Sample records for high activity waste

  1. Technical baseline description of high-level waste andlow-activity waste feed mobilization and delivery

    SciTech Connect (OSTI)

    Papp, I.G. [Numatec Hanford Corporation, Richland, WA 99352 (United States)

    1997-06-01T23:59:59.000Z

    This document is a compilation of information related to the high-level waste (HLW) and low-activity waste (LAW) feed staging, mobilization, and transfer/delivery issues. Information relevant to current Tank Waste Remediation System (TWRS) inventories and activities designed to feed the Phase I Privatization effort at the Hanford Site is included. Discussions on the higher level Phase II activities are offered for a perspective on the interfaces.

  2. Maximization of waste loading for a vitrified Hanford high-activity simulated waste

    SciTech Connect (OSTI)

    Fini, P.T. [State Univ. of New York, Alfred, NY (United States). Coll. of Ceramics; Hrma, P. [Pacific Northwest Lab., Richland, WA (United States)

    1994-04-01T23:59:59.000Z

    Simulated high-level nuclear waste glasses incorporating up to 70 wt % Neutralized Current Acid Waste (NCAW) were prepared. For the waste loading (W) range of 40 to 55 wt %, alkaliborosilicate glasses were formulated with a melting temperature of 1,150 C; for W > 55 wt %, only silica was added to the waste and the melting temperature was 1,150 C. Properties measured included durability and crystallinity of slowly cooled glasses and glasses heat treated for 24 hours at 1,050 C. Acceptable durability (by the Environmental Assessment glass standard) was retained up to W = 70 wt %, which is the maximum NCAW waste loading if no limit on crystallinity is imposed. If < 1 vol% of spinel is acceptable in the melt at 1,050 C, a waste loading of approximately 50 wt % is possible. If no crystallinity is permissible at 1,050 C, W = 34 wt % is the estimated maximum.

  3. WASTE SOLIDIFICATION BUILDING BENCH SCALE HIGH ACTIVITY WASTE SIMULANT VARIABILITY STUDY FY2008

    SciTech Connect (OSTI)

    Hansen, E; Timothy Jones, T; Tommy Edwards, T; Alex Cozzi, A

    2009-03-20T23:59:59.000Z

    The primary objective of this task was to perform a variability study of the high activity waste (HAW) acidic feed to determine the impact of feed variability on the quality of the final grout and on the mixability of the salt solution into the dry powders. The HAW acidic feeds were processed through the neutralization/pH process, targeting a final pH of 12. These fluids were then blended with the dry materials to make the final waste forms. A secondary objective was to determine if elemental substitution for cost prohibitive or toxic elements in the simulant affects the mixing response, thus providing a more economical simulant for use in full scale tests. Though not an objective, the HAW simulant used in the full scale tests was also tested and compared to the results from this task. A statistically designed test matrix was developed based on the maximum molarity inputs used to make the acidic solutions. The maximum molarity inputs were: 7.39 HNO{sub 3}, 0.11618 gallium, 0.5423 silver, and 1.1032 'other' metals based on their NO{sub 3}{sup -} contribution. Substitution of the elements aluminum for gallium and copper for silver was also considered in this test matrix, resulting in a total of 40 tests. During the NaOH addition, the neutralization/pH adjustment process was controlled to a maximum temperature of 60 C. The neutralized/pH adjusted simulants were blended with Portland cement and zircon flour at a water to cement mass ratio of 0.30. The mass ratio of zircon flour to Portland cement was 1/12. The grout was made using a Hobart N-50 mixer running at low speed for two minutes to incorporate and properly wet the dry solids with liquid and at medium speed for five minutes for mixing. The resulting fresh grout was measured for three consecutive yield stress measurements. The cured grout was measured for set, bleed, and density. Given the conditions of preparing the grout in this task, all of the grouts were visually well mixed prior to preparing the grouts for measurements. All of the cured grouts were measured for bleed and set. All of the cured grouts satisfied the bleed and set requirements, where no bleed water was observed on any of the grout samples after one day and all had set within 3 days of curing. This data indicates, for a well mixed product, bleed and set requirement are satisfied for the range of acidic feeds tested in this task. The yield stress measurements provide both an indication on the mixability of the salt solution with dry materials and an indication of how quickly the grout is starting to form structure. The inability to properly mix these two streams into a well mixed grout product will lead to a non-homogeneous mixture that will impact product quality. Product quality issues could be unmixed regions of dry material and hot spots having high concentrations of americium 241. Mixes that were more difficult to incorporate typically resulted in grouts with higher yield stresses. The mixability from these tests will provide Waste Solidification Building (WSB) an indication of which grouts will be more challenging to mix. The first yield stress measurements were statistically compared to a list of variables, specifically the batched chemicals used to make the acidic solutions. The first yield stress was also compared to the physical properties of the acidic solutions, physical and pH properties of the neutralized/pH adjusted solutions, and chemical and physical properties of the grout.

  4. Measurement of Tc-99 in Savannah River Site High Activity Waste

    SciTech Connect (OSTI)

    DiPrete, D.P.

    2003-03-27T23:59:59.000Z

    Waste cleanup efforts currently underway at the Savannah River Site have created a need to characterize Tc-99 in the various high activity waste matrices currently in Site inventories. The traditional method our laboratory used for analyzing Tc-99 in higher activity matrices was a solvent-solvent extraction method using Aliquat-336 in xylene, which resulted in the problematic generation of mixed waste. In an effort to eliminate the generation of mixed wastes resulting from the Aliquat 336/xylene process, a variety of different separation methodologies have been studied. Eichrom TEVA solid phase extractions using column technology have been employed in a case by case basis over the last several years. More recently, applications using Eichrom TEVA extraction discs and 3M Empore Tc extraction discs have also been explored.

  5. Balancing Cost and Risk by Optimizing the High-Level Waste and Low-Activity Waste Vitrification

    SciTech Connect (OSTI)

    Hrma, Pavel R.; Vienna, John D.

    2000-02-23T23:59:59.000Z

    In the currently used melters, the waste loading for nearly all high-level waste (HLW) is limited by crystallization. Above a certain level of waste loading, precipitation, settling, and accumulation of crystalline phases can cause severe processing problems and shorten the melter lifetime. To decrease the cost without putting the vitrification process at an unreasonable risk, several options, such as developing melters that operate above the liquidus temperature of glass, can be considered. Alternatively, if the melter is stirred, either mechanically, by bubbling, or by temperature gradients in induction heating, the melt can contain a substantial fraction of a crystalline phase that would not settle because it would be removed from the melter with glass. In addition, an induction melter can be nearly completely drained. For current melters that operate at a fixed temperature of 1150C, optimized glass formulation within currently accepted constaints has been developed. This approach is based on mathematically formulated relationships between glass properties and glass composition. Finally, re-evaluating the liquidus-temperature constraint, which may be unnecessarily restrictive for some HLWs, has recently been investigated. An attempt is being made to assess the rate of settling of crystalline phases in the melter and evaluate the risk for melter operation. Based on a reliable estimate of such a risk, waste loading could be increased, and a substantial saving can accrue. For low-activity waste (LAW), the waste loading in glass is limited either by the product quality or by segregation of sulfate during melting. The formulation of constraints on LAW glass in terms of relevant properties has not been completed, and no property-composition relationships have been established so far for this type of waste glass.

  6. UP2 400 High Activity Oxide Legacy Waste Retrieval Project Scope and Progress-13048

    SciTech Connect (OSTI)

    Chabeuf, Jean-Michel; Varet, Thierry [AREVA Site Value Development Business Unit, La Hague Site (France)] [AREVA Site Value Development Business Unit, La Hague Site (France)

    2013-07-01T23:59:59.000Z

    The High Activity Oxide facility (HAO) reprocessed sheared and dissolved 4500 metric tons of light water reactor fuel the fuel of the emerging light water reactor spent fuel between 1976 and 1998. Over the period, approximately 2200 tons of process waste, composed primarily of sheared hulls, was produced and stored in a vast silo in the first place, and in canisters stored in pools in subsequent years. Upon shutdown of the facility, AREVA D and D Division in La Hague launched a thorough investigation and characterization of the silos and pools content, which then served as input data for the definition of a legacy waste retrieval and reconditioning program. Basic design was conducted between 2005 and 2007, and was followed by an optimization phase which lead to the definition of a final scenario and budget, 12% under the initial estimates. The scenario planned for the construction of a retrieval and reconditioning cell to be built on top of the storage silo. The retrieved waste would then be rinsed and sorted, so that hulls could subsequently be sent to La Hague high activity compacting facility, while resins and sludge would be cemented within the retrieval cell. Detailed design was conducted successfully from 2008 until 2011, while a thorough research and development program was conducted in order to qualify each stage of the retrieval and reconditioning process, and assist in the elaboration of the final waste package specification. This R and D program was defined and conducted as a response and mitigation of the major project risks identified during the basic design process. Procurement and site preparatory works were then launched in 2011. By the end of 2012, R and D is nearly completed, the retrieval and reconditioning process have been secured, the final waste package specification is being completed, the first equipment for the retrieval cell is being delivered on site, while preparation works are allowing to free up space above and around the silo, to allow for construction which is scheduled to being during the first semester of 2013. The elaboration of the final waste package is still undergoing and expected to be completed by then end of 2013, following some final elements of R and D required to demonstrate the full compatibility of the package with deep geological repository. The HAO legacy waste retrieval project is so far the largest such project entering operational phase on the site of La Hague. It is on schedule, under budget, and in conformity with the delivery requirements set by the French Safety Authority, as well as other stakeholders. This project paves the way for the successful completion of AREVA La Hague other legacy waste retrieval projects, which are currently being drafted or already in active R and D phase. (authors)

  7. Vitrification of Low-Activity Radioactive Waste Streams and a High-Level Radioactive Waste Stream in Support of the Hanford River Protection Program

    SciTech Connect (OSTI)

    Crawford, C.L.

