow the computer control and automation of mechanical components such as valves and pumps. In some cases, the chamber and chamber shroud require complete replacement while others require only mechanical component retrofit or replacement. The project of refurbishment and automation began in 1996 and has resulted in the computer control of one Facility (Facility #225) and the integration of electronically controlled devices and PLCs within several other facilities. Facility 225 has been successfully controlled by PLC and SCADA for over one year. Insignificant anomalies have occurred and were resolved with minimal impact to testing and operations. The amount of work remaining to be performed will occur over the next four to five years. Fiscal year 1998 includes the complete refurbishment of one facility, computer control of the thermal systems in two facilities, implementation of SCADA and PLC systems to support multiple facilities and the implementation of a Database server to allow efficient test management and data analysis.

  19. SLUDGE BATCH 5 ACCEPTANCE EVALUATION RADIONUCLIDE CONCENTRATIONS IN TANK 51 SB5 QUALIFICATION SAMPLE PREPARED AT SRNL

    International Nuclear Information System (INIS)

    Bannochie, C; Ned Bibler, N; David Diprete, D

    2008-01-01

    Presented in this report are radionuclide concentrations required as part of the program of qualifying Sludge Batch Five (SB5) for processing in the Defense Waste Processing Facility (DWPF). Part of this SB5 material is currently in Tank 51 being washed and prepared for transfer to Tank 40. The acceptance evaluation needs to be completed prior to the transfer of the material in Tank 51 to Tank 40 to complete the formation of SB5. The sludge slurry in Tank 40 has already been qualified for DWPF and is currently being processed as SB4. The radionuclide concentrations were measured or estimated in the Tank 51 SB5 Qualification Sample prepared at Savannah River National Laboratory (SRNL). This sample was prepared from the three liter sample of Tank 51 sludge slurry taken on March 21, 2008. The sample was delivered to SRNL where it was initially characterized in the Shielded Cells. Under direction of the Liquid Waste Organization it was then modified by five washes, six decants, an addition of Pu/Be from Canyon Tank 16.4, and an addition of NaNO2. This final slurry now has a composition expected to be similar to that of the slurry in Tank 51 after final preparations have been made for transfer of that slurry to Ta Determining the radionuclide concentrations in this Tank 51 SB5 Qualification Sample is part of the work requested in Technical Task Request (TTR) No. HLW-DWPF-TTR-2008-0010. The work with this qualification sample is covered by a Task Technical and Quality Assurance Plan and an Analytical Study Plan. The radionuclides included in this report are needed for the DWPF Radiological Program Evaluation, the DWPF Waste Acceptance Criteria (TSR/WAC) Evaluation, and the DWPF Solid Waste Characterization Program (TTR Task 2). Radionuclides required to meet the Waste Acceptance Product Specifications (TTR Task 5) will be measured at a later date after the slurry from Tank 51 has been transferred to Tank 40. Then a sample of the as-processed SB5 will be taken and

  20. Remote handling facility and equipment used for space truss assembly

    International Nuclear Information System (INIS)

    Burgess, T.W.

    1987-01-01

    The ACCESS truss remote handling experiments were performed at Oak Ridge National Laboratory's (ORNL's) Remote Operation and Maintenance Demonstration (ROMD) facility. The ROMD facility has been developed by the US Department of Energy's (DOE's) Consolidated Fuel Reprocessing Program to develop and demonstrate remote maintenance techniques for advanced nuclear fuel reprocessing equipment and other programs of national interest. The facility is a large-volume, high-bay area that encloses a complete, technologically advanced remote maintenance system that first began operation in FY 1982. The maintenance system consists of a full complement of teleoperated manipulators, manipulator transport systems, and overhead hoists that provide the capability of performing a large variety of remote handling tasks. This system has been used to demonstrate remote manipulation techniques for the DOE, the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan, and the US Navy in addition to the National Aeronautics and Space Administration. ACCESS truss remote assembly was performed in the ROMD facility using the Central Research Laboratory's (CRL) model M-2 servomanipulator. The model M-2 is a dual-arm, bilateral force-reflecting, master/slave servomanipulator which was jointly developed by CRL and ORNL and represents the state of the art in teleoperated manipulators commercially available in the United States today. The model M-2 servomanipulator incorporates a distributed, microprocessor-based digital control system and was the first successful implementation of an entirely digitally controlled servomanipulator. The system has been in operation since FY 1983. 3 refs., 2 figs

  1. Construction and Commissioning of PAL-XFEL Facility

    Directory of Open Access Journals (Sweden)

    In Soo Ko

    2017-05-01

    Full Text Available The construction of Pohang Accelerator Laboratory X-ray Free-Electron Laser (PAL-XFEL, a 0.1-nm hard X-ray free-electron laser (FEL facility based on a 10-GeV S-band linear accelerator (LINAC, is achieved in Pohang, Korea by the end of 2016. The construction of the 1.11 km-long building was completed by the end of 2014, and the installation of the 10-GeV LINAC and undulators started in January 2015. The installation of the 10-GeV LINAC, together with the undulators and beamlines, was completed by the end of 2015. The commissioning began in April 2016, and the first lasing of the hard X-ray FEL line was achieved on 14 June 2016. The progress of the PAL-XFEL construction and its commission are reported here.

  2. Technical Facilities Management, Loan Pool, and Calibration

    Science.gov (United States)

    Smith, Jacob

    2011-01-01

    My work at JPL for the SURF program began on June 11, 2012 with the Technical Facilities Management group (TFM). As well as TFM, I worked with Loan Pool and Metrology to help them out with various tasks. Unlike a lot of other interns, I did not have a specific project rather many different tasks to be completed over the course of the 10 weeks.The first task to be completed was to sort through old certification reports in 6 different boxes to locate reports that needed to be archived into a digital database. There were no reports within these boxes that needed to be archived but rather were to be shredded. The reports went back to the early 1980's and up to the early 2000's. I was looking for reports dated from 2002 to 2012

  3. Reactor cover gas monitoring at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Bechtold, R.A.; Holt, F.E.; Meadows, G.E.; Schenter, R.E.

    1986-09-01

    The Fast Flux Test Facility (FFTF) is a 400-megawatt (thermal) sodium-cooled reactor designed for irradiation testing of fuels, materials and components for LMRs. It is operated by the Westinghouse Hanford Company for the US Department of Energy on the government-owned Hanford reservation near Richland, Washington. The first 100-day operating cycle began in April 1982 and the eighth operating cycle was completed in July 1986. Argon is used as the cover gas for all sodium systems at the plant. A program for cover gas monitoring has been in effect since the start of sodium fill in 1978. The argon is supplied to the FFTF by a liquid argon Dewar System and used without further purification

  4. HEPCloud, a New Paradigm for HEP Facilities: CMS Amazon Web Services Investigation

    Energy Technology Data Exchange (ETDEWEB)

    Holzman, Burt [Fermilab; Bauerdick, Lothar A.T. [Fermilab; Bockelman, Brian [Nebraska U.; Dykstra, Dave [Fermilab; Fisk, Ian [New York U.; Fuess, Stuart [Fermilab; Garzoglio, Gabriele [Fermilab; Girone, Maria [CERN; Gutsche, Oliver [Fermilab; Hufnagel, Dirk [Fermilab; Kim, Hyunwoo [Fermilab; Kennedy, Robert [Fermilab; Magini, Nicolo [Fermilab; Mason, David [Fermilab; Spentzouris, Panagiotis [Fermilab; Tiradani, Anthony [Fermilab; Timm, Steve [Fermilab; Vaandering, Eric W. [Fermilab

    2017-09-29

    Historically, high energy physics computing has been performed on large purpose-built computing systems. These began as single-site compute facilities, but have evolved into the distributed computing grids used today. Recently, there has been an exponential increase in the capacity and capability of commercial clouds. Cloud resources are highly virtualized and intended to be able to be flexibly deployed for a variety of computing tasks. There is a growing nterest among the cloud providers to demonstrate the capability to perform large-scale scientific computing. In this paper, we discuss results from the CMS experiment using the Fermilab HEPCloud facility, which utilized both local Fermilab resources and virtual machines in the Amazon Web Services Elastic Compute Cloud. We discuss the planning, technical challenges, and lessons learned involved in performing physics workflows on a large-scale set of virtualized resources. In addition, we will discuss the economics and operational efficiencies when executing workflows both in the cloud and on dedicated resources.

  5. How life began.

    Science.gov (United States)

    Cloud, P

    1986-11-01

    Study of the origin of life has become a legitimate scientific inquiry, with an international, multidisciplinary membership and a cogent body of data. Experiments involving plausible early Earth conditions and biogeochemical analyses of carbonaceous meteorites imply a variety of available starting molecules. Biogeological evidence indicates microbial beginnings about 3800 million years (3.8 aeons) ago. By then the known universe had been in existence for perhaps 15 aeons and galaxies abundant for ten. Conditions suitable for the origin of life may require a long prior cosmic evolution. The natural origin of life on the early Earth is now widely agreed upon but not the pathways. The beginnings of catalysis, replication and a functional cell remain moot. Much discussion has centered on the templating role that crystals such as clays and zeolites might have played in prebiotic evolution. Recent discovery of the catalytic and replicative functions of RNA recommend it as the key molecule in the transition from chemical to biological evolution. Copyright © 1986. Published by Elsevier Ltd.

  6. Tomorrow Began Yesterday

    Directory of Open Access Journals (Sweden)

    Konstantin Lidin

    2014-01-01

    Full Text Available The phenomenon of the sixtiers was prepared by the previous historical period. The period after the World War II comprises a fundamental change of the world order – from a multipolar world to a confrontation of two superpowers and two ideological systems, and, at the same time, formation of a complex of international organizations on a global scale. In this context, the Soviet architecture made a sharp turn from Stalin’s Empire style to an extreme ascetism – the continuation of constructivism of the early XXth century. The Irkutsk architectural school, unlike the main flow of the 1960s, developed the style of Neo-Brutalism. The article draws parallels between Neo-Brutalism of the Irkutsk school and “a severe style” of the Soviet pictorial art of the same period.

  7. Critique of Hanford Waste Vitrification Plant off-gas sampling requirements

    International Nuclear Information System (INIS)

    Goles, R.W.

    1996-03-01

    Off-gas sampling and monitoring activities needed to support operations safety, process control, waste form qualification, and environmental protection requirements of the Hanford Waste Vitrification Plant (HWVP) have been evaluated. The locations of necessary sampling sites have been identified on the basis of plant requirements, and the applicability of Defense Waste Processing Facility (DWPF) reference sampling equipment to these HWVP requirements has been assessed for all sampling sites. Equipment deficiencies, if present, have been described and the bases for modifications and/or alternative approaches have been developed

  8. The Berkeley Accelerator Space Effects (BASE) Facility - A new mission for the 88-Inch Cyclotron at LBNL

    International Nuclear Information System (INIS)

    McMahan, M.A.

    2005-01-01

    In FY04, the 88-Inch Cyclotron began a new operating mode that supports a local research program in nuclear science, R and D in accelerator technology and a test facility for the National Security Space (NSS) community (the US Air Force and NRO). The NSS community (and others on a cost recovery basis) can take advantage of both the light- and heavy-ion capabilities of the cyclotron to simulate the space radiation environment. A significant portion of this work involves the testing of microcircuits for single event effects. The experimental areas within the building that are used for the radiation effects testing are now called the Berkeley Accelerator Space Effects (BASE) Facility. Improvements to the facility to provide increased reliability, quality assurance and new capabilities are underway and will be discussed. These include a 16 A MeV 'cocktail' of beams for heavy ion testing, a neutron beam, more robust dosimetry, and other upgrades

  9. Manufacture and installation of reactor auxiliary facilities for advanced thermal prototype reactor 'Fugen'

    International Nuclear Information System (INIS)

    Kawahara, Toshio; Matsushita, Tadashi

    1977-01-01

    The facilities of reactor auxiliary systems for the advanced thermal prtotype reactor ''Fugen'' were manufactured in factories since 1972, and the installation at the site began in November, 1974. It was almost completed in March, 1977, except a part of the tests and inspections, therefore the outline of the works is reported. The ATR ''Fugen'' is a heavy water-moderated, boiling light water reactor, and its reactor auxiliary systems comprise mainly the facilities for handling heavy water, such as heavy water cooling system, heavy water cleaning system, poison supplying system, helium circulating system, helium cleaning system, and carbon dioxide system. The poison supplying system supplies liquid poison to the heavy water cooling system to absorb excess reactivity in the initial reactor core. The helium circulating system covers heavy water surface with helium to prevent the deterioration of heavy water and maintains heavy water level by pressure difference. The carbon dioxide system flows highly pure CO 2 gas in the space of pressure tubes and carandria tubes, and provides thermal shielding. The design, manufacture and installation of the facilities of reactor auxiliary systems, and the helium leak test, synthetic pressure test and total cleaning are explained. (Kako, I.)

  10. Marshall Space Flight Center's Impact Testing Facility Capabilities

    Science.gov (United States)

    Finchum, Andy; Hubbs, Whitney; Evans, Steve

    2008-01-01

    Marshall Space Flight Center s (MSFC) Impact Testing Facility (ITF) serves as an important installation for space and missile related materials science research. The ITF was established and began its research in spacecraft debris shielding in the early 1960s, then played a major role in the International Space Station debris shield development. As NASA became more interested in launch debris and in-flight impact concerns, the ITF grew to include research in a variety of impact genres. Collaborative partnerships with the DoD led to a wider range of impact capabilities being relocated to MSFC as a result of the closure of Particle Impact Facilities in Santa Barbara, California. The Particle Impact Facility had a 30 year history in providing evaluations of aerospace materials and components during flights through rain, ice, and solid particle environments at subsonic through hypersonic velocities. The facility s unique capabilities were deemed a "National Asset" by the DoD. The ITF now has capabilities including environmental, ballistic, and hypervelocity impact testing utilizing an array of air, powder, and two-stage light gas guns to accommodate a variety of projectile and target types and sizes. Numerous upgrades including new instrumentation, triggering circuitry, high speed photography, and optimized sabot designs have been implemented. Other recent research has included rain drop demise characterization tests to obtain data for inclusion in on-going model development. The current and proposed ITF capabilities range from rain to micrometeoroids allowing the widest test parameter range possible for materials investigations in support of space, atmospheric, and ground environments. These test capabilities including hydrometeor, single/multi-particle, ballistic gas guns, exploding wire gun, and light gas guns combined with Smooth Particle Hydrodynamics Code (SPHC) simulations represent the widest range of impact test capabilities in the country.

  11. Lessons Learned at Envirocare of Utah's Containerized Waste Facility (CWF): Dose Minimization Through ALARA Techniques and Tools

    International Nuclear Information System (INIS)

    Heckman, J.; Gardner, J.; Ledoux, M. R.

    2003-01-01

    Envirocare of Utah, Inc. (Envirocare) commenced operation of its Class A Containerized Waste Facility (CWF) on October 25, 2001. The opening of this facility began a new era for Envirocare, in that; their core business had always been low level, high volume, bulk radioactive waste. The CWF commenced operations to dispose of low level, low volume, high activity, containerized radioactive waste. Due to the potential for high dose rates on the waste disposal containers, the ALARA principle (As Low As Reasonably Achievable) plays an important role in the operation of the CWF and its mission to properly dispose of waste while minimizing doses to the workers, public, and the environment. This paper will enumerate some of the efforts made by the management and staff of the CWF that have contributed to significant dose reductions

  12. Irradiation Facilities at the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2005-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC) (formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens

  13. 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

  14. Determination of the impact of glycolate on ARP and MCU operations

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, K. M.L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Peters, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fondeur, F. F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Shehee, T. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Washington, A. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-05-17

    Savannah River Remediation (SRR) is evaluating an alternate flowsheet for the Defense Waste Processing Facility (DWPF) using glycolic acid as a reductant. An important aspect of the development of the glycolic acid flowsheet is determining if glycolate has any detrimental downstream impacts. Testing was performed to determine if there is any impact to the strontium and actinide sorption by monosodium titanate (MST) and modified monosodium titanate (mMST) or if there is an impact to the cesium removal, phase separation, or coalescer performance at the Modular Caustic-Side Solvent Extraction Processing Unit (MCU).

  15. 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

  16. 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

  17. Nuclear power-related facilities and neighboring land price: a case study on the Mutsu-Ogawara region, Japan.

    Science.gov (United States)

    Yamane, Fumihiro; Ohgaki, Hideaki; Asano, Kota

    2011-12-01

    From the perspective of risk, nuclear-power-related facilities (NPRFs) are often regarded as locally undesirable land use. However, construction of NPRFs contributes to social infrastructural improvement and job creation in the host communities. This raises a question: How large are these positive and negative effects? To approach this question from an economic viewpoint, we estimated the hedonic land price function for the Mutsu-Ogawara region of Japan from 1976 to 2004 and analyzed year-by-year fluctuations in land prices around the NPRFs located there. Land prices increased gradually in the neighborhood of the nuclear fuel cycle facilities (NFCFs) in Rokkasho Village, except for some falling (i) from 1982 to 1983 (the first official announcement of the project of construction came in 1983), (ii) from 1987 to 1988 (in 1988, the construction began and opposition movements against the project reached their peak), and (iii) from 1998 to 1999 (the pilot carry-in of spent fuels into the reprocessing plant began in 1998). Land prices around the Higashidori Nuclear Power Plant decreased during the period 1981-1982, when the Tohoku Electric Power Corp. and Tokyo Electric Power Corp. announced their joint construction plan. On the other hand, we obtained some results, even though not significant, indicating that land prices around Ohminato and Sekinehama harbors changed with the arrival and departure of the nuclear ship Mutsu, which suffered a radiation leak in 1974. © 2011 Society for Risk Analysis.

  18. LSST summit facility construction progress report: reacting to design refinements and field conditions

    Science.gov (United States)

    Barr, Jeffrey D.; Gressler, William; Sebag, Jacques; Seriche, Jaime; Serrano, Eduardo

    2016-07-01

    The civil work, site infrastructure and buildings for the summit facility of the Large Synoptic Survey Telescope (LSST) are among the first major elements that need to be designed, bid and constructed to support the subsequent integration of the dome, telescope, optics, camera and supporting systems. As the contracts for those other major subsystems now move forward under the management of the LSST Telescope and Site (T and S) team, there has been inevitable and beneficial evolution in their designs, which has resulted in significant modifications to the facility and infrastructure. The earliest design requirements for the LSST summit facility were first documented in 2005, its contracted full design was initiated in 2010, and construction began in January, 2015. During that entire development period, and extending now roughly halfway through construction, there continue to be necessary modifications to the facility design resulting from the refinement of interfaces to other major elements of the LSST project and now, during construction, due to unanticipated field conditions. Changes from evolving interfaces have principally involved the telescope mount, the dome and mirror handling/coating facilities which have included significant variations in mass, dimensions, heat loads and anchorage conditions. Modifications related to field conditions have included specifying and testing alternative methods of excavation and contending with the lack of competent rock substrate where it was predicted to be. While these and other necessary changes are somewhat specific to the LSST project and site, they also exemplify inherent challenges related to the typical timeline for the design and construction of astronomical observatory support facilities relative to the overall development of the project.