    2002-07-10T23:59:59.000Z

    Hanford tank waste consists of about 190 million curies in 54 million gallons of highly radioactive and mixed hazardous waste stored in underground storage tanks at the Hanford Site in Washington State. The tank waste includes solids (sludge), liquids (supernatant), and salt cake (dried salts that dissolve in water to form supernatant). The tank waste will be remediated through treatment and immobilization to protect the environment and meet regulatory requirements. The U.S. Department of Energy's (DOE's) preferred alternative to remediate the Hanford tank waste is to pretreat the waste by separating it into low-activity waste (LAW) and high-level waste (HLW), followed by immobilization of the LAW for on-site disposal and immobilization of the HLW for ultimate disposal in a national repository. This paper describes the crucible-scale vitrification and associated wasteform product tests in support of the WTP at Hanford. The two different LAW glasses produced in this study were from pretreated Envelope A (Tank 241-AN-103) and Envelope C (Tank 241-AN-102) waste. The HLW glass was produced from Tank C-106 HLW sludge and the HLW radionuclide products separated from Hanford Site tank samples AN-103, AN-102 and AZ-102. Pretreatment of these three supernates consisted of characterization, strontium and transuranics removal by precipitation and filtration, and final Cs-137 and Tc-99 removal by ion exchange (IX). The glasses were produced from formulations supplied by Vitreous State Laboratory of the Catholic University of America (CUA). Formulations were based on previous surrogate testing and the actual characterization data from the radioactive feed streams. Crucible-scale vitrifications were performed in platinum/gold crucibles in a custom-designed furnace fit with an offgas containment system. Both LAW and HLW melter feed slurries were evaporated, calcined, and then melted at 1150 degrees C. The LAW and HLW glasses were heat-treated per a modeled centerline cooling curve for the LAW canister and HLW canister, respectively.

  8. Independent Oversight Activity Report, Hanford Waste Treatment...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Department of Energy Office of River Protection review of the High Level Waste Facility heating, ventilation, and air conditioning systems. Independent Oversight Activity Report,...

  9. Extraction chromatographic separation of promethium from high active waste solutions of Purex origin

    SciTech Connect (OSTI)

    Ramanujam, A.; Achuthan, P.V.; Dhami, P.S.; Gopalakrishnan, V.; Kannan, R.; Mathur, J.N. [Bhabha Atomic Research Centre, Bombay (India)

    1995-03-01T23:59:59.000Z

    An extraction chromatographic procedure for the separation of {sup 147}Pm from High Active Waste solutions of Purex process has been developed. Octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide(CMPO) and 2-ethylhexyl-2-ethylhexylphosphonic acid (KSM-17), both sorbed separately on an inert support(chromosorb-102) have been sequentially employed for this purpose. In the CMPO column, the rare earths and the trivalent actinides are sorbed together with uranium, plutonium and traces of few other fission products. The elution of this column with 0.04 M HNO{sub 3} gives an eluate containing trivalent actinides and lanthanides. This solution, after adjusting the pH to 2.0, is used as feed for the second extraction chromatographic column based on KSM-17. All the trivalent metal ions are sorbed on the column leaving the trace impurities in the effluent. Fractional elution of the metal ions from this column is carried out with nitric acid of varying concentrations. At 0.09 M HNO{sub 3}, the pure beta emitting fraction of {sup 147}Pm has been obtained. 16 refs., 3 figs., 2 tabs.

  10. Waste to Energy Time Activities

    E-Print Network [OSTI]

    SEMINAR Waste to Energy Time Activities 9:30-9:40 Brief introduction of participants 9:40-10:10 Presentation of Dr. Kalogirou, "Waste to Energy: An Integral Part of Worldwide Sustainable Waste Management" 10. Sofia Bethanis, "Production of synthetic aggregates for use in structural concrete from waste to energy

  11. High-Level Waste Requirements

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1999-07-09T23:59:59.000Z

    The guide provides the criteria for determining which DOE radioactive wastes are to be managed as high-level waste in accordance with DOE M 435.1-1.

  12. MEASUREMENT AND CALCULATION OF RADIONUCLIDE ACTIVITIES IN SAVANNAH RIVER SITE HIGH LEVEL WASTE SLUDGE FOR ACCEPTANCE OF DEFENSE WASTE PROCESSING FACILITY GLASS IN A FEDERAL REPOSITORY

    SciTech Connect (OSTI)

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

    2008-12-31T23:59:59.000Z

    This paper describes the results of the analyses of High Level Waste (HLW) sludge slurry samples and of the calculations necessary to decay the radionuclides to meet the reporting requirement in the Waste Acceptance Product Specifications (WAPS) [1]. The concentrations of 45 radionuclides were measured. The results of these analyses provide input for radioactive decay calculations used to project the radionuclide inventory at the specified index years, 2015 and 3115. This information is necessary to complete the Production Records at Savannah River Site's Defense Waste Processing Facility (DWPF) so that the final glass product resulting from Macrobatch 5 (MB5) can eventually be submitted to a Federal Repository. Five of the necessary input radionuclides for the decay calculations could not be measured directly due to their low concentrations and/or analytical interferences. These isotopes are Nb-93m, Pd-107, Cd-113m, Cs-135, and Cm-248. Methods for calculating these species from concentrations of appropriate other radionuclides will be discussed. Also the average age of the MB5 HLW had to be calculated from decay of Sr-90 in order to predict the initial concentration of Nb-93m. As a result of the measurements and calculations, thirty-one WAPS reportable radioactive isotopes were identified for MB5. The total activity of MB5 sludge solids will decrease from 1.6E+04 {micro}Ci (1 {micro}Ci = 3.7E+04 Bq) per gram of total solids in 2008 to 2.3E+01 {micro}Ci per gram of total solids in 3115, a decrease of approximately 700 fold. Finally, evidence will be given for the low observed concentrations of the radionuclides Tc-99, I-129, and Sm-151 in the HLW sludges. These radionuclides were reduced in the MB5 sludge slurry to a fraction of their expected production levels due to SRS processing conditions.

  13. Independent Oversight Activity Report, Hanford Waste Treatment...

    Office of Environmental Management (EM)

    Observation of the Waste Treatment and Immobilization Plant Low Activity Waste Facility Heating, Ventilation, and Air Conditioning Systems Hazards Analysis Activities...

  14. Vitrification of high sulfate wastes

    SciTech Connect (OSTI)

    Merrill, R.A.; Whittington, K.F.; Peters, R.D.

    1994-09-01T23:59:59.000Z

    The US Department of Energy (DOE) through the Mixed Waste Integrated Program (MWIP) is investigating the application of vitrification technology to mixed wastes within the DOE system This work involves identifying waste streams, laboratory testing to identify glass formulations and characterize the vitrified product, and demonstration testing with the actual waste in a pilot-scale system. Part of this program is investigating process limits for various waste components, specifically those components that typically create problems for the application of vitrification, such as sulfate, chloride, and phosphate. This work describes results from vitrification testing for a high-sulfate waste, the 183-H Solar Evaporation Basin waste at Hanford. A low melting phosphate glass formulation has been developed for a waste stream high in sodium and sulfate. At melt temperatures in the range of 1,000 C to 1,200 C, sulfate in the waste is decomposed to gaseous oxides and driven off during melting, while the remainder of the oxides stay in the melt. Decomposition of the sulfates eliminates the processing problems typically encountered in vitrification of sulfate-containing wastes, resulting in separation of the sulfate from the remainder of the waste and allowing the sulfate to be collected in the off-gas system and treated as a secondary waste stream. Both the vitreous product and intentionally devitrified samples are durable when compared to reference glasses by TCLP and DI water leach tests. Simple, short tests to evaluate the compatibility of the glasses with potential melter materials found minimal corrosion with most materials.

  15. SECONDARY WASTE MANAGEMENT STRATEGY FOR EARLY LOW ACTIVITY WASTE TREATMENT

    SciTech Connect (OSTI)

    CRAWFORD TW

    2008-07-17T23:59:59.000Z

    This study evaluates parameters relevant to River Protection Project secondary waste streams generated during Early Low Activity Waste operations and recommends a strategy for secondary waste management that considers groundwater impact, cost, and programmatic risk. The recommended strategy for managing River Protection Project secondary waste is focused on improvements in the Effiuent Treatment Facility. Baseline plans to build a Solidification Treatment Unit adjacent to Effluent Treatment Facility should be enhanced to improve solid waste performance and mitigate corrosion of tanks and piping supporting the Effiuent Treatment Facility evaporator. This approach provides a life-cycle benefit to solid waste performance and reduction of groundwater contaminants.

  16. Independent Oversight Activity Report, Hanford Waste Treatment...

    Broader source: Energy.gov (indexed) [DOE]

    Hanford Waste Treatment and Immobilization Plant Engineering Activities and Tank Farm Operations HIAR-HANFORD-2014-01-13 This Independent Oversight Activity Report documents...

  17. Qualitative and Quantitative Assessment of Nuclear Materials Contained in High-Activity Waste Arising from the Operations at the 'SHELTER' Facility

    SciTech Connect (OSTI)

    Cherkas, Dmytro

    2011-10-01T23:59:59.000Z

    As a result of the nuclear accident at the Chernobyl NPP in 1986, the explosion dispeesed nuclear materials contained in the nuclear fuel of the reactor core over the destroyed facilities at Unit No. 4 and over the territory immediately adjacent to the destroyed unit. The debris was buried under the Cascade Wall. Nuclear materials at the SHELTER can be characterized as spent nuclear fuel, fresh fuel assemblies (including fuel assemblies with damaged geometry and integrity, and individual fuel elements), core fragments of the Chernobyl NPP Unit No. 4, finely-dispersed fuel (powder/dust), uranium and plutonium compounds in water solutions, and lava-like nuclear fuel-containing masses. The new safe confinement (NSC) is a facility designed to enclose the Chernobyl NPP Unit No. 4 destroyed by the accident. Construction of the NSC involves excavating operations, which are continuously monitored including for the level of radiation. The findings of such monitoring at the SHELTER site will allow us to characterize the recovered radioactive waste. When a process material categorized as high activity waste (HAW) is detected the following HLW management operations should be involved: HLW collection; HLW fragmentation (if appropriate); loading HAW into the primary package KT-0.2; loading the primary package filled with HAW into the transportation cask KTZV-0.2; and storing the cask in temporary storage facilities for high-level solid waste. The CDAS system is a system of 3He tubes for neutron coincidence counting, and is designed to measure the percentage ratio of specific nuclear materials in a 200-liter drum containing nuclear material intermixed with a matrix. The CDAS consists of panels with helium counter tubes and a polyethylene moderator. The panels are configured to allow one to position a waste-containing drum and a drum manipulator. The system operates on the ‘add a source’ basis using a small Cf-252 source to identify irregularities in the matrix during an assay. The platform with the source is placed under the measurement chamber. The platform with the source material is moved under the measurement chamber. The design allows one to move the platform with the source in and out, thus moving the drum. The CDAS system and radioactive waste containers have been built. For each drum filled with waste two individual measurements (passive/active) will be made. This paper briefly describes the work carried out to assess qualitatively and quantitatively the nuclear materials contained in high-level waste at the SHELTER facility. These efforts substantially increased nuclear safety and security at the facility.

  18. Permitting plan for the immobilized low-activity waste project

    SciTech Connect (OSTI)

    Deffenbaugh, M.L.