  19. 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

  20. Defense-Waste-Processing Faclity, Savannah River Plant, Aiken, SC: Draft environmental impact statement

    International Nuclear Information System (INIS)

    1981-09-01

    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 Energy (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

  1. 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

  2. Tritium surveillance around nuclear facilities in Japan

    International Nuclear Information System (INIS)

    Inoue, Y.; Kasida, Y.

    1978-01-01

    In order to measure the tritium levels in the environmental water around the nuclear facilities, the tritium surveillance program began in 1967 locally at Tsuruga and Mihama districts. Nowadays it has been expanded to the ten commercial nuclear power stations and three nuclear facilities. For samples whose tritium concentration is believed less than about 100 pCi/l, they were electrolytically enriched, and then counted by the liquid scintillation counter. Some of samples believed higher than 100 pCi/l were analysed without any enrichment by the low background liquid scintillation counters, Aloka LB 600 or Aloka LB 1. The results of each station are listed in Table. The sampling points corresponding to each results are shown in Figure. Tritium from the effluent was not reflected in all the land water and the tap water around the nuclear power stations and the nuclear facilities. Tritium concentration in rivers, streams, and reservoirs (pools) decreased exponentially from about 600 pCi/l in 1967 to about 150 pCi/l in 1972 at Tsuruga and Mihama, and 360 pCi/l in 1968 to 120 pCi/l in 1973 at Genkai, with the half life of about 2.5 years in both cases. After around 1972, tritium levels of river system in all districts of Japan kept nearly constant up to the end of 1975 and they were in the range from 100 to 300 pCi/l corresponding to the districts. Thereafter, it seems to start to decrease again in 1976. Sea water sampled at the intake of the station or on the seashore far from the outlet was regarded not to be influenced by the effluent from the nuclear reactors or facilities. Tritium concentration in these coastal waters decreased from 100 - 300 pCi/l in 1971 to 30 - 40 pCi/l in 1972 in Fukushima, Ibaraki and Fukui prefectures. (author)

  3. User's manual for BINIAC: A computer code to translate APET bins

    International Nuclear Information System (INIS)

    Gough, S.T.

    1994-03-01

    This report serves as the user's manual for the FORTRAN code BINIAC. BINIAC is a utility code designed to format the output from the Defense Waste Processing Facility (DWPF) Accident Progression Event Tree (APET) methodology. BINIAC inputs the accident progression bins from the APET methodology, converts the frequency from occurrences per hour to occurrences per year, sorts the progression bins, and converts the individual dimension character codes into facility attributes. Without the use of BINIAC, this process would be done manually at great time expense. BINIAC was written under the quality assurance control of IQ34 QAP IV-1, revision 0, section 4.1.4. Configuration control is established through the use of a proprietor and a cognizant users list

  4. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

  5. Decontamination and decommissioning of the initial engine test facility and the IET two-inch hot-waste line

    International Nuclear Information System (INIS)

    Stoll, F.E.

    1987-04-01

    The Initial Engine Test Decommissioning Project is described in this report. The Initial Engine Test facility was constructed and operated at the National Reactor Testing Station, now known as the Idaho National Engineering Laboratory, to support the Aircraft Nuclear Propulsion Program and the Systems for Nuclear Auxiliary Power Transient test program, circa 1950 through 1960s. Due to the severe nature of these nuclear test programs, a significant amount of radioactive contamination was deposited in various portions of the Initial Engine Test Facility. Characterizations, decision analyses, and plans for decontamination and decommissioning were prepared from 1982 through 1985. Decontamination and decommissioning activities were performed in such a way that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory remains. These decontamination and decommissioning activities began in 1985 and were completed in 1987. 13 figs

  6. Clean-out and Reuse of GCEP Facilities at Portsmouth

    International Nuclear Information System (INIS)

    Franz, William; Hickman, Mark; Wiehle, Kristi

    2008-01-01

    PORTS began operations in 1956 to enrich uranium for both civilian and military use. It operated under Goodyear Atomic Corporation and Lockheed-Martin as a government-owned contractor-operated facility until the formation of the United States Enrichment Corporation (USEC) as a government corporation in 1993. In 1998, USEC was privatized as a publicly traded corporation. USEC leases the Portsmouth and Paducah GDPs from the U.S. Department of Energy (DOE). Enrichment operations were terminated at Portsmouth in 2001, although USEC continues to lease and maintain the Portsmouth GDP in Cold Shutdown and conduct some DOE projects there. LPP found that removal of old equipment to allow reuse of a facility can present unexpected challenges. Classified components create significant logistics issues. In this case, teamwork and attention to detail by USEC, LATA/Parallax, and DOE resulted in solutions and success. Lessons learnt: - Transportation logistics are particularly important when shipping waste streams with special requirements. - Investment in extra equipment yields tangible benefits where other resources (cleared drivers) are scarce. - An early start to providing specially qualified drivers, in this case providing security clearances, is essential. Availability is limited, and the time required to qualify new drivers may be lengthy. - A dedicated communications station, rather than reliance on existing resources (shift superintendent), is invaluable

  7. 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

  8. Historical records of the electric power and cooling water systems for the experimental facilities in the I.P.P. (1964-1983)

    International Nuclear Information System (INIS)

    Kitajima, Miyoshi.

    1984-01-01

    Twenty years have passed, since experimental studies began in the IPP. This is a status report on the electric power and cooling water systems in this Institute. Historical records on consumption of the electric power and cooling water for experimental facilities are summarized, also. Valuable stories obtained through operation, control and periodic inspection of these systems are described, which will be helpful for construction of new systems in future. Finally problems on safety, reliability, and economy are commented. (author)

  9. Establishment of the Neutron Beam Research Facility at the OPAL Reactor

    International Nuclear Information System (INIS)

    Kennedy, S.J.; Robinson, R.A.

    2012-01-01

    Full text: Australia's first research reactor, HIFAR, reached criticality in January 1958. At that time Australia's main agenda was establishment of a nuclear power program. HIFAR operated for nearly 50 years, providing a firm foundation for the establishment of Australia's second generation research Reactor OPAL, which reached criticality in August 006. In HIFAR's early years a neutron beam facility was established for materials characterization, partly in aid of the nuclear energy agenda and partly in response to interest from Australia's scientific community. By the time Australia's nuclear energy program ceased (in the 1970s), radioisotope production and research had also been established at Lucas Heights. Also, by this time the neutron beam facility for scientific research had evolved into a major utilization programme, warranting establishment of an independent body to facilitate scientific access (the Australian Institute for Nuclear Science and Engineering). In HIFAR's lifetime, ANSTO established a radiopharmaceuticals service for the Australian medical community and NDT silicon production was also established and grew to maturity. So when time came to determine the strategy for nuclear research in Australia into the 21st century, it was clear that the replacement for HIFAR should be multipurpose, with major emphases on scientific applications of neutron beams and medical isotope production. With this strategy in mind, ANSTO set about to design and build OPAL with a world-class neutron beam facility, capable of supporting a large and diverse scientific research community. The establishment of the neutron beam facility became the mission of the Bragg Institute management team. This journey began in 1997 with establishment of a working budget, and reached its first major objective when OPAL reached 20 MW thermal power nearly one decade later (in 2006). The first neutron beam instruments began operation soon after (in 2007), and quickly proved themselves to be

  10. The Rock Characterization Facility

    International Nuclear Information System (INIS)

    Holmes, J.

    1994-01-01

    In 1989, UK Nirex began a programme of surface-based characterization of the geology and hydrogeology of a site at Sellafield to evaluate its suitability to host a deep repository for radioactive waste. The next major stage in site characterization will be the construction and operation of a Rock Characterization Facility (RCF). It will be designed to provide rock characterization information and scope for model validation to permit firmer assessment of long-term safety. It will also provide information needed to decide the detailed location, design and orientation of a repository and to inform repository construction methods. A three-phase programme is planned for the RCF. During each phase, testwork will steadily improve our geological, hydrogeological and geotechnical understanding of the site. The first phase will involve sinking two shafts. That will be preceded by the establishment of a network of monitoring boreholes to ensure that the impact of shaft sinking can be measured. This will provide valuable data for model validation. In phase two, initial galleries will be excavated, probably at a depth of 650 m below Ordnance datum, which will host a comprehensive suite of experiments. These galleries will be extended in phase three to permit access to most of the rock volume that would host the repository. (Author)

  11. Power Systems Development Facility

    Energy Technology Data Exchange (ETDEWEB)

    None

    2003-07-01

    This report discusses Test Campaign TC12 of the Kellogg Brown & Root, Inc. (KBR) Transport Gasifier train with a Siemens Westinghouse Power Corporation (SW) particle filter system at the Power Systems Development Facility (PSDF) located in Wilsonville, Alabama. The Transport Gasifier is an advanced circulating fluidized-bed reactor designed to operate as either a combustor or a gasifier using a particulate control device (PCD). While operating as a gasifier, either air or oxygen can be used as the oxidant. Test run TC12 began on May 16, 2003, with the startup of the main air compressor and the lighting of the gasifier start-up burner. The Transport Gasifier operated until May 24, 2003, when a scheduled outage occurred to allow maintenance crews to install the fuel cell test unit and modify the gas clean-up system. On June 18, 2003, the test run resumed when operations relit the start-up burner, and testing continued until the scheduled end of the run on July 14, 2003. TC12 had a total of 733 hours using Powder River Basin (PRB) subbituminous coal. Over the course of the entire test run, gasifier temperatures varied between 1,675 and 1,850 F at pressures from 130 to 210 psig.