    1997-09-04T23:59:59.000Z

    This document addresses the environmental permitting requirements for the transportation and interim storage of the Immobilized Low-Activity Waste (ILAW) produced during Phase 1 of the Hanford Site privatization effort. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage and disposal of Tank Waste Remediation Systems (TWRS) immobilized low-activity tank waste (ILAW) and (2) interim storage of TWRS immobilized HLW (IHLW) and other canistered high-level waste forms. Low-activity waste (LAW), low-level waste (LLW), and high-level waste (HLW) are defined by the TWRS, Hanford Site, Richland, Washington, Final Environmental Impact Statement (EIS) DOE/EIS-0189, August 1996 (TWRS, Final EIS). By definition, HLW requires permanent isolation in a deep geologic repository. Also by definition, LAW is ``the waste that remains after separating from high-level waste as much of the radioactivity as is practicable that when solidified may be disposed of as LLW in a near-surface facility according to the NRC regulations.`` It is planned to store/dispose of (ILAW) inside four empty vaults of the five that were originally constructed for the Group Program. Additional disposal facilities will be constructed to accommodate immobilized LLW packages produced after the Grout Vaults are filled. The specifications for performance of the low-activity vitrified waste form have been established with strong consideration of risk to the public. The specifications for glass waste form performance are being closely coordinated with analysis of risk. RL has pursued discussions with the NRC for a determination of the classification of the Hanford Site`s low-activity tank waste fraction. There is no known RL action to change law with respect to onsite disposal of waste.

  19. Independent Oversight Activity Report, Hanford Waste Treatment...

    Office of Environmental Management (EM)

    that focuses on the technical adequacy of the low activity waste (LAW) documented safety analysis and supporting basis. BNI uses the VSL experimental data and reports to...

  20. Hanford's Simulated Low Activity Waste Cast Stone Processing

    SciTech Connect (OSTI)

    Kim, Young

    2013-08-20T23:59:59.000Z

    Cast Stone is undergoing evaluation as the supplemental treatment technology for Hanford’s (Washington) high activity waste (HAW) and low activity waste (LAW). This report will only cover the LAW Cast Stone. The programs used for this simulated Cast Stone were gradient density change, compressive strength, and salt waste form phase identification. Gradient density changes show a favorable outcome by showing uniformity even though it was hypothesized differently. Compressive strength exceeded the minimum strength required by Hanford and greater compressive strength increase seen between the uses of different salt solution The salt waste form phase is still an ongoing process as this time and could not be concluded.

  1. High Level Waste System Plan Revision 9

    SciTech Connect (OSTI)

    Davis, N.R.; Wells, M.N.; Choi, A.S.; Paul, P.; Wise, F.E.

    1998-04-01T23:59:59.000Z

    Revision 9 of the High Level Waste System Plan documents the current operating strategy of the HLW System at SRS to receive, store, treat, and dispose of high-level waste.

  2. Hanford Tank Waste - Near Source Treatment of Low Activity Waste

    SciTech Connect (OSTI)

    Ramsey, William Gene

    2013-08-15T23:59:59.000Z

    Abstract only. Treatment and disposition of Hanford Site waste as currently planned consists of 100+ waste retrievals, waste delivery through up to 8+ miles of dedicated, in-ground piping, centralized mixing and blending operations- all leading to pre-treatment combination and separation processes followed by vitrification at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The sequential nature of Tank Farm and WTP operations requires nominally 15-20 years of continuous operations before all waste can be retrieved from many Single Shell Tanks (SSTs). Also, the infrastructure necessary to mobilize and deliver the waste requires significant investment beyond that required for the WTP. Treating waste as closely as possible to individual tanks or groups- as allowed by the waste characteristics- is being investigated to determine the potential to 1) defer, reduce, and/or eliminate infrastructure requirements, and 2) significantly mitigate project risk by reducing the potential and impact of single point failures. The inventory of Hanford waste slated for processing and disposition as LAW is currently managed as high-level waste (HLW), i.e., the separation of fission products and other radionuclides has not commenced. A significant inventory of this waste (over 20M gallons) is in the form of precipitated saltcake maintained in single shell tanks, many of which are identified as potential leaking tanks. Retrieval and transport (as a liquid) must be staged within the waste feed delivery capability established by site infrastructure and WTP. Near Source treatment, if employed, would provide for the separation and stabilization processing necessary for waste located in remote farms (wherein most of the leaking tanks reside) significantly earlier than currently projected. Near Source treatment is intended to address the currently accepted site risk and also provides means to mitigate future issues likely to be faced over the coming decades. This paper describes the potential near source treatment and waste disposition options as well as the impact these options could have on reducing infrastructure requirements, project cost and mission schedule.

  3. Pump Jet Mixing and Pipeline Transfer Assessment for High-Activity Radioactive Wastes in Hanford Tank 241-AZ-102

    SciTech Connect (OSTI)

    Y Onishi; KP Recknagle; BE Wells

    2000-08-09T23:59:59.000Z

    The authors evaluated how well two 300-hp mixer pumps would mix solid and liquid radioactive wastes stored in Hanford double-shell Tank 241-AZ-102 (AZ-102) and confirmed the adequacy of a three-inch (7.6-cm) pipeline system to transfer the resulting mixed waste slurry to the AP Tank Farm and a planned waste treatment (vitrification) plant on the Hanford Site. Tank AZ-102 contains 854,000 gallons (3,230 m{sup 3}) of supernatant liquid and 95,000 gallons (360 m{sup 3}) of sludge made up of aging waste (or neutralized current acid waste). The study comprises three assessments: waste chemistry, pump jet mixing, and pipeline transfer. The waste chemical modeling assessment indicates that the sludge, consisting of the solids and interstitial solution, and the supernatant liquid are basically in an equilibrium condition. Thus, pump jet mixing would not cause much solids precipitation and dissolution, only 1.5% or less of the total AZ-102 sludge. The pump jet mixing modeling indicates that two 300-hp mixer pumps would mobilize up to about 23 ft (7.0 m) of the sludge nearest the pump but would not erode the waste within seven inches (0.18 m) of the tank bottom. This results in about half of the sludge being uniformly mixed in the tank and the other half being unmixed (not eroded) at the tank bottom.

  4. Tank waste remediation system phase I high-level waste feed processability assessment report

    SciTech Connect (OSTI)

    Lambert, S.L.; Stegen, G.E., Westinghouse Hanford

    1996-08-01T23:59:59.000Z

    This report evaluates the effects of feed composition on the Phase I high-level waste immobilization process and interim storage facility requirements for the high-level waste glass.Several different Phase I staging (retrieval, blending, and pretreatment) scenarios were used to generate example feed compositions for glass formulations, testing, and glass sensitivity analysis. Glass models and data form laboratory glass studies were used to estimate achievable waste loading and corresponding glass volumes for various Phase I feeds. Key issues related to feed process ability, feed composition, uncertainty, and immobilization process technology are identified for future consideration in other tank waste disposal program activities.

  5. Hanford Immobilized Low-Activity Waste Product Acceptance Test Plan

    SciTech Connect (OSTI)

    Peeler, D.

    1999-06-22T23:59:59.000Z

    'The Hanford Site has been used to produce nuclear materials for the U.S. Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during Pu production, exists in 177 underground single- and double-shell tanks. These wastes are to be retrieved and separated into low-activity waste (LAW) and high-level waste (HLW) fractions. The DOE is proceeding with an approach to privatize the treatment and immobilization of Handord''s LAW and HLW.'

  6. An Overview of Thermoelectric Waste Heat Recovery Activities...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    An Overview of Thermoelectric Waste Heat Recovery Activities in Europe An Overview of Thermoelectric Waste Heat Recovery Activities in Europe An overview presentation of R&D...

  7. PAIRWISE BLENDING OF HIGH LEVEL WASTE (HLW)

    SciTech Connect (OSTI)

    CERTA, P.J.

    2006-02-22T23:59:59.000Z

    The primary objective of this study is to demonstrate a mission scenario that uses pairwise and incidental blending of high level waste (HLW) to reduce the total mass of HLW glass. Secondary objectives include understanding how recent refinements to the tank waste inventory and solubility assumptions affect the mass of HLW glass and how logistical constraints may affect the efficacy of HLW blending.

  8. Low and high Temperature Dual Thermoelectric Generation Waste...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    and high Temperature Dual Thermoelectric Generation Waste Heat Recovery System for Light-Duty Vehicles Low and high Temperature Dual Thermoelectric Generation Waste Heat Recovery...

  9. Phosphate Glasses for Vitrification of Waste with High Sulfur Content

    SciTech Connect (OSTI)

    Kim, Dong-Sang; Vienna, John D.; Hrma, Pavel R.; Cassingham, Nathan J.

    2002-10-31T23:59:59.000Z

    The low solubility of sulfate in silicate-based glasses, approximately 1 mass% as SO3, limits the loading of high-level waste (HLW) and low-activity waste (LAW) containing high concentrations of sulfur. Based on crucible melting studies, we have shown that the phosphate glasses may incorporate more than 5 mass% SO3; hence, the waste loading can be increased until another constraint is met, such as glass durability. A high-sulfate HLW glass has been formulated and tested to demonstrate the advantages of phosphate glasses. The effect of waste loading on the chemical durability of quenched and slow-cooled phosphate glasses was determined using the Product Consistency Test.

  10. activity radioactive waste: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    is: not high-level radioactive waste or irradiated nuclear fuel not uranium, thorium or other ore tailings or waste from extraction and concentration for source material...

  11. activities radioactive waste: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    is: not high-level radioactive waste or irradiated nuclear fuel not uranium, thorium or other ore tailings or waste from extraction and concentration for source material...

  12. activity radioactive wastes: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    is: not high-level radioactive waste or irradiated nuclear fuel not uranium, thorium or other ore tailings or waste from extraction and concentration for source material...

  13. Stability of High Level Radioactive Waste Forms

    SciTech Connect (OSTI)

    Besmann, T.M.; Kulkarni, N.S.; Spear, K.E.; Vienna, J.D.; Hanni, J.B.; Crum, J.D.; Hrma, P.

    2005-01-20T23:59:59.000Z

    This presentation was given at the DOE Office of Science-Environmental Management Science Program (EMSP) High-Level Waste Workshop held on January 19-20, 2005 at the Savannah River Site.

  14. Crystallization in High-Level Waste Glasses

    SciTech Connect (OSTI)

    Hrma, Pavel R. (BATTELLE (PACIFIC NW LAB)); Dane R Spearing, Gary L Smith, SK Sundaram

    2002-01-01T23:59:59.000Z

    This review outlines important aspects of crystallization in HLW glasses, such as equilibrium, nucleation, growth, and dissolution. The impact of crystallization on continuous melters and the chemical durability of high-level waste glass are briefly discussed.

  15. Hanford immobilized low-activity tank waste performance assessment

    SciTech Connect (OSTI)

    Mann, F.M.

    1998-03-26T23:59:59.000Z

    The Hanford Immobilized Low-Activity Tank Waste Performance Assessment examines the long-term environmental and human health effects associated with the planned disposal of the vitrified low-level fraction of waste presently contained in Hanford Site tanks. The tank waste is the by-product of separating special nuclear materials from irradiated nuclear fuels over the past 50 years. This waste has been stored in underground single and double-shell tanks. The tank waste is to be retrieved, separated into low and high-activity fractions, and then immobilized by private vendors. The US Department of Energy (DOE) will receive the vitrified waste from private vendors and plans to dispose of the low-activity fraction in the Hanford Site 200 East Area. The high-level fraction will be stored at Hanford until a national repository is approved. This report provides the site-specific long-term environmental information needed by the DOE to issue a Disposal Authorization Statement that would allow the modification of the four existing concrete disposal vaults to provide better access for emplacement of the immobilized low-activity waste (ILAW) containers; filling of the modified vaults with the approximately 5,000 ILAW containers and filler material with the intent to dispose of the containers; construction of the first set of next-generation disposal facilities. The performance assessment activity will continue beyond this assessment. The activity will collect additional data on the geotechnical features of the disposal sites, the disposal facility design and construction, and the long-term performance of the waste. Better estimates of long-term performance will be produced and reviewed on a regular basis. Performance assessments supporting closure of filled facilities will be issued seeking approval of those actions necessary to conclude active disposal facility operations. This report also analyzes the long-term performance of the currently planned disposal system as a basis to set requirements on the waste form and the facility design that will protect the long-term public health and safety and protect the environment.

  16. High-Level Waste Melter Study Report

    SciTech Connect (OSTI)

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

    2001-07-13T23:59:59.000Z

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

  17. EIS-0287: Idaho High-Level Waste & Facilities Disposition

    Broader source: Energy.gov [DOE]

    This EIS analyzes the potential environmental consequences of alternatives for managing high-level waste (HLW) calcine, mixed transuranic waste/sodium bearing waste (SBW) and newly generated liquid...

  18. High-level waste melter alternatives assessment report

    SciTech Connect (OSTI)

    Calmus, R.B.

    1995-02-01T23:59:59.000Z

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

  19. Hanford Low Activity Waste (LAW) Fluidized Bed Steam Reformer...

    Office of Environmental Management (EM)

    Low Activity Waste (LAW) Fluidized Bed Steam Reformer (FBSR) Na-Al-Si (NAS) Waste Form Qualification C.M. Jantzen and E.M. Pierce November 18, 2010 2 Participating Organizations 3...

  20. Waste Treatment Technology Process Development Plan For Hanford Waste Treatment Plant Low Activity Waste Recycle

    SciTech Connect (OSTI)

    McCabe, Daniel J.; Wilmarth, William R.; Nash, Charles A.

    2013-08-29T23:59:59.000Z

    The purpose of this Process Development Plan is to summarize the objectives and plans for the technology development activities for an alternative path for disposition of the recycle stream that will be generated in the Hanford Waste Treatment Plant Low Activity Waste (LAW) vitrification facility (LAW Recycle). This plan covers the first phase of the development activities. The baseline plan for disposition of this stream is to recycle it to the WTP Pretreatment Facility, where it will be concentrated by evaporation and returned to the LAW vitrification facility. Because this stream contains components that are volatile at melter temperatures and are also problematic for the glass waste form, they accumulate in the Recycle stream, exacerbating their impact on the number of LAW glass containers. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and reducing the halides in the Recycle is a key component of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, this stream does not have a proven disposition path, and resolving this gap becomes vitally important. This task seeks to examine the impact of potential future disposition of this stream in the Hanford tank farms, and to develop a process that will remove radionuclides from this stream and allow its diversion to another disposition path, greatly decreasing the LAW vitrification mission duration and quantity of glass waste. The origin of this LAW Recycle stream will be from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover or precipitates of scrubbed components (e.g. carbonates). The soluble components are mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet, and will not be available until the WTP begins operation, causing uncertainty in its composition, particularly the radionuclide content. This plan will provide an estimate of the likely composition and the basis for it, assess likely treatment technologies, identify potential disposition paths, establish target treatment limits, and recommend the testing needed to show feasibility. Two primary disposition options are proposed for investigation, one is concentration for storage in the tank farms, and the other is treatment prior to disposition in the Effluent Treatment Facility. One of the radionuclides that is volatile and expected to be in high concentration in this LAW Recycle stream is Technetium-99 ({sup 99}Tc), a long-lived radionuclide with a half-life of 210,000 years. Technetium will not be removed from the aqueous waste in the Hanford Waste Treatment and Immobilization Plant (WTP), and will primarily end up immobilized in the LAW glass, which will be disposed in the Integrated Disposal Facility (IDF). Because {sup 99}Tc has a very long half-life and is highly mobile, it is the largest dose contributor to the Performance Assessment (PA) of the IDF. Other radionuclides that are also expected to be in appreciable concentration in the LAW Recycle are {sup 129}I, {sup 90}Sr, {sup 137}Cs, and {sup 241}Am. The concentrations of these radionuclides in this stream will be much lower than in the LAW, but they will still be higher than limits for some of the other disposition pathways currently available. Although the baseline process will recycle this stream to the Pretreatment Facility, if the LAW facility begins operation first, this stream will not have a disposition path internal to WTP. One potential solution is to return the stream to the tank farms where it can be evaporated in the 242-A evaporator, or perhaps deploy an auxiliary evaporator to concentrate it prior to return to the tank farms. In either case, testing is needed to evaluat

  1. High-level radioactive wastes. Supplement 1

    SciTech Connect (OSTI)

    McLaren, L.H. (ed.)

    1984-09-01T23:59:59.000Z

    This bibliography contains information on high-level radioactive wastes included in the Department of Energy's Energy Data Base from August 1982 through December 1983. These citations are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes, each preceded by a brief description, are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number. 1452 citations.

  2. High Level Waste Management Division High. Level Waste System Plan

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC) EnvironmentalGyroSolé(tm) Harmonicbet WhenHiggs Boson May| ArgonneHigh Level

  3. High-level waste qualification: Managing uncertainty

    SciTech Connect (OSTI)

    Pulsipher, B.A. [Pacific Northwest Lab., Richland, WA (United States)

    1993-12-31T23:59:59.000Z

    Qualification of high-level waste implies specifications driven by risk against which performance can be assessed. The inherent uncertainties should be addressed in the specifications and statistical methods should be employed to appropriately manage these uncertainties. Uncertainties exist whenever measurements are obtained, sampling is employed, or processes are affected by systematic or random perturbations. This paper presents the approach and statistical methods currently employed by Pacific Northwest Laboratory (PNL) and West Valley Nuclear Services (WVNS) to characterize, minimize, and control uncertainties pertinent to a waste-form acceptance specification concerned with product consistency.

  4. Technical basis for classification of low-activity waste fraction from Hanford site tanks

    SciTech Connect (OSTI)

    Petersen, C.A.

    1996-09-20T23:59:59.000Z

    The overall objective of this report is to provide a technical basis to support a U.S. Nuclear Regulatory Commission determination to classify the low-activity waste from the Hanford Site single-shell and double-shell tanks as `incidental` wastes after removal of additional radionuclides and immobilization.The proposed processing method, in addition to the previous radionuclide removal efforts, will remove the largest practical amount of total site radioactivity, attributable to high-level waste, for disposal is a deep geologic repository. The remainder of the waste would be considered `incidental` waste and could be disposed onsite.

  5. activities waste fixation: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Peter S 1998-01-01 34 Calculations of Induced Activity in the ATLAS Experiment for Nuclear Waste Zoning. CERN Preprints Summary: Extensive calculations were performed with...

  6. activity waste ilaw: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    composting processes and related student, volunteer and staff activities. Desired 34 Assessing Information Waste in Lean Product Development MIT - DSpace Summary: Lean Product...

  7. Development of high-waste loaded high-level nuclear waste glasses for high-temperature melter

    SciTech Connect (OSTI)

    Kim, D.S.; Hrma, P.; Lamar, D.A.; Elliott, M.L. [Pacific Northwest Lab., Richland, WA (United States)

    1994-12-31T23:59:59.000Z

    This paper describes the approach taken in formulating glasses that can be processed at 1150 to 1500{degrees}C by applying glass property/composition models developed at Pacific Northwest Laboratory. Compositions and melting temperatures for glasses with high waste loading that are acceptable and able to be processed were determined for two different Hanford waste types. The glasses meet high-level waste glass acceptability criteria and are suitable for processing in a continuous Joule-heated melter.

  8. Development of high-waste loaded high-level nuclear waste glasses for high-temperature melter

    SciTech Connect (OSTI)

    Kim, D.S.; Hrma, P.R.; Lamar, D.A.; Elliott, M.L.

    1994-04-01T23:59:59.000Z

    This paper describes the approach taken in formulating glasses that can be processed at 1150 to 1500{degrees}C by applying glass property/composition models developed at Pacific Northwest Laboratory. Compositions and melting temperatures for glasses with high waste loading that are acceptable and able to be processed were determined for two different Hanford waste types. The glasses meet high-level waste glass acceptability criteria and are suitable for processing in a continuous Joule-heated melter.

  9. Identification of items and activities important to waste form acceptance by Westinghouse GoCo sites

    SciTech Connect (OSTI)

    Plodinec, M.J.; Marra, S.L. [Westinghouse Savannah River Co., Aiken, SC (United States); Dempster, J. [West Valley Demonstration Project, NY (United States); Randklev, E.H. [Hanford Waste Vitrification Plant (United States)

    1993-10-12T23:59:59.000Z

    The Department of Energy has established specifications (Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms, or WAPS) for canistered waste forms produced at Hanford, Savannah River, and West Valley. Compliance with these specifications requires that each waste form producer identify the items and activities which must be controlled to ensure compliance. As part of quality assurance oversight activities, reviewers have tried to compare the methodologies used by the waste form producers to identify items and activities important to waste form acceptance. Due to the lack of a documented comparison of the methods used by each producer, confusion has resulted over whether the methods being used are consistent. This confusion has been exacerbated by different systems of nomenclature used by each producer, and the different stages of development of each project. The waste form producers have met three times in the last two years, most recently on June 28, 1993, to exchange information on each producer`s program. These meetings have been sponsored by the Westinghouse GoCo HLW Vitrification Committee. This document is the result of this most recent exchange. It fills the need for a documented comparison of the methodologies used to identify items and activities important to waste form acceptance. In this document, the methodology being used by each waste form producer is summarized, and the degree of consistency among the waste form producers is determined.