  12. Strategy and plan for siting and licensing a Rocky Mountain low-level radioactive waste facility

    International Nuclear Information System (INIS)

    Whitman, M.

    1983-09-01

    In 1979, the States of Nevada and Washington temporarily closed their commercial low-level radioactive waste (LLW) disposal facilities and South Carolina, the only other state hosting such a facility, restricted the amount of waste it would accept. All three states then announced that they did not intend to continue the status quo of accepting all of the country's commercial low-level radioactive waste. Faced with this situation, other states began considering alternative LLW management and disposal options. In the Rocky Mountain region, this evolved into discussions for the development of an interstate compact to manage low-level waste. Inherent in this management plan was a strategy to site and license a new LLW disposal facility for the Rocky Mountain region. The Rocky Mountain Low-Level Radioactive Waste Compact was negotiated over the course of a year, with final agreement on the language of the compact agreed to in early 1982. States eligible to join the compact are Arizona, Colorado, Nevada, New Mexico, Utah, and Wyoming. Colorado adopted the compact into law in 1982, and Nevada, New Mexico and Wyoming adopted it in 1983. Utah has joined the Northwest Compact, although it may decide to join the Rocky Mountain Compact after a new disposal facility is developed for the region. Arizona has taken no action on the Rocky Mountain Compact

  13. WIPP Facility Work Plan for Solid Waste Management Units

    International Nuclear Information System (INIS)

    2000-01-01

    This Facility Work Plan (FWP) has been prepared as required by Module VII,Section VII.M.1 of the Waste Isolation Pilot Plant (WIPP) Hazardous Waste Permit, NM4890139088-TSDF (the Permit); (NMED, 1999a). This work plan describes the programmatic facility-wide approach to future investigations at Solid Waste Management Units (SWMUs) and Areas of Concern (AOCs) specified in the Permit. This FWP addresses the current Permit requirements. It uses the results of previous investigations performed at WIPP and expands the investigations as required by the Permit. As an alternative to the Resource Conservation and Recovery Act (RCRA) Facility Investigation (RFI) specified in Module VII of the Permit, current New Mexico Environment Department (NMED) guidance identifies an Accelerated Corrective Action Approach (ACAA) that may be used for any SWMU or AOC (NMED, 1998). This accelerated approach is used to replace the standard RFI Work Plan and Report sequence with a more flexible decision-making approach. The ACAA process allows a Facility to exit the schedule of compliance contained in the Facility's's Hazardous and Solid Waste Amendments (HSWA) permit module and proceed on an accelerated time frame. Thus, the ACAA process can be entered either before or after an RFI Work Plan. According to NMED's guidance, a facility can prepare an RFI Work Plan or Sampling and Analysis Plan (SAP) for any SWMU or AOC (NMED, 1998). Based on this guidance, a SAP constitutes an acceptable alternative to the RFI Work Plan specified in the Permit. The scope of work for the RFI Work Plan or SAP is being developed by the Permittees. The final content of the RFI Work Plan or SAP will be coordinated with the NMED for submittal on May 24, 2000. Specific project-related planning information will be included in the RFI Work Plan or SAP. The SWMU program at WIPP began in 1994 under U.S. Environmental Protection Agency (EPA) regulatory authority. NMED subsequently received regulatory authority from EPA. A

  14. 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

  15. New challenges in the safety assessment of radioactive waste storage and disposal facilities in the Russian Federation

    Energy Technology Data Exchange (ETDEWEB)

    Linge, I.; Utkin, S. [Nuclear Safety Inst. (IBRAE RAN), Moscow (Russian Federation)

    2014-07-01

    Russian radioactive waste (RW) management practice (disposal, in particular) is characterized by a number of features which makes it fundamentally different from the international one. The technologies used in the middle of the XX century became widespread even after the nuclear arms race was over. As a result: Industrial sites comprise a large number of old solid RW storage facilities and surface water reservoirs (ponds, lakes), which capacity varies from one to several hundred million cubic meters, storing liquid RW; Deep well injection of liquid RW into aquifers has been in practice since the early 1960's. Major changes aimed at addressing the accumulated problems began to occur only a decade ago. In 2008, a large-scale state nuclear legacy program was initiated, and in 2011, the framework act «On RW management» was passed. New tasks were set before the Russian nuclear industry for the purpose of establishing a unified state system for RW management. It was accompanied by a number of new challenges in the safety justification and calculation tools development. They are discussed in the paper with significant consideration to the existing nuclear legacy facilities; unique liquid radioactive waste storage and disposal facilities; and new-built disposal facilities. (author)

  16. High-Level Waste Systems Plan. Revision 7 (U)

    International Nuclear Information System (INIS)

    Brooke, J.N.; Gregory, M.V.; Paul, P.; Taylor, G.; Wise, F.E.; Davis, N.R.; Wells, M.N.

    1996-10-01

    This revision of the High-Level Waste (HLW) System Plan aligns SRS HLW program planning with the DOE Savannah River (DOE-SR) Ten Year Plan (QC-96-0005, Draft 8/6), which was issued in July 1996. The objective of the Ten Year Plan is to complete cleanup at most nuclear sites within the next ten years. The two key principles of the Ten Year Plan are to accelerate the reduction of the most urgent risks to human health and the environment and to reduce mortgage costs. Accordingly, this System Plan describes the HLW program that will remove HLW from all 24 old-style tanks, and close 20 of those tanks, by 2006 with vitrification of all HLW by 2018. To achieve these goals, the DWPF canister production rate is projected to climb to 300 canisters per year starting in FY06, and remain at that rate through the end of the program in FY18, (Compare that to past System Plans, in which DWPF production peaked at 200 canisters per year, and the program did not complete until 2026.) An additional $247M (FY98 dollars) must be made available as requested over the ten year planning period, including a one-time $10M to enhance Late Wash attainment. If appropriate resources are made available, facility attainment issues are resolved and regulatory support is sufficient, then completion of the HLW program in 2018 would achieve a $3.3 billion cost savings to DOE, versus the cost of completing the program in 2026. Facility status information is current as of October 31, 1996

  17. 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

  18. The Pinellas Plant RCRA facility investigation - A case study

    International Nuclear Information System (INIS)

    Kilbury, Richard; Keshian, Berg; Farley, Dwain; Meyer, David; Ingle, David; Biedermann, Charles

    1992-01-01

    Under the direction of the U.S. Department of Energy Albuquerque Field Office Environmental Restoration Program, a Resource Conservation and Recovery Act (RCRA) Facility Investigation (RFI) was completed at the Pinellas Plant to fulfill requirements of the Hazardous and Solid Waste Amendments of 1984 (HSWA) permit issued on February 9, 1990 by the U.S. Environmental Protection Agency (EPA). This RFI addressed potential contaminant releases and environmental conditions at 15 solid waste management units (SWMUs). The RFI characterization program began in April 1990 and was completed in May 1991. The scope of RFI data collection activities is presented in the Pinellas Plant RFI Workplan issued in May 1990 and approved by EPA on April 16, 1991. An RFI Report was submitted to EPA on September 1, 1991. This paper presents a summary of RFI results and conclusions. Primary environmental concerns at the Pinellas Plant are emphasized. (author)

  19. Dating the period when intensive anthropogenic activity began to influence the Sanjiang Plain, Northeast China

    Science.gov (United States)

    Cong, Jinxin; Gao, Chuanyu; Zhang, Yan; Zhang, Shaoqing; He, Jiabao; Wang, Guoping

    2016-01-01

    Dating the start of intensive anthropogenic influence on ecosystems is important for identifying the conditions necessary for ecosystem recovery. However, few studies have focused on determining when anthropogenic influences on wetland began through sedimentary archives. To fill this critical gap in our knowledge, combustion sources and emission intensities, reconstructed via black carbon (BC) and polycyclic aromatic hydrocarbons (PAHs) were analyzed in two wetlands in the Sanjiang Plain in Northeast China. 14C provided age control for the sedimentary records. By combining previous sedimentary and archaeological studies, we attempt to date the beginning of intensive anthropogenic influences on the Sanjiang Plain. Our results showed that BC deposition fluxes increased from 0.02 to 0.7 g C/m2.yr during the last 10,000 years. An upward trend was apparent during the last 500 years. Before 1200 cal yr BP, human activities were minor, such that the wetland ecosystem in the Sanjiang Plain before this period may represent the reference conditions that for the recovery of these wetlands. As the human population increased after 1200 cal yr BP, combustion sources changed and residential areas became a major source of BC and PAHs. In this way, the wetland ecosystem gradually became more heavily influenced by human activities. PMID:26907560

  20. Dating the period when intensive anthropogenic activity began to influence the Sanjiang Plain, Northeast China

    Science.gov (United States)

    Cong, Jinxin; Gao, Chuanyu; Zhang, Yan; Zhang, Shaoqing; He, Jiabao; Wang, Guoping

    2016-02-01

    Dating the start of intensive anthropogenic influence on ecosystems is important for identifying the conditions necessary for ecosystem recovery. However, few studies have focused on determining when anthropogenic influences on wetland began through sedimentary archives. To fill this critical gap in our knowledge, combustion sources and emission intensities, reconstructed via black carbon (BC) and polycyclic aromatic hydrocarbons (PAHs) were analyzed in two wetlands in the Sanjiang Plain in Northeast China. 14C provided age control for the sedimentary records. By combining previous sedimentary and archaeological studies, we attempt to date the beginning of intensive anthropogenic influences on the Sanjiang Plain. Our results showed that BC deposition fluxes increased from 0.02 to 0.7 g C/m2.yr during the last 10,000 years. An upward trend was apparent during the last 500 years. Before 1200 cal yr BP, human activities were minor, such that the wetland ecosystem in the Sanjiang Plain before this period may represent the reference conditions that for the recovery of these wetlands. As the human population increased after 1200 cal yr BP, combustion sources changed and residential areas became a major source of BC and PAHs. In this way, the wetland ecosystem gradually became more heavily influenced by human activities.