  10. Repackaging of High Fissile TRU Waste at the Transuranic Waste Processing Center - 13240

    SciTech Connect (OSTI)

    Oakley, Brian; Heacker, Fred [WAI, TRU Waste Processing Center, 100 WIPP Road Lenoir City, TN 37771 (United States)] [WAI, TRU Waste Processing Center, 100 WIPP Road Lenoir City, TN 37771 (United States); McMillan, Bill [DOE, Oak Ridge Operations, Bldg. 2714, Oak Ridge, TN 37830 (United States)] [DOE, Oak Ridge Operations, Bldg. 2714, Oak Ridge, TN 37830 (United States)

    2013-07-01T23:59:59.000Z

    Twenty-six drums of high fissile transuranic (TRU) waste from Oak Ridge National Laboratory (ORNL) operations were declared waste in the mid-1980's and placed in storage with the legacy TRU waste inventory for future treatment and disposal at the Waste Isolation Pilot Plant (WIPP). Repackaging and treatment of the waste at the TRU Waste Packaging Center (TWPC) will require the installation of additional equipment and capabilities to address the hazards for handling and repackaging the waste compared to typical Contact Handled (CH) TRU waste that is processed at the TWPC, including potential hydrogen accumulation in legacy 6M/2R packaging configurations, potential presence of reactive plutonium hydrides, and significant low energy gamma radiation dose rates. All of the waste is anticipated to be repackaged at the TWPC and certified for disposal at WIPP. The waste is currently packaged in multiple layers of containers which presents additional challenges for repackaging activities due to the potential for the accumulation of hydrogen gas in the container headspace in quantities than could exceed the Lower Flammability Limit (LFL). The outer container for each waste package is a stainless steel 0.21 m{sup 3} (55-gal) drum which contains either a 0.04 m{sup 3} or 0.06 m{sup 3} (10-gal or 15-gal) 6M drum. The inner 2R container in each 6M drum is ?12 cm (5 in) outside diameter x 30-36 cm (12-14 in) long and is considered to be a > 4 liter sealed container relative to TRU waste packaging criteria. Inside the 2R containers are multiple configurations of food pack cans, pipe nipples, and welded capsules. The waste contains significant quantities of high burn-up plutonium oxides and metals with a heavy weight percentage of higher atomic mass isotopes and the subsequent in-growth of significant quantities of americium. Significant low energy gamma radiation is expected to be present due to the americium in-growth. Radiation dose rates on inner containers are estimated to be 1-3 mSv/hr (100-300 mrem/hr) with an unshielded dose rate on the waste itself of over 10 mSv/hr (1 rem/hr). Additional equipment to be installed at the TWPC will include a new perma-con enclosure and a shielded/inert glovebox in the process building to repackage and stabilize the waste. All of the waste will be repackaged into Standard Pipe Overpacks. Most of the waste (21 of the 26 drums) is expected to be repackaged at the food-pack can level (i.e. the food-pack cans will not be opened). Five of the incoming waste containers are expected to be repackaged at the primary waste level. Three of the containers exceed the 200 gram Pu-239 Fissile Gram Equivalent (FGE) limit for the Standard Pipe Overpack. These three containers will be repackaged down to the primary waste level and divided into eight Standard Pipe Overpacks for shipment to WIPP. Two containers must be stabilized to eliminate any reactive plutonium hydrides that may be present. These containers will be opened in the inert, shielded glovebox, and the remaining corroded plutonium metal converted to a stable oxide form by using a 600 deg. C tube furnace with controlled oxygen feed in a helium carrier gas. The stabilized waste will then be packaged into two Standard Pipe Overpacks. Design and build out activities for the additional repackaging capabilities at the TWPC are scheduled to begin in Fiscal Year 2013 with repackaging, stabilization, and certification activities scheduled to begin in Fiscal Year 2014. Following repackaging and stabilization activities, the Standard Pipe Overpacks will be certified for disposal at WIPP utilizing Non-Destructive Examination (NDE) to verify the absence of prohibited items and Non-Destructive Assay (NDA) to verify the isotopic content under the TWPC WIPP certification program implemented by the Central Characterization Project (CCP). (authors)

  11. Stabilization of high and low solids Consolidated Incinerator Facility (CIF) waste with super cement

    SciTech Connect (OSTI)

    Walker, B.W.

    2000-01-11T23:59:59.000Z

    This report details solidification activities using selected Mixed Waste Focus Area technologies with the High and Low Solid waste streams. Ceramicrete and Super Cement technologies were chosen as the best possible replacement solidification candidates for the waste streams generated by the SRS incinerator from a list of several suggested Mixed Waste Focus Area technologies. These technologies were tested, evaluated, and compared to the current Portland cement technology being employed. Recommendation of a technology for replacement depends on waste form performance, process flexibility, process complexity, and cost of equipment and/or raw materials.

  12. High Waste Loading Glass Formulations for Hanford High-Aluminum High-Level Waste Streams

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC) EnvironmentalGyroSolé(tm) Harmonicbet WhenHiggs BosonAccurate knowledge ofHIGH WASTE

  13. Waste management activities and carbon emissions in Africa

    SciTech Connect (OSTI)

    Couth, R. [University of KwaZulu-Natal, CRECHE, School of Civil Engineering, Survey and Construction, Durban 4041 (South Africa); Trois, C., E-mail: troisc@ukzn.ac.za [University of KwaZulu-Natal, CRECHE, School of Civil Engineering, Survey and Construction, Durban 4041 (South Africa)

    2011-01-15T23:59:59.000Z

    This paper summarizes research into waste management activities and carbon emissions from territories in sub-Saharan Africa with the main objective of quantifying emission reductions (ERs) that can be gained through viable improvements to waste management in Africa. It demonstrates that data on waste and carbon emissions is poor and generally inadequate for prediction models. The paper shows that the amount of waste produced and its composition are linked to national Gross Domestic Product (GDP). Waste production per person is around half that in developed countries with a mean around 230 kg/hd/yr. Sub-Saharan territories produce waste with a biogenic carbon content of around 56% (+/-25%), which is approximately 40% greater than developed countries. This waste is disposed in uncontrolled dumps that produce large amounts of methane gas. Greenhouse gas (GHG) emissions from waste will rise with increasing urbanization and can only be controlled through funding mechanisms from developed countries.

  14. HIGH ALUMINUM HLW (HIGH LEVEL WASTE ) GLASSES FOR HANFORDS WTP (WASTE TREATMENT PROJECT)

    SciTech Connect (OSTI)

    KRUGER AA; BOWAN BW; JOSEPH I; GAN H; KOT WK; MATLACK KS; PEGG IL

    2010-01-04T23:59:59.000Z

    This paper presents the results of glass formulation development and melter testing to identify high waste loading glasses to treat high-Al high level waste (HLW) at Hanford. Previous glass formulations developed for this HLW had high waste loadings but their processing rates were lower that desired. The present work was aimed at improving the glass processing rate while maintaining high waste loadings. Glass formulations were designed, prepared at crucible-scale and characterized to determine their properties relevant to processing and product quality. Glass formulations that met these requirements were screened for melt rates using small-scale tests. The small-scale melt rate screening included vertical gradient furnace (VGF) and direct feed consumption (DFC) melter tests. Based on the results of these tests, modified glass formulations were developed and selected for larger scale melter tests to determine their processing rate. Melter tests were conducted on the DuraMelter 100 (DMIOO) with a melt surface area of 0.11 m{sup 2} and the DuraMelter 1200 (DMI200) HLW Pilot Melter with a melt surface area of 1.2 m{sup 2}. The newly developed glass formulations had waste loadings as high as 50 wt%, with corresponding Al{sub 2}O{sub 3} concentration in the glass of 26.63 wt%. The new glass formulations showed glass production rates as high as 1900 kg/(m{sup 2}.day) under nominal melter operating conditions. The demonstrated glass production rates are much higher than the current requirement of 800 kg/(m{sup 2}.day) and anticipated future enhanced Hanford Tank Waste Treatment and Immobilization Plant (WTP) requirement of 1000 kg/(m{sup 2}.day).

  15. High Level Waste ManagemenfDivision ..

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC) EnvironmentalGyroSolé(tm) Harmonicbet WhenHiggs Boson May| ArgonneHigh Level Waste

  16. Supplemental Immobilization of Hanford Low-Activity Waste: Cast Stone Screening Tests

    SciTech Connect (OSTI)

    Westsik, Joseph H.; Piepel, Gregory F.; Lindberg, Michael J.; Heasler, Patrick G.; Mercier, Theresa M.; Russell, Renee L.; Cozzi, Alex; Daniel, William E.; Eibling, Russell E.; Hansen, E. K.; Reigel, Marissa M.; Swanberg, David J.

    2013-09-30T23:59:59.000Z

    More than 56 million gallons of radioactive and hazardous waste are stored in 177 underground storage tanks at the U.S. Department of Energy’s (DOE’s) Hanford Site in southeastern Washington State. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the wastes and immobilize them in a glass waste form. The WTP includes a pretreatment facility to separate the wastes into a small volume of high-level waste (HLW) containing most of the radioactivity and a larger volume of low-activity waste (LAW) containing most of the nonradioactive chemicals. The HLW will be converted to glass in the HLW vitrification facility for ultimate disposal at an offsite federal repository. At least a portion (~35%) of the LAW will be converted to glass in the LAW vitrification facility and will be disposed of onsite at the Integrated Disposal Facility (IDF). The pretreatment and HLW vitrification facilities will have the capacity to treat and immobilize the wastes destined for each facility. However, a second LAW immobilization facility will be needed for the expected volume of LAW requiring immobilization. A cementitious waste form known as Cast Stone is being considered to provide the required additional LAW immobilization capacity. The Cast Stone waste form must be acceptable for disposal in the IDF. The Cast Stone waste form and immobilization process must be tested to demonstrate that the final Cast Stone waste form can comply with the waste acceptance criteria for the disposal facility and that the immobilization processes can be controlled to consistently provide an acceptable waste form product. Further, the waste form must be tested to provide the technical basis for understanding the long-term performance of the waste form in the disposal environment. These waste form performance data are needed to support risk assessment and performance assessment (PA) analyses of the long-term environmental impact of the waste disposal in the IDF. The PA is needed to satisfy both Washington State IDF Permit and DOE Order requirements. Cast Stone has been selected for solidification of radioactive wastes including WTP aqueous secondary wastes treated at the Effluent Treatment Facility (ETF) at Hanford. A similar waste form called Saltstone is used at the Savannah River Site (SRS) to solidify its LAW tank wastes.