  1. 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

  2. Rheology of Savannah River site tank 42 and tank 51 HLW radioactive sludges

    International Nuclear Information System (INIS)

    Ha, B.C.; Bibler, N.E.

    1996-01-01

    Knowledge of the rheology of the radioactive sludge slurries at the Savannah River Site (SRS) is necessary in order to ensure that they can be retrieved from waste tanks and processed for final disposal. The high activity radioactive wastes stored as caustic slurries at SRS result from the neutralization of acid waste generated from production of nuclear defense materials. During storage, the wastes separate into a supernate layer and a sludge layer. In the Defense Waste Processing Facility (DWPF) at SRS, the radionuclides from the sludge and supernate will be immobilized into borosilicate glass for long term storage and eventual disposal. Before transferring the waste from a storage tank to the DWPF, a portion of the aluminum in the waste sludge will be dissolved and the sludge will be extensively washed to remove sodium. Tank 51 and Tank 42 radioactive sludges represent the first batch of HLW sludge to be processed in the DWPF. This paper presents results of rheology measurements of Tank 51 and Tank 42 at various solids concentrations. The rheologies of Tank 51 and Tank 42 radioactive slurries were measured remotely in the Shielded Cells Operations (SCO) at the Savannah River Technology Center (SRTC) using a modified Haake Rotovisco RV-12 with an M150 measuring drive unit and TI sensor system. Rheological properties of the Tank 51 and Tank 42 radioactive sludges were measured as a function of weight percent solids. The weight percent solids of Tank 42 sludge was 27, as received. Tank 51 sludge had already been washed. The weight percent solids were adjusted by dilution with water or by concentration through drying. At 12, 15, and 18 weight percent solids, the yield stresses of Tank 51 sludge were 5, 11, and 14 dynes/cm2, respectively. The apparent viscosities were 6, 10, and 12 centipoises at 300 sec-1 shear rate, respectively

  3. Two years of outstanding AFGD performance, pure air on the Lakes Bailly Scrubber Facility

    Energy Technology Data Exchange (ETDEWEB)

    Henderson, J.; Vymazal, D.C. [Pure Air, Allentown, PA (United States); Styf, D.A. [Northern Indiana Public Service Company (NIPSCO), Chesterton, IN (United States)] [and others

    1994-12-31

    The {open_quotes}Advanced Flue Gas Desulfurization (AFGD) Demonstration Project{close_quotes} is a $151.3 million cooperative effort between the U.S. Department of Energy and a project company of Pure Air, a general partnership of Air Products and Chemicals, Inc. and Mitsubishi Heavy Industries America, Inc. The goal of the AFGD project is to demonstrate that, by combining state-of-the-art technology, highly efficient plant operation and maintenance capabilities, and by-product gypsum sales, significant reductions of SO{sub 2} emissions can be achieved at approximately one-half the life cycle of a conventional Flue Gas Desulfurization (FGD) system. Further, this emission reduction is achieved without generating solid waste and while minimizing liquid wastewater effluent. Briefly, this project entails the design, construction and operation of a nominal 600 MWe AFGD facility to remove SO{sub 2} from coal-fired power plant flue gas at the Northern Indiana Public Service Company`s Bailly Generating Station, located approximately 40 miles southeast of Chicago, Illinois. The facility is used to demonstrate a variety of advanced technical and business-related features, during a three-year period of operation which began in the summer of 1992. The aim of this demonstration is to accelerate near-term commercialization. Key features of the AFGD project are discussed. Construction of the scrubber is complete; operations began in June 1992, ahead of schedule and within budget. The Clean Coal demonstration project calls for three years of operations. After the three-year demonstration period, Pure Air on the Lake will continue to Own-and-Operate the scrubber for the next 17 years. This paper review the advanced wet flue gas desulfurization (FGD) design features, and the environmental and business features of the project. Also included are data on the first two years of successful operation.

  4. 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

  5. Waste acceptance product specifications for vitrified high-level waste forms

    International Nuclear Information System (INIS)

    Applewhite-Ramsey, A.; Sproull, J.F.

    1994-01-01

    The Department of Energy (DOE) Office of Environmental Restoration and Waste Management (EM) has developed Waste Acceptance Product Specifications (EM-WAPS). The EM-WAPS will be the basis for defining product acceptance criteria compatible with the requirements of the Civilian Radioactive Waste Management System (CRWMS). The relationship between the EM-WAPS and the CRWMS Systems Requirements document (WA-SRD) will be discussed. The impact of the EM-WAPS on the Savannah River Sit (SRS) Defense Waste Processing Facility's (DWPF) Waste Acceptance Program, Waste Qualification Run planning, and startup schedule will also be reported. 14 refs., 2 tabs

  6. 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

  7. 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

  8. High Sodium Simulant Testing To Support SB8 Sludge Preparation

    International Nuclear Information System (INIS)

    Newell, J. D.

    2012-01-01

    Scoping studies were completed for high sodium simulant SRAT/SME cycles to determine any impact to CPC processing. Two SRAT/SME cycles were performed with simulant having sodium supernate concentration of 1.9M at 130% and 100% of the Koopman Minimum Acid requirement. Both of these failed to meet DWPF processing objectives related to nitrite destruction and hydrogen generation. Another set of SRAT/SME cycles were performed with simulant having a sodium supernate concentration of 1.6M at 130%, 125%, 110%, and 100% of the Koopman Minimum Acid requirement. Only the run at 110% met DWPF processing objectives. Neither simulant had a stoichiometric factor window of 30% between nitrite destruction and excessive hydrogen generation. Based on the 2M-110 results it was anticipated that the 2.5M stoichiometric window for processing would likely be smaller than from 110-130%, since it appeared that it would be necessary to increase the KMA factor by at least 10% above the minimum calculated requirement to achieve nitrite destruction due to the high oxalate content. The 2.5M-130 run exceeded the DWPF hydrogen limits in both the SRAT and SME cycle. Therefore, testing of this wash endpoint was halted. This wash endpoint with this minimum acid requirement and mercury-noble metal concentration profile appears to be something DWPF should not process due to an overly narrow window of stoichiometry. The 2M case was potentially processable in DWPF, but modifications would likely be needed in DWPF such as occasionally accepting SRAT batches with undestroyed nitrite for further acid addition and reprocessing, running near the bottom of the as yet ill-defined window of allowable stoichiometric factors, potentially extending the SRAT cycle to burn off unreacted formic acid before transferring to the SME cycle, and eliminating formic acid additions in the frit slurry

  9. Outline of construction and facility features of Onagawa nuclear power station Unit No. 2

    International Nuclear Information System (INIS)

    Umimura, Yoshiharu; Tsunoda, Ryohei; Watanabe, Kazunori

    1996-01-01

    Tohoku Electric Power Company promotes development of various power sources to provide a stable supply of electricity in the future, and nuclear power takes a leading part. In August 1989, construction of Onagawa nuclear power plant Unit No. 2 (825MW) was started, following Unit No. 1 (524MW) which went on line in 1984 as Tohoku Electric's first nuclear power plant unit. Unit No. 2 began commercial operation in July 1995 through satisfactory construction work such as RPV hydraulic test in March 1994, fuel loading in October 1994, and various startup tests in each power stage. The design and construction of Unit No. 2 reflect construction and operation experience gained from Unit No. 1, and the latest technology, including that of the LWR Improvement and Standardization Program, was adopted to enhance facility reliability, improve operation and maintenance performance, and reduce worker dosage. Features of the facility, construction techniques, and a description of preoperation of Onagawa nuclear power plant Unit No. 2 are described in this paper. (author)

  10. How nonimaging optics began

    Science.gov (United States)

    Winston, Roland

    2016-09-01

    Classical optics was traditionally the mapping of point sources by lenses, mirrors and occasionally holograms , i.e. making an image. The subject has had many famous scientists associated with it; Fermat, Huygens, Descartes, Hamilton just to name a few. By the mid 20th Century it was a well-developed field as exemplified by such luminaries as Walter T. Welford, Emil Wolf and many others. The theory of aberrations which addresses the imperfections of the mapping codified the state of the art. Then arose the need to collect energy, not just images. To the author's knowledge it can be traced back to WWII Germany with collection of infra-red radiation (the work by D. E. Williamson, was not published until 1952). The radiation collector, a simple right-circular cone, was a harbinger of things to come.

  11. 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.

  12. 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

  13. Remote handling in the Plutonium Immobilization Project: Puck handling

    International Nuclear Information System (INIS)

    Brault, J.R.

    2000-01-01

    Since the break up of the Soviet Union at the end of the Cold War, the US and Russia have been negotiating ways to reduce their nuclear stockpiles. Economics is one of the reasons behind this, but another important reason is safeguarding these materials from unstable organizations and countries. With the downsizing of the nuclear stockpiles, large quantities of plutonium are being declared excess and must be safely disposed of. The Savannah River Site (SRS) has been selected as the site where the immobilization facility will be located. Conceptual design and process development commenced in 1998. SRS will immobilize excess plutonium in a ceramic waste form and encapsulate it in vitrified high level waste in the Defense Waste Processing Facility (DWPF) canister. These canisters will then be interred in the national repository at Yucca Mountain, New Mexico. The facility is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. This paper will discuss the first two operations

  14. Remote handling in the Plutonium Immobilization Project: Plutonium conversion and first stage immobilization

    International Nuclear Information System (INIS)

    Brault, J.R.

    2000-01-01

    Since the break up of the Soviet Union at the end of the Cold War, the United States and Russia have been negotiating ways to reduce their nuclear stockpiles. Economics is one of the reasons behind this, but another important reason is safeguarding these materials from unstable organizations and countries. With the downsizing of the nuclear stockpiles, large quantities of plutonium are being declared excess and must be safely disposed of. The Savannah River Site (SRS) has been selected as the site where the immobilization facility will be located. Conceptual design and process development commenced in 1998. SRS will immobilize excess plutonium in a ceramic waste form and encapsulate it in vitrified high level waste in the Defense Waste Processing Facility (DWPF) canister. These canisters will then be interred in the national repository at Yucca Mountain, New Mexico. The facility is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. This paper will discuss the first two operations

  15. Lightning protection for the process canyons at the Savannah River site

    International Nuclear Information System (INIS)

    McAfee, D.E.

    1995-01-01

    Westinghouse Savannah River Company (WSRC) has performed Lightning Studies for the existing Process Canyons at the Savannah River Site (SRS). These studies were initiated to verify the lightning protection systems for the facilities and to compare the installations to the National Fire Protection (NFPA) Standard 780, Lighting Protection Code, 1992. The original study of the F-Canyon was initiated to develop answers to concerns raised by the Defense Nuclear Facility Safety Board (DNFSB). Once this study was completed it was determined that a similar study for H-Canyon would be prudent; followed by an evaluation of the Defense Waste Processing Facility (DWPF) Vitrification Building (S-Canyon). This paper will provide an overview of the nature of lightning and the principals of lightning protection. This will provide the reader with a basic understanding of the phenomena of lighting and its potential for damaging structures, components, and injuring personnel in or near the structure

  16. Emission Facilities - Erosion & Sediment Control Facilities

    Data.gov (United States)

    NSGIC Education | GIS Inventory — An Erosion and Sediment Control Facility is a DEP primary facility type related to the Water Pollution Control program. The following sub-facility types related to...

  17. The effect of the National Health Insurance Scheme (NHIS) on health service delivery in mission facilities in Ghana: a retrospective study.

    Science.gov (United States)

    Aryeetey, Genevieve Cecilia; Nonvignon, Justice; Amissah, Caroline; Buckle, Gilbert; Aikins, Moses

    2016-06-07

    In 2004, Ghana began implementation of a National Health Insurance Scheme (NHIS) to minimize out-of-pocket expenditure at the point of use of service. The implementation of the scheme was accompanied by increased access and use of health care services. Evidence suggests most health facilities are faced with management challenges in the delivery of services. The study aimed to assess the effect of the introduction of the NHIS on health service delivery in mission health facilities in Ghana. We conceptualised the effect of NHIS on facilities using service delivery indicators such as outpatient and inpatient turn out, estimation of general service readiness, revenue and expenditure, claims processing and availability of essential medicines. We collected data from 38 mission facilities, grouped into the three ecological zones; southern, middle and northern. Structured questionnaires and exit interviews were used to collect data for the periods 2003 and 2010. The data was analysed in SPSS and MS Excel. The facilities displayed high readiness to deliver services. There were significant increases in outpatient and inpatient attendance, revenue, expenditure and improved access to medicines. Generally, facilities reported increased readiness to deliver services. However, challenging issues around high rates of non-reimbursement of NHIS claims due to errors in claims processing, lack of feedback regarding errors, and lack of clarity on claims reporting procedures were reported. The implementation of the NHIS saw improvement and expansion of services resulting in benefits to the facilities as well as constraints. The constraints could be minimized if claims processing is improved at the facility level and delays in reimbursements also reduced.

  18. Study of the application of near-real-time materials accountancy to safeguards for reprocessing facilities

    International Nuclear Information System (INIS)

    Ikawa, Koji; Ihara, Hitoshi; Nishimura, Hideo; Hirata, Mitsuho; Sakuragi, Hirotaka; Ido, Masaru.

    1983-09-01

    This report describes the results of TASTEX task F, the basic purpose of which was to investigate the feasibility of applying the basic concepts of near-real-time materials accountancy to small or medium-sized spent fuel reprocessing facilities, using the PNC-Tokai facility as a model. The background of Task-F and the proposed IAEA requirements on reprocessing plant safeguards are briefly shown. A model of near-real-time materials accountancy based on weekly material balances covering the entire process MBA is outlined, and the effectiveness of this model is evaluated based on simulation and analysis procedures developed for the study. The results show that the proposed materials accountancy model should provide sufficient information to satisfy IAEA guidelines for detection goals. Field testing of the model began in 1980, and the preliminary evaluation of this field test data shows that weekly in-process physical inventories are possible without affecting process operations. This report also describes studies related to IAEA verification procedures, and identifies necessary further work. (author)

  19. Dismantlement of nuclear facilities decommissioned from the Russian navy: Enhancing regulatory supervision of nuclear and radiation safety

    International Nuclear Information System (INIS)

    Sneve, M.K.

    2013-01-01

    The availability of up to date regulatory norms and standards for nuclear and radiation safety, relevant to the management of nuclear legacy situations, combined with effective and efficient regulatory procedures for licensing and monitoring compliance, are considered to be extremely important. Accordingly the NRPA has set up regulatory cooperation programs with corresponding authorities in the Russian Federation. Cooperation began with the civilian regulatory authorities and was more recently extended to include the military authority and this joint cooperation supposed to develop the regulatory documents to improve supervision over nuclear and radiation safety while managing the nuclear military legacy facilities in Northwest Russia and other regions of the country. (Author)

  20. Dismantlement of nuclear facilities decommissioned from the Russian navy: Enhancing regulatory supervision of nuclear and radiation safety

    Energy Technology Data Exchange (ETDEWEB)

    Sneve, M.K.

    2013-03-01

    The availability of up to date regulatory norms and standards for nuclear and radiation safety, relevant to the management of nuclear legacy situations, combined with effective and efficient regulatory procedures for licensing and monitoring compliance, are considered to be extremely important. Accordingly the NRPA has set up regulatory cooperation programs with corresponding authorities in the Russian Federation. Cooperation began with the civilian regulatory authorities and was more recently extended to include the military authority and this joint cooperation supposed to develop the regulatory documents to improve supervision over nuclear and radiation safety while managing the nuclear military legacy facilities in Northwest Russia and other regions of the country. (Author)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  2. PROJECT EXPERIENCE REPORT DEMOLITION OF HANFORDS 233-S PLUTONIUM CONCENTRATION FACILITY

    International Nuclear Information System (INIS)

    BERLIN, G.T.; ORGILL, T.K.

    2004-01-01

    This report provides a summary of the preparation, operations, innovative work practices, and lessons learned associated with demolition of the 2334 Plutonium Concentration Facility. This project represented the first open-air demolition of a highly-contaminated plutonium facility at the Hanford Site. This project may also represent the first plutonium facility in the US. Department of Energy (DOE) complex to have been demolished without first decontaminating surfaces to near ''free release'' standards. Demolition of plutonium contaminated structures, if not properly managed, can subject cleanup personnel and the environment to significant risk. However, with proper sequencing and innovative use of commercially available equipment, materials, and services, this project demonstrated that a plutonium processing facility can be demolished while avoiding the need to perform extensive decontamination or to construct large enclosures. This project utilized an excavator with concrete shears, diamond circular saws, water misting and fogging equipment, commercially available fixatives and dust suppressants, conventional mobile crane and rigging services, and near real-time modeling of meteorological and radiological conditions. Following a significant amount of preparation, actual demolition of the 233-S Facility began in October 2003 and was completed in late April 2004. The knowledge and experience gained on this project are important to the Hanford Site as additional plutonium processing facilities are scheduled for demolition in the near future. Other sites throughout the DOE Complex may also be faced with similar challenges. Numerous innovations and effective work practices were implemented on this project. Accordingly, a series of ''Lessons Learned and Innovative Practices Fact Sheets'' were developed and are included as an appendix to this report. This collection of fact sheets is not intended to capture every innovative work practice and lesson learned, but rather

  3. PROJECT EXPERIENCE REPORT DEMOLITION OF HANFORDS 233-S PLUTONIUM CONCENTRATION FACILITY

    International Nuclear Information System (INIS)

    BERLIN, G.T.

    2004-01-01

    This report provides a summary of the preparation, operations, innovative work practices, and lessons learned associated with demolition of the 2334 Plutonium Concentration Facility. This project represented the first open-air demolition of a highly-contaminated plutonium facility at the Hanford Site. This project may also represent the first plutonium facility in the US. Department of Energy (DOE) complex to have been demolished without first decontaminating surfaces to near ''free release'' standards. Demolition of plutonium contaminated structures, if not properly managed, can subject cleanup personnel and the environment to significant risk. However, with proper sequencing and innovative use of commercially available equipment, materials, and services, this project demonstrated that a plutonium processing facility can be demolished while avoiding the need to perform extensive decontamination or to construct large enclosures. This project utilized an excavator with concrete shears, diamond circular saws, water misting and fogging equipment, commercially available fixatives and dust suppressants, conventional mobile crane and rigging services, and near real-time modeling of meteorological and radiological conditions. Following a significant amount of preparation, actual demolition of the 2333 Facility began in October 2003 and was completed in late April 2004. The knowledge and experience gained on this project are important to the Hanford Site as additional plutonium processing facilities are scheduled for demolition in the near future. Other sites throughout the DOE Complex may also be faced with similar challenges. Numerous innovations and effective work practices were implemented on this project. Accordingly, a series of ''Lessons Learned and Innovative Practices Fact Sheets'' were developed and are included as an appendix to this report. This collection of fact sheets is not intended to capture every innovative work practice and lesson learned, but rather to

  4. 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

  5. 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

  6. 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

  7. Process innovations to minimize waste volumes at Savannah River

    International Nuclear Information System (INIS)

    Doherty, J.P.