  17. LOW ACTIVITY WASTE FEED SOLIDS CARACTERIZATION AND FILTERABILITY TESTS

    SciTech Connect (OSTI)

    McCabe, D.; Crawford, C.; Duignan, M.; Williams, M.; Burket, P.

    2014-04-03T23:59:59.000Z

    The primary treatment of the tank waste at the DOE Hanford site will be done in the Waste Treatment and Immobilization Plant (WTP) that is currently under construction. The baseline plan for the WTP Pretreatment facility is to treat the waste, splitting it into High Level Waste (HLW) feed and Low Activity Waste (LAW) feed. Both waste streams are then separately vitrified as glass and sealed in canisters. The LAW glass will be disposed onsite in the Integrated Disposal Facility (IDF). There are currently no plans to treat the waste to remove technetium in the WTP Pretreatment facility, so its disposition path is the LAW glass. Options are being explored to immobilize the LAW portion of the tank waste, i.e., the LAW feed from the WTP Pretreatment facility. Removal of {sup 99}Tc from the LAW Feed, followed by off-site disposal of the {sup 99}Tc, would eliminate a key risk contributor for the IDF Performance Assessment (PA) for supplemental waste forms, and has potential to reduce treatment and disposal costs. Washington River Protection Solutions (WRPS) is developing some conceptual flow sheets for LAW treatment and disposal that could benefit from technetium removal. One of these flowsheets will specifically examine removing {sup 99}Tc from the LAW feed stream to supplemental immobilization. The conceptual flow sheet of the {sup 99}Tc removal process includes a filter to remove insoluble solids prior to processing the stream in an ion exchange column, but the characteristics and behavior of the liquid and solid phases has not previously been investigated. This report contains results of testing of a simulant that represents the projected composition of the feed to the Supplemental LAW process. This feed composition is not identical to the aqueous tank waste fed to the Waste Treatment Plant because it has been processed through WTP Pretreatment facility and therefore contains internal changes and recycle streams that will be generated within the WTP process. Although a Supplemental LAW feed simulant has previously been prepared, this feed composition differs from that simulant because those tests examined only the fully soluble aqueous solution at room temperature, not the composition formed after evaporation, including the insoluble solids that precipitate after it cools. The conceptual flow sheet for Supplemental LAW immobilization has an option for removal of {sup 99}Tc from the feed stream, if needed. Elutable ion exchange has been selected for that process. If implemented, the stream would need filtration to remove the insoluble solids prior to processing in an ion exchange column. The characteristics, chemical speciation, physical properties, and filterability of the solids are important to judge the feasibility of the concept, and to estimate the size and cost of a facility. The insoluble solids formed during these tests were primarily natrophosphate, natroxalate, and a sodium aluminosilicate compound. At the elevated temperature and 8 M [Na+], appreciable insoluble solids (1.39 wt%) were present. Cooling to room temperature and dilution of the slurry from 8 M to 5 M [Na+] resulted in a slurry containing 0.8 wt% insoluble solids. The solids (natrophosphate, natroxalate, sodium aluminum silicate, and a hydrated sodium phosphate) were relatively stable and settled quickly. Filtration rates were in the range of those observed with iron-based simulated Hanford tank sludge simulants, e.g., 6 M [Na+] Hanford tank 241-AN-102, even though their chemical speciation is considerably different. Chemical cleaning of the crossflow filter was readily accomplished with acid. As this simulant formulation was based on an average composition of a wide range of feeds using an integrated computer model, this exact composition may never be observed. But the test conditions were selected to enable comparison to the model to enable improving its chemical prediction capability.

  18. Glassy slags as novel waste forms for remediating mixed wastes with high metal contents

    SciTech Connect (OSTI)

    Feng, X.; Wronkiewicz, D.J.; Bates, J.K.; Brown, N.R.; Buck, E.C.; Gong, M.; Ebert, W.L.

    1994-03-01T23:59:59.000Z

    Argonne National Laboratory (ANL) is developing a glassy slag final waste form for the remediation of low-level radioactive and mixed wastes with high metal contents. This waste form is composed of various crystalline and metal oxide phases embedded in a silicate glass phase. This work indicates that glassy slag shows promise as final waste form because (1) it has similar or better chemical durability than high-level nuclear waste (HLW) glasses, (2) it can incorporate large amounts of metal wastes, (3) it can incorporate waste streams having low contents of flux components (boron and alkalis), (4) it has less stringent processing requirements (e.g., viscosity and electric conductivity) than glass waste forms, (5) its production can require little or no purchased additives, which can result in greater reduction in waste volume and overall treatment costs. By using glassy slag waste forms, minimum additive waste stabilization approach can be applied to a much wider range of waste streams than those amenable only to glass waste forms.

  19. Remote Handling Equipment for a High-Level Waste Waste Package Closure System

    SciTech Connect (OSTI)

    Kevin M. Croft; Scott M. Allen; Mark W. Borland

    2006-04-01T23:59:59.000Z

    High-level waste will be placed in sealed waste packages inside a shielded closure cell. The Idaho National Laboratory (INL) has designed a system for closing the waste packages including all cell interior equipment and support systems. This paper discusses the material handling aspects of the equipment used and operations that will take place as part of the waste package closure operations. Prior to construction, the cell and support system will be assembled in a full-scale mockup at INL.

  20. Handbook of high-level radioactive waste transportation

    SciTech Connect (OSTI)

    Sattler, L.R.

    1992-10-01T23:59:59.000Z

    The High-Level Radioactive Waste Transportation Handbook serves as a reference to which state officials and members of the general public may turn for information on radioactive waste transportation and on the federal government`s system for transporting this waste under the Civilian Radioactive Waste Management Program. The Handbook condenses and updates information contained in the Midwestern High-Level Radioactive Waste Transportation Primer. It is intended primarily to assist legislators who, in the future, may be called upon to enact legislation pertaining to the transportation of radioactive waste through their jurisdictions. The Handbook is divided into two sections. The first section places the federal government`s program for transporting radioactive waste in context. It provides background information on nuclear waste production in the United States and traces the emergence of federal policy for disposing of radioactive waste. The second section covers the history of radioactive waste transportation; summarizes major pieces of legislation pertaining to the transportation of radioactive waste; and provides an overview of the radioactive waste transportation program developed by the US Department of Energy (DOE). To supplement this information, a summary of pertinent federal and state legislation and a glossary of terms are included as appendices, as is a list of publications produced by the Midwestern Office of The Council of State Governments (CSG-MW) as part of the Midwestern High-Level Radioactive Waste Transportation Project.

  1. High-Level waste process and product data annotated bibliography

    SciTech Connect (OSTI)

    Stegen, G.E.

    1996-02-13T23:59:59.000Z

    The objective of this document is to provide information on available issued documents that will assist interested parties in finding available data on high-level waste and transuranic waste feed compositions, properties, behavior in candidate processing operations, and behavior on candidate product glasses made from those wastes. This initial compilation is only a partial list of available references.

  2. RADIOACTIVE DEMONSTRATIONS OF FLUIDIZED BED STEAM REFORMING AS A SUPPLEMENTARY TREATMENT FOR HANFORD'S LOW ACTIVITY WASTE AND SECONDARY WASTES

    SciTech Connect (OSTI)

    Jantzen, C.; Crawford, C.; Cozzi, A.; Bannochie, C.; Burket, P.; Daniel, G.

    2011-02-24T23:59:59.000Z

    The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP's LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as Cs-137, I-129, Tc-99, Cl, F, and SO4 that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap. The current waste disposal path for the WTP-SW is to recycle it to the supplemental LAW treatment to avoid a large steady state accumulation in the pretreatment-vitrification loop. Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750 C) continuous method by which LAW and/or WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product is being investigated to prevent dispersion during transport or burial/storage but is not necessary for performance. A Benchscale Steam Reformer (BSR) was designed and constructed at the Savannah River National Laboratory (SRNL) to treat actual radioactive wastes to confirm the findings of the non-radioactive FBSR pilot scale tests and to qualify the waste form for applications at Hanford. Radioactive testing commenced in 2010 with a demonstration of Hanford's WTP-SW where Savannah River Site (SRS) High Level Waste (HLW) secondary waste from the Defense Waste Processing Facility (DWPF) was shimmed with a mixture of I-125/129 and Tc-99 to chemically resemble WTP-SW. Ninety six grams of radioactive product were made for testing. The second campaign commenced using SRS LAW chemically trimmed to look like Hanford's LAW. Six hundred grams of radioactive product were made for extensive testing and comparison to the non-radioactive pilot scale tests. The same mineral phases were found in the radioactive and non-radioactive testing.

  3. Radioactive Demonstrations Of Fluidized Bed Steam Reforming (FBSR) With Hanford Low Activity Wastes

    SciTech Connect (OSTI)

    Jantzen, C. M.; Crawford, C. L.; Burket, P. R.; Bannochie, C. J.; Daniel, W. G.; Nash, C. A.; Cozzi, A. D.; Herman, C. C.

    2012-10-22T23:59:59.000Z

    Several supplemental technologies for treating and immobilizing Hanford low activity waste (LAW) are being evaluated. One immobilization technology being considered is Fluidized Bed Steam Reforming (FBSR) which offers a low temperature (700-750?C) continuous method by which wastes high in organics, nitrates, sulfates/sulfides, or other aqueous components may be processed into a crystalline ceramic (mineral) waste form. The granular waste form produced by co-processing the waste with kaolin clay has been shown to be as durable as LAW glass. The FBSR granular product will be monolithed into a final waste form. The granular component is composed of insoluble sodium aluminosilicate (NAS) feldspathoid minerals such as sodalite. Production of the FBSR mineral product has been demonstrated both at the industrial, engineering, pilot, and laboratory scales on simulants. Radioactive testing at SRNL commenced in late 2010 to demonstrate the technology on radioactive LAW streams which is the focus of this study.

  4. Development of Ceramic Waste Forms for High-Level Nuclear Waste Over the Last 30 Years

    SciTech Connect (OSTI)

    Vance, Eric [Institute of Materials and Engineering Science, Australian Nuclear Science and Technology Organisation, New Illawarra Road, Menai, NSW, 2234 (Australia)

    2007-07-01T23:59:59.000Z

    Many types of ceramics have been put forward for immobilisation of high-level waste (HLW) from reprocessing of nuclear power plant fuel or weapons production. After describing some historical aspects of waste form research, the essential features of the chemical design and processing of these different ceramic types will be discussed briefly. Given acceptable laboratory and long-term predicted performance based on appropriately rigorous chemical design, the important processing parameters are mostly waste loading, waste throughput, footprint, offgas control/minimization, and the need for secondary waste treatment. It is concluded that the 'problem of high-level nuclear waste' is largely solved from a technical point of view, within the current regulatory framework, and that the main remaining question is which technical disposition method is optimum for a given waste. (author)

  5. Summary - System Planning for Low-Activity Waste Treatment at...

    Office of Environmental Management (EM)

    Management (DOE-EM) External Technical Review of System Planning for Low-Activity Waste Treatment at Hanford Why DOE-EM Did This Review Construction of the facilities of...