    1986-01-01

    In 1983 approximately 1.6 x 10 3 m 3 (427,000 gallons) of radioactive salt solution were decontaminated in a full-scale demonstration. The cesium decontamination factor (DF) was in excess of 4 x 10 4 vs. a goal of 1 x 10 4 . Data from this test were combined with pilot data and used to design the permanent facilities currently under construction. Startup of the Salt Decontamination Process is scheduled for 1987 and will decontaminate 2 x 10 4 m 3 (5.2 million gallons) of radioactive salt solution and generate 2 x 10 3 m 3 (520,000 gallons) of concentrated and washed precipitate per year. The Defense Waste Processing Facility (DWPF) will begin processing this concentrate in the Precipitate Hydrolysis Process starting in 1989. Laboratory data using simulated salt solution and nonradioactive cesium are being used to design this process. A 1/5-scale pilot plant is under construction and will be used to gain large-scale operating experience using nonradioactive simulants. This pilot plant is scheduled to startup in early 1987. The incentives to reduce the volume of waste that must be treated are self-evident. At Savannah River process development innovations to minimize the DWPF feed volumes have directly improved the economics of the process. The integrity of the final borosilicate glass water form has not been compromised by these developments. Many of the unit operations are familiar to chemical engineers and were put to use in a unique environment. As a result, tax dollars have been saved, and the objective of safely disposing of the nation's high-level defense waste has moved forward

  8. Hanford Waste Vitrification Plant - the project and process systems

    International Nuclear Information System (INIS)

    Swenson, L.D.; Miller, W.C.; Smith, R.A.

    1990-01-01

    The Hanford Waste Vitrification Plant (HWVP) project is scheduled to start construction on the Hanford reservation in southeastern Washington State in 1991. The project will immobilize the liquid high-level defense waste stored there. The HWVP represents the third phase of the U.S. Department of Energy (DOE) activities that are focused on the permanent disposal of high-level radioactive waste, building on the experience of Defense Waste Processing Facility (DWPF) at the Savannah River site, South Carolina, and of the West Valley Demonstration Plant (WVDP), New York. This sequential approach to disposal of the country's commercial and defense high-level radioactive waste allows HWVP to make extensive use of lessons learned from the experience of its predecessors, using mature designs from the earlier facilities to achieve economies in design and construction costs while enhancing operational effectiveness

  9. Final work plan : indoor air and ambient air sampling near the former CCC/USDA grain storage facility in Everest, Kansas.

    Energy Technology Data Exchange (ETDEWEB)

    LaFreniere, L. M. (Environmental Science Division)

    2010-05-24

    The Commodity Credit Corporation of the U.S. Department of Agriculture (CCC/USDA) operated a grain storage facility at the western edge of Everest, Kansas, from the early 1950s to the early 1970s. Sampling by the Kansas Department of Health and Environment (KDHE) in 1997 resulted in the detection of carbon tetrachloride in one domestic well (the Nigh well) northwest of the former facility. On behalf of the CCC/USDA, Argonne National Laboratory subsequently conducted a series of investigations to characterize the contamination (Argonne 2003, 2006a,b,c). Automatic, continuous monitoring of groundwater levels began in 2002 and is ongoing at six locations. The results have consistently indicated groundwater flow toward the north-northwest from the former CCC/USDA property to the Nigh property, then west-southwest from the Nigh property to the intermittent creek. Sitewide periodic groundwater and surface water sampling with analysis for volatile organic compounds (VOCs) began in 2008. Argonne's combined data indicate no significant downgradient extension of contamination since 2000. At present, the sampling is annual, as approved by the KDHE (2009) in response to a plan developed for the CCC/USDA (Argonne 2009). This document presents a plan for collecting indoor air samples in homes located along and adjacent to the defined extent of the carbon tetrachloride contamination. The plan was requested by the KDHE. Ambient air samples to represent the conditions along this pathway will also be taken. The purpose of the proposed work is to satisfy KDHE requirements and to collect additional data for assessing the risk to human health due to the potential upward migration of carbon tetrachloride and its primary degradation product (chloroform) into homes located in close proximity to the former grain storage facility, as well as along and within 100 ft laterally from the currently defined plume emanating from the former Everest facility. Investigation of the indoor air

  10. Closure Report for Corrective Action Unit 116: Area 25 Test Cell C Facility, Nevada National Security Site, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    NSTec Environmental Restoration

    2011-09-29

    This Closure Report (CR) presents information supporting closure of Corrective Action Unit (CAU) 116, Area 25 Test Cell C Facility. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; the U.S. Department of Energy (DOE), Environmental Management; the U.S. Department of Defense; and DOE, Legacy Management (FFACO, 1996 [as amended March 2010]). CAU 116 consists of the following two Corrective Action Sites (CASs), located in Area 25 of the Nevada National Security Site: (1) CAS 25-23-20, Nuclear Furnace Piping and (2) CAS 25-41-05, Test Cell C Facility. CAS 25-41-05 consisted of Building 3210 and the attached concrete shield wall. CAS 25-23-20 consisted of the nuclear furnace piping and tanks. Closure activities began in January 2007 and were completed in August 2011. Activities were conducted according to Revision 1 of the Streamlined Approach for Environmental Restoration Plan for CAU 116 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office [NNSA/NSO], 2008). This CR provides documentation supporting the completed corrective actions and provides data confirming that closure objectives for CAU 116 were met. Site characterization data and process knowledge indicated that surface areas were radiologically contaminated above release limits and that regulated and/or hazardous wastes were present in the facility.

  11. Evaluation Of ARG-1 Samples Prepared By Cesium Carbonate Dissolution During The Isolok SME Acceptability Testing

    International Nuclear Information System (INIS)

    Edwards, T.; Hera, K.; Coleman, C.

    2011-01-01

    Evaluation of Defense Waste Processing Facility (DWPF) Chemical Process Cell (CPC) cycle time identified several opportunities to improve the CPC processing time. The Mechanical Systems and Custom Equipment Development (MS and CED) Section of the Savannah River National Laboratory (SRNL) recently completed the evaluation of one of these opportunities - the possibility of using an Isolok sampling valve as an alternative to the Hydragard valve for taking DWPF process samples at the Slurry Mix Evaporator (SME). The use of an Isolok for SME sampling has the potential to improve operability, reduce maintenance time, and decrease CPC cycle time. The SME acceptability testing for the Isolok was requested in Task Technical Request (TTR) HLW-DWPF-TTR-2010-0036 and was conducted as outlined in Task Technical and Quality Assurance Plan (TTQAP) SRNLRP-2011-00145. RW-0333P QA requirements applied to the task, and the results from the investigation were documented in SRNL-STI-2011-00693. Measurement of the chemical composition of study samples was a critical component of the SME acceptability testing of the Isolok. A sampling and analytical plan supported the investigation with the analytical plan directing that the study samples be prepared by a cesium carbonate (Cs 2 CO 3 ) fusion dissolution method and analyzed by Inductively Coupled Plasma - Optical Emission Spectroscopy (ICP-OES). The use of the cesium carbonate preparation method for the Isolok testing provided an opportunity for an additional assessment of this dissolution method, which is being investigated as a potential replacement for the two methods (i.e., sodium peroxide fusion and mixed acid dissolution) that have been used at the DWPF for the analysis of SME samples. Earlier testing of the Cs 2 CO 3 method yielded promising results which led to a TTR from Savannah River Remediation, LLC (SRR) to SRNL for additional support and an associated TTQAP to direct the SRNL efforts. A technical report resulting from this

  12. 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

  13. Development Of A Macro-Batch Qualification Strategy For The Hanford Tank Waste Treatment And Immobilization Plant

    International Nuclear Information System (INIS)

    Herman, Connie C.

    2013-01-01

    The Savannah River National Laboratory (SRNL) has evaluated the existing waste feed qualification strategy for the Hanford Tank Waste Treatment and Immobilization Plant (WTP) based on experience from the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) waste qualification program. The current waste qualification programs for each of the sites are discussed in the report to provide a baseline for comparison. Recommendations on strategies are then provided that could be implemented at Hanford based on the successful Macrobatch qualification strategy utilized at SRS to reduce the risk of processing upsets or the production of a staged waste campaign that does not meet the processing requirements of the WTP. Considerations included the baseline WTP process, as well as options involving Direct High Level Waste (HLW) and Low Activity Waste (LAW) processing, and the potential use of a Tank Waste Characterization and Staging Facility (TWCSF). The main objectives of the Hanford waste feed qualification program are to demonstrate compliance with the Waste Acceptance Criteria (WAC), determine waste processability, and demonstrate unit operations at a laboratory scale. Risks to acceptability and successful implementation of this program, as compared to the DWPF Macro-Batch qualification strategy, include: Limitations of mixing/blending capability of the Hanford Tank Farm; The complexity of unit operations (i.e., multiple chemical and mechanical separations processes) involved in the WTP pretreatment qualification process; The need to account for effects of blending of LAW and HLW streams, as well as a recycle stream, within the PT unit operations; and The reliance on only a single set of unit operations demonstrations with the radioactive qualification sample. This later limitation is further complicated because of the 180-day completion requirement for all of the necessary waste feed qualification steps. The primary recommendations/changes include the

  14. Progress and future directions for remediation of Hanford facilities and contaminated sites

    International Nuclear Information System (INIS)

    McClain, L.K.; Nemec, J.F.