  6. Towards Increased Waste Loading in High Level Waste Glasses: Developing a Better Understanding of Crystallization Behavior

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

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

    2014-01-01T23:59:59.000Z

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JCHM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these ''troublesome'' waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating. The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (with higher Al2O3). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group.

  7. Towards Increased Waste Loading in High Level Waste Glasses: Developing a Better Understanding of Crystallization Behavior

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Marra, James C.; Kim, Dong -Sang

    2014-01-01T23:59:59.000Z

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JCHM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these ''troublesome'' waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advancedmore »glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating. The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (with higher Al2O3). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group.« less

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

    SciTech Connect (OSTI)

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

    2009-12-30T23:59:59.000Z

    This Test Plan describes work to support the development and testing of high waste loading glass formulations that achieve high glass melting rates for Hanford high aluminum high level waste (HLW). In particular, the present testing is designed to evaluate the effect of using low activity waste (LAW) waste streams as a source of sodium in place ofchemical additives, sugar or cellulose as a reductant, boehmite as an aluminum source, and further enhancements to waste processing rate while meeting all processing and product quality requirements. The work will include preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM 100) tests designed to examine the effects of enhanced glass formulations, glass processing temperature, incorporation of the LAW waste stream as a sodium source, type of organic reductant, and feed solids content on waste processing rate and product quality. Also included is a confirmatory test on the HLW Pilot Melter (DM1200) with a composition selected from those tested on the DM100. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy's (DOE's) Office of River Protection (ORP) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same waste composition. This Test Plan is prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is about 12,500. This estimate is based upon the inventory ofthe tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat transfer and glass melting rate. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of {approx}1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150 C and by increasing the waste loading in the glass product Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage. The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP contract requirements. The WTP's overall mission will require the immobilization oftank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in waste-loading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.

  9. Reportable Nuclide Criteria for ORNL Radioactive Waste Management Activities - 13005

    SciTech Connect (OSTI)

    McDowell, Kip; Forrester, Tim [Oak Ridge National Laboratory, PO Box 2008 MS-6322, Oak Ridge, TN 37831 (United States)] [Oak Ridge National Laboratory, PO Box 2008 MS-6322, Oak Ridge, TN 37831 (United States); Saunders, Mark [Fairfield Services Group, PO Box 31468, KNOxville, TN 37930 (United States)] [Fairfield Services Group, PO Box 31468, KNOxville, TN 37930 (United States)

    2013-07-01T23:59:59.000Z

    The U.S. Department of Energy's Oak Ridge National Laboratory (ORNL) in Oak Ridge, Tennessee generates numerous radioactive waste streams. Many of those streams contain a large number of radionuclides with an extremely broad range of concentrations. To feasibly manage the radionuclide information, ORNL developed reportable nuclide criteria to distinguish between those nuclides in a waste stream that require waste tracking versus those nuclides of such minimal activity that do not require tracking. The criteria include tracking thresholds drawn from ORNL onsite management requirements, transportation requirements, and relevant treatment and disposal facility acceptance criteria. As a management practice, ORNL maintains waste tracking on a nuclide in a specific waste stream if it exceeds any of the reportable nuclide criteria. Nuclides in a specific waste stream that screen out as non-reportable under all these criteria may be dropped from ORNL waste tracking. The benefit of these criteria is to ensure that nuclides in a waste stream with activities which meaningfully affect safety and compliance are tracked, while documenting the basis for removing certain isotopes from further consideration. (authors)

  10. Progress of the High Level Waste Program at the Defense Waste Processing Facility - 13178

    SciTech Connect (OSTI)

    Bricker, Jonathan M.; Fellinger, Terri L.; Staub, Aaron V.; Ray, Jeff W.; Iaukea, John F. [Savannah River Remediation, Aiken, South Carolina, 29808 (United States)] [Savannah River Remediation, Aiken, South Carolina, 29808 (United States)

    2013-07-01T23:59:59.000Z

    The Defense Waste Processing Facility at the Savannah River Site treats and immobilizes High Level Waste into a durable borosilicate glass for safe, permanent storage. The High Level Waste program significantly reduces environmental risks associated with the storage of radioactive waste from legacy efforts to separate fissionable nuclear material from irradiated targets and fuels. In an effort to support the disposition of radioactive waste and accelerate tank closure at the Savannah River Site, the Defense Waste Processing Facility recently implemented facility and flowsheet modifications to improve production by 25%. These improvements, while low in cost, translated to record facility production in fiscal years 2011 and 2012. In addition, significant progress has been accomplished on longer term projects aimed at simplifying and expanding the flexibility of the existing flowsheet in order to accommodate future processing needs and goals. (authors)

  11. Stability of High-Level Waste Forms

    SciTech Connect (OSTI)

    Besmann, Theodore M.; Vienna, John D.

    2006-11-10T23:59:59.000Z

    The objective of the proposed effort is to use a new approach to develop solution models of complex waste glass systems and spent fuel that are predictive with regard to composition, phase separation, and volatility. The effort will also yield thermodynamic values for waste components that are fundamentally required for corrosion models used to predict the leaching/corrosion behavior for waste glass and spent fuel material. This basic information and understanding of chemical behavior can subsequently be used directly in computational models of leaching and transport in geologic media, in designing and engineering waste forms and barrier systems, and in prediction of chemical interactions.

  12. High-level waste at Hanford: Potential for waste loading maximization

    SciTech Connect (OSTI)

    Hrma, P.; Bailey, A.W. [Pacific Northwest Lab., Richland, WA (United States)

    1995-12-31T23:59:59.000Z

    The loading of Hanford nuclear waste in borosilicate glass is limited by phase-related phenomena, such as crystallization or formation of immiscible liquids, and by the breakdown of the glass structure due to an excessive concentration of modifiers. The phase-related phenomena cause both processing and product quality problems. The deterioration of the product durability determines the ultimate waste loading limit if all processing problems are resolved. Concrete examples and mass-balance based calculations show that a substantial potential exists for increasing waste loading of high-level wastes that contain a large fraction of refractory components.

  13. DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

    SciTech Connect (OSTI)

    G. Radulesscu; J.S. Tang

    2000-06-07T23:59:59.000Z

    The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M&O 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M&Q 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M&O 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable canisters. The intended use of this analysis is to support Site Recommendation reports and to assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the Development Plan ''Design Analysis for the Defense High-Level Waste Disposal Container'' (CRWMS M&O 2000c) with no deviations from the plan.

  14. active nuclear wastes: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    active nuclear wastes First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Calculations of Induced Activity...

  15. MELT RATE ENHANCEMENT FOR HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASS FORMULATION FINAL REPORT 08R1360-1

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; KOT W; PEGG IL; JOSEPH I; BARDAKCI T; GAN H; GONG W; CHAUDHURI M

    2010-01-04T23:59:59.000Z

    This report describes the development and testing of new glass formulations for high aluminum waste streams that achieve high waste loadings while maintaining high processing rates. The testing was based on the compositions of Hanford High Level Waste (HLW) with limiting concentrations of aluminum specified by the Office of River Protection (ORP). The testing identified glass formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts and small scale melt rate screening tests. The results were used to select compositions for subsequent testing in a DuraMelter 100 (DM100) system. These tests were used to determine processing rates for the selected formulations as well as to examine the effects of increased glass processing temperature, and the form of aluminum in the waste simulant. Finally, one of the formulations was selected for large-scale confirmatory testing on the HLW Pilot Melter (DM1200), which is a one third scale prototype of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW melter and off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy (DOE) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same high-aluminum waste composition used in the present work and other Hanford HLW compositions. The scope of this study was outlined in a Test Plan that was prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the WTP is about 13,500 (equivalent to 40,500 MT glass). This estimate is based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat transfer and glass melting rate. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of {approx}1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150 C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage. The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP Contract terms. The WTP's overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in wasteloading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.

  16. TWRS retrieval and storage mission, immobilized low-activity waste disposal plan

    SciTech Connect (OSTI)

    Shade, J.W.

    1998-01-07T23:59:59.000Z

    The TWRS mission is to store, treat, and immobilize highly radioactive Hanford waste (current and future tank waste and the encapsulated cesium and strontium) in a safe, environmentally sound, and cost-effective manner (TWRS JMN Justification for mission need). The mission includes retrieval, pretreatment, immobilization, interim storage and disposal, and tank closure. As part of this mission, DOE has established the TWRS Office to manage all Hanford Site tank waste activities. The TWRS program has identified the need to store, treat, immobilize, and dispose of the highly radioactive Hanford Site tank waste and encapsulated cesium and strontium materials in an environmentally sound, safe, and cost-effective manner. To support environmental remediation and restoration at the Hanford Site a two-phase approach to using private contractors to treat and immobilize the low-activity and high-level waste currently stored in underground tanks is planned. The request for proposals (RFP) for the first phase of waste treatment and immobilization was issued in February 1996 (Wagoner 1996) and initial contracts for two private contractor teams led by British Nuclear Fuels Ltd. and Lockheed-Martin Advanced Environmental Services were signed in September 1996. Phase 1 is a proof-of-concept and commercial demonstration effort to demonstrate the technical and business feasibility of using private facilities to treat Hanford Site waste, maintain radiological, nuclear, process, and occupational safety; and maintain environmental protection and compliance while reducing lifecycle costs and waste treatment times. Phase 1 production of ILAW is planned to begin in June 2002 and could treat up to about 13 percent of the waste. Phase 1 production is expected to be completed in 2007 for minimum order quantities or 2011 for maximum order quantities. Phase 2 is a full-scale production effort that will begin after Phase 1 and treat and immobilize most of the waste. Phase 2 production is expected to be completed in 2025. DOE will supply the feed to the private contractors and will receive the ILAW product from the private treatment facilities during Phase 1. For Phase 2, retrieval and feed delivery, as well as waste treatment and immobilization, will be done by private contractors. DOE will pay the private contractors for each ILAW package that meets the product specifications as stated in the RFP or subsequently negotiated. Acceptance of immobilized waste will be based on private contractor activities to qualify, verify, document, and certify the product and DOE activities to audit, review, inspect, and evaluate the treatment and immobilization process and products. The acceptance process is expected to result in ILAW product packages certified for transport and disposal at the Hanford Site safely and in compliance with environmental regulations.

  17. Minor component study for simulated high-level nuclear waste glasses (Draft)

    SciTech Connect (OSTI)

    Li, H.; Langowskim, M.H.; Hrma, P.R.; Schweiger, M.J.; Vienna, J.D.; Smith, D.E.

    1996-02-01T23:59:59.000Z

    Hanford Site single-shell tank (SSI) and double-shell tank (DSI) wastes are planned to be separated into low activity (or low-level waste, LLW) and high activity (or high-level waste, HLW) fractions, and to be vitrified for disposal. Formulation of HLW glass must comply with glass processibility and durability requirements, including constraints on melt viscosity, electrical conductivity, liquidus temperature, tendency for phase segregation on the molten glass surface, and chemical durability of the final waste form. A wide variety of HLW compositions are expected to be vitrified. In addition these wastes will likely vary in composition from current estimates. High concentrations of certain troublesome components, such as sulfate, phosphate, and chrome, raise concerns about their potential hinderance to the waste vitrification process. For example, phosphate segregation in the cold cap (the layer of feed on top of the glass melt) in a Joule-heated melter may inhibit the melting process (Bunnell, 1988). This has been reported during a pilot-scale ceramic melter run, PSCM-19, (Perez, 1985). Molten salt segregation of either sulfate or chromate is also hazardous to the waste vitrification process. Excessive (Cr, Fe, Mn, Ni) spinel crystal formation in molten glass can also be detrimental to melter operation.

  18. Cementitious Stabilization of Mixed Wastes with High Salt Loadings

    SciTech Connect (OSTI)

    Spence, R.D.; Burgess, M.W.; Fedorov, V.V.; Downing, D.J.

    1999-04-01T23:59:59.000Z

    Salt loadings approaching 50 wt % were tolerated in cementitious waste forms that still met leach and strength criteria, addressing a Technology Deficiency of low salt loadings previously identified by the Mixed Waste Focus Area. A statistical design quantified the effect of different stabilizing ingredients and salt loading on performance at lower loadings, allowing selection of the more effective ingredients for studying the higher salt loadings. In general, the final waste form needed to consist of 25 wt % of the dry stabilizing ingredients to meet the criteria used and 25 wt % water to form a workable paste, leaving 50 wt % for waste solids. The salt loading depends on the salt content of the waste solids but could be as high as 50 wt % if all the waste solids are salt.

  19. High Level Waste Management Division High-Level Waste System Plan

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC) EnvironmentalGyroSolé(tm) Harmonicbet WhenHiggs Boson May| ArgonneHigh Level Waste.6

  20. Analysis of alternatives for immobilized low activity waste disposal

    SciTech Connect (OSTI)

    Burbank, D.A.

    1997-10-28T23:59:59.000Z

    This report presents a study of alternative disposal system architectures and implementation strategies to provide onsite near-surface disposal capacity to receive the immobilized low-activity waste produced by the private vendors. The analysis shows that a flexible unit strategy that provides a suite of design solutions tailored to the characteristics of the immobilized low-activity waste will provide a disposal system that best meets the program goals of reducing the environmental, health, and safety impacts; meeting the schedule milestones; and minimizing the life-cycle cost of the program.

  1. High-Efficiency Quantum-Well Thermoelectrics for Waste Heat Power...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    High-Efficiency Quantum-Well Thermoelectrics for Waste Heat Power Generation High-Efficiency Quantum-Well Thermoelectrics for Waste Heat Power Generation 2005 Diesel Engine...

  2. High-Temperature Components for Rankine-Cycle-Based Waste Heat...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    High-Temperature Components for Rankine-Cycle-Based Waste Heat Recovery Systems on Combustion Engines High-Temperature Components for Rankine-Cycle-Based Waste Heat Recovery...

  3. Glass Formulations for Immobilizing Hanford Low-Activity Wastes

    SciTech Connect (OSTI)

    Kim, Dong-Sang; Elliott, Michael L.; Smith, Harry D.; Bagaasen, Larry M.; Hrma, Pavel R.

    2006-02-28T23:59:59.000Z

    Researchers at Pacific Northwest National Laboratory (PNNL) are developing and testing glasses for immobilizing low-activity wastes (LAW) for the full Hanford mission. PNNL is performing testing for low-activity waste glasses for both the Hanford Waste Treatment Plant (WTP) and the Bulk Vitrification Plant. The objective of this work is to increase the waste content of the glasses and ultimately increase the waste throughput of the LAW vitrification plants. This paper focuses on PNNL’s development and testing of glasses for the Bulk Vitrification process. Bulk Vitrification was selected as a potential supplemental treatment to accelerate the cleanup of LAW at Hanford. Also known as In-Container Vitrification™ (ICV™), the Bulk Vitrification process combines soil, LAW, and chemical amendments; dries the mixture; and then vitrifies the material in a batch process in a refractory lined box. The process was developed by AMEC Earth and Environmental, Inc. (AMEC). Working with AMEC, PNNL developed a glass formulation that could incorporate a broad range of Hanford LAW. The initial glass development involved a “nominal” waste composition, and a baseline glass was formulated and tested at crucible, engineering, and full scales. The performance of the baseline glass was then verified using a battery of laboratory tests as well as engineering-scale and full-scale ICV™ tests. Future testing is planned for optimizing the glass waste loading and qualifying a broader range of waste streams for treatment in the Bulk Vitrification process. This paper reviews the glass development and qualification process completed to date. This includes several series of crucible studies as well as confirmation testing at engineering-scale and full-scale. This formulation paper complements information presented by AMEC in an ICV™ processing paper.

  4. Active Waste Materials Corrosion and Decontamination Tests

    SciTech Connect (OSTI)

    MJ Danielson; MR Elmore; SG Pitman

    2000-08-15T23:59:59.000Z

    Stainless steel alloys, 304L and 316L, were corrosion tested in representative radioactive samples of three actual Hanford tank waste solutions (Tanks AW-101, C-104, AN-107). Both the 304L and 316L exhibited good corrosion performance when immersed in boiling waste solutions. The maximum general corrosion rate was 0.015 mm/y (0.60 mils per year). Generally, the 304L had a slightly higher rate than the 316L. No localized attack was observed after 122 days of testing in the liquid phase, liquid/vapor phase, or vapor phase. Radioactive plate-out decontamination tests indicated that a 24-hour exposure to 1 {und M} HNO{sub 3} could remove about 99% of the radioactive components in the metal film when exposed to the C-104 and AN-107 solutions. The decontamination results are less certain for the AW-101 solution, since the initial contamination readings exceeded the capacity of the meter used for this test.

  5. RETENTION OF SULFATE IN HIGH LEVEL RADIOACTIVE WASTE GLASS

    SciTech Connect (OSTI)

    Fox, K.

    2010-09-07T23:59:59.000Z

    High level radioactive wastes are being vitrified at the Savannah River Site for long term disposal. Many of the wastes contain sulfate at concentrations that can be difficult to retain in borosilicate glass. This study involves efforts to optimize the composition of a glass frit for combination with the waste to improve sulfate retention while meeting other process and product performance constraints. The fabrication and characterization of several series of simulated waste glasses are described. The experiments are detailed chronologically, to provide insight into part of the engineering studies used in developing frit compositions for an operating high level waste vitrification facility. The results lead to the recommendation of a specific frit composition and a concentration limit for sulfate in the glass for the next batch of sludge to be processed at Savannah River.

  6. I-NERI-2007-004-K, DEVELOPMENT AND CHARACTERIZATION OF NEW HIGH-LEVEL WASTE FORMS FOR ACHIEVING WASTE MINIMIZATION FROM PYROPROCESSING

    SciTech Connect (OSTI)

    S.M. Frank

    2011-09-01T23:59:59.000Z

    Work describe in this report represents the final year activities for the 3-year International Nuclear Energy Research Initiative (I-NERI) project: Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing. Used electrorefiner salt that contained actinide chlorides and was highly loaded with surrogate fission products was processed into three candidate waste forms. The first waste form, a high-loaded ceramic waste form is a variant to the CWF produced during the treatment of Experimental Breeder Reactor-II used fuel at the Idaho National Laboratory (INL). The two other waste forms were developed by researchers at the Korean Atomic Energy Research Institute (KAERI). These materials are based on a silica-alumina-phosphate matrix and a zinc/titanium oxide matrix. The proposed waste forms, and the processes to fabricate them, were designed to immobilize spent electrorefiner chloride salts containing alkali, alkaline earth, lanthanide, and halide fission products that accumulate in the salt during the processing of used nuclear fuel. This aspect of the I-NERI project was to demonstrate 'hot cell' fabrication and characterization of the proposed waste forms. The outline of the report includes the processing of the spent electrorefiner salt and the fabrication of each of the three waste forms. Also described is the characterization of the waste forms, and chemical durability testing of the material. While waste form fabrication and sample preparation for characterization must be accomplished in a radiological hot cell facility due to hazardous radioactivity levels, smaller quantities of each waste form were removed from the hot cell to perform various analyses. Characterization included density measurement, elemental analysis, x-ray diffraction, scanning electron microscopy and the Product Consistency Test, which is a leaching method to measure chemical durability. Favorable results from this demonstration project will provide additional options for fission product immobilization and waste management associated the electrochemical/pyrometallurgical processing of used nuclear fuel.

  7. Laboratory Evaporation Testing Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

    SciTech Connect (OSTI)

    Adamson, Duane J.; Nash, Charles A.; McCabe, Daniel J.; Crawford, Charles L.; Wilmarth, William R.

    2014-01-27T23:59:59.000Z

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream, LAW Off-Gas Condensate, from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of canistered glass waste forms. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to be within acceptable concentration ranges in the LAW glass. Diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task examines the impact of potential future disposition of this stream in the Hanford tank farms, and investigates auxiliary evaporation to enable another disposition path. Unless an auxiliary evaporator is used, returning the stream to the tank farms would require evaporation in the 242-A evaporator. This stream is expected to be unusual because it will be very high in corrosive species that are volatile in the melter (chloride, fluoride, sulfur), will have high ammonia, and will contain carryover particulates of glass-former chemicals. These species have potential to cause corrosion of tanks and equipment, precipitation of solids, release of ammonia gas vapors, and scale in the tank farm evaporator. Routing this stream to the tank farms does not permanently divert it from recycling into the WTP, only temporarily stores it prior to reprocessing. Testing is normally performed to demonstrate acceptable conditions and limits for these compounds in wastes sent to the tank farms. The primary parameter of this phase of the test program was measuring the formation of solids during evaporation in order to assess the compatibility of the stream with the evaporator and transfer and storage equipment. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW facility melter offgas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet, and, thus, the composition will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. This report discusses results of evaporation testing of the simulant. Two conditions were tested, one with the simulant at near neutral pH, and a second at alkaline pH. The neutral pH test is comparable to the conditions in the Hanford Effluent Treatment Facility