    1996-01-01

    A great deal of physical progress is being made in the Hanford Environmental Restoration (ER) Project, which is responsible for the portion of work at Hanford that deals with contaminated soil and groundwater, and with inactive nuclear facilities. This work accounts for 10 to 15 percent of the Hanford site budget. (Other US Department of Energy [DOE] programs and contractors are responsible for the high-level liquid waste in underground storage tanks and the spent nuclear fuel). The project open-quotes closed the circleclose quotes on environmental restoration at Hanford this summer when the Environmental Restoration Disposal Facility (ERDF) went into operation and began receiving wastes being excavated from contaminated areas in Hanford's open-quotes 100 Areaclose quotes along the Columbia River. With this milestone event, environmental restoration at Hanford now has a clear path forward: (1) Waste areas along the Columbia River have been identified, volume estimates are being refined, and excavation has started. (2) The million-cubic-yard capacity ERDF is receiving waste from excavation in the 100 Area. (3) Deactivation of the N Reactor will be completed within a year. (4) Numerous other facilities in the 100 Area are being decommissioned, eliminating hazards and reducing the costs of surveillance and maintenance (S ampersand M). (5) A demonstration of long-term protective storage for one of the reactor blocks is in progress. (6) A comprehensive groundwater treatment strategy is in place. This paper describes the Hanford ER project, the progress being made, and the management techniques that are making the project successful

  15. SLUDGE MASS REDUCTION: PRIMARY COMPOSITIONAL FACTORS THAT INFLUENCE MELT RATE FOR FUTURE SLUDGE BATCH PROJECTIONS

    International Nuclear Information System (INIS)

    Newell, J; Miller, D; Stone, M; Pickenheim, B

    2008-01-01

    The Savannah River National Laboratory (SRNL) was tasked to provide an assessment of the downstream impacts to the Defense Waste Processing Facility (DWPF) of decisions regarding the implementation of Al-dissolution to support sludge mass reduction and processing. Based on future sludge batch compositional projections from the Liquid Waste Organization's (LWO) sludge batch plan, assessments have been made with respect to the ability to maintain comparable projected operating windows for sludges with and without Al-dissolution. As part of that previous assessment, candidate frits were identified to provide insight into melt rate for average sludge batches representing with and without Al-dissolution flowsheets. Initial melt rate studies using the melt rate furnace (MRF) were performed using five frits each for Cluster 2 and Cluster 4 compositions representing average without and with Al-dissolution. It was determined, however, that the REDOX endpoint (Fe 2+ /ΣFe for the glass) for Clusters 2 and 4 resulted in an overly oxidized feed which negatively affected the initial melt rate tests. After the sludge was adjusted to a more reduced state, additional testing was performed with frits that contained both high and low concentrations of sodium and boron oxides. These frits were selected strictly based on the ability to ascertain compositional trends in melt rate and did not necessarily apply to any acceptability criteria for DWPF processing. The melt rate data are in general agreement with historical trends observed at SRNL and during processing of SB3 (Sludge Batch 3)and SB4 in DWPF. When MAR acceptability criteria were applied, Frit 510 was seen to have the highest melt rate at 0.67 in/hr for Cluster 2 (without Al-dissolution), which is compositionally similar to SB4. For Cluster 4 (with Al-dissolution), which is compositionally similar to SB3, Frit 418 had the highest melt rate at 0.63 in/hr. Based on this data, there appears to be a slight advantage of the Frit

  16. Safety of fuel cycle facilities. Topical issues paper no. 3

    International Nuclear Information System (INIS)

    Ranguelova, V.; Niehaus, F.; Delattre, D.

    2001-01-01

    that the relative hazards vary from facility to facility depending upon the processes employed. However, all installations should be designed and operated so as to keep all sources of radiation exposure under strict technical and administrative control. The design management should ensure that the structures, systems and components important to safety have the appropriate characteristics, specifications and material composition to perform the required safety functions. Emphasis should be placed on the use of proven engineered safety features in the implementation of the defence in depth concept in the facility design and operation. The development of nuclear safety standards is one of the tasks given by the Statute of the IAEA. Over a number of years, the IAEA has developed a comprehensive set of publications which address, in a structured manner, the safety of nuclear installations. The Safety Fundamentals publication entitled 'The Safety of Nuclear Installations' presents an international consensus on the basic concepts underlying the principles for the regulation, design and operation of nuclear installations, including fuel cycle facilities. Detailed requirements and guides for activities relating to siting, design, construction, operation, decommissioning and regulation of nuclear installations are addressed by lower hierarchy safety standards, which cover nuclear power plants (NPPs) and research reactors. Although some of the IAEA NPP standards can be adapted and applied to fuel cycle facilities and some standards on criticality safety are published by the International Organization for Standardization (ISO), in general there is a lack of international safety standards to cover the safety of such facilities. With a view to establishing a plan for the development of safety standards for fuel cycle facilities, the IAEA began to compile information on the status of national regulations and safety issues concerning such facilities. It also held a Technical Committee

  17. WIPP Facility Work Plan for Solid Waste Management Units

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2000-02-25

    This Facility Work Plan (FWP) has been prepared as required by Module VII,Section VII.M.1 of the Waste Isolation Pilot Plant (WIPP) Hazardous Waste Permit, NM4890139088-TSDF (the Permit); (NMED, 1999a). This work plan describes the programmatic facility-wide approach to future investigations at Solid Waste Management Units (SWMUs) and Areas of Concern (AOCs) specified in the Permit. This FWP addresses the current Permit requirements. It uses the results of previous investigations performed at WIPP and expands the investigations as required by the Permit. As an alternative to the Resource Conservation and Recovery Act (RCRA) Facility Investigation (RFI) specified in Module VII of the Permit, current New Mexico Environment Department (NMED) guidance identifies an Accelerated Corrective Action Approach (ACAA) that may be used for any SWMU or AOC (NMED, 1998). This accelerated approach is used to replace the standard RFI Work Plan and Report sequence with a more flexible decision-making approach. The ACAA process allows a Facility to exit the schedule of compliance contained in the Facility’s Hazardous and Solid Waste Amendments (HSWA) permit module and proceed on an accelerated time frame. Thus, the ACAA process can be entered either before or after an RFI Work Plan. According to NMED’s guidance, a facility can prepare an RFI Work Plan or Sampling and Analysis Plan (SAP) for any SWMU or AOC (NMED, 1998). Based on this guidance, a SAP constitutes an acceptable alternative to the RFI Work Plan specified in the Permit. The scope of work for the RFI Work Plan or SAP is being developed by the Permittees. The final content of the RFI Work Plan or SAP will be coordinated with the NMED for submittal on May 24, 2000. Specific project-related planning information will be included in the RFI Work Plan or SAP. The SWMU program at WIPP began in 1994 under U.S. Environmental Protection Agency (EPA) regulatory authority. NMED subsequently received regulatory authority from EPA

  18. Environmental fate of hexabromocyclododecane from a new Canadian electronic recycling facility.

    Science.gov (United States)

    Tomko, Geoffrey; McDonald, Karen M

    2013-01-15

    An electronics recycling facility began operation at the municipal landfill site for the City of Edmonton, Canada in March 2008 with the goal of processing 30,000 tonnes of electronic wastes per year. Of the many by-products from the process, brominated fire retardants such as hexabromocyclododecane (HBCD) can evolve off of e-wastes and be released into the environmental media. HBCD has been identified by many countries and international bodies as a chemical of concern because of its ability to bioaccumulate in the ecosystem. An evaluation of the potential emission of HBCD indicates that up to 500 kg per year may be released from a landfill and recycling facility such as that operating in Edmonton. A multimedia fugacity model was used to evaluate the dispersion and fate of atmospherically emitted HBCD traveling into surrounding agricultural land and forested parkland. The model indicates that the three isomers of HBCD partitioned into environmental media similarly. Much of the HBCD is lost through atmospheric advection, but it is also found in soil and sediment. Modeled air concentrations are similar to those measured at locations with a history of e-waste recycling. Since HBCD has been shown to bioaccumulate, the HBCD released from this source has the long-term potential to affect agricultural food crops and the park ecosystem. Copyright © 2012 Elsevier Ltd. All rights reserved.

  19. Ressenya a Henry Ettinghausen, How the Press Began. The Pre-Periodical Printed News in Early Modern Europe, A Coruña, SIELAE – Facultad de Filología, Universidade da Coruña, 2015,

    Directory of Open Access Journals (Sweden)

    Ricard Expósito i Amagat

    2017-06-01

    Full Text Available Ressenya a Henry Ettinghausen, How the Press Began. The Pre-Periodical Printed News in Early Modern Europe, A Coruña, SIELAE – Facultad de Filología, Universidade da Coruña, 2015, 302 pp., 80 il·ls., ISBN: 978-84-608-3423-6 Review to  Henry Ettinghausen, How the Press Began. The Pre-Periodical Printed News in Early Modern Europe, A Coruña, SIELAE – Facultad de Filología, Universidade da Coruña, 2015, 302 pp., 80 il·ls., ISBN: 978-84-608-3423-6

  20. 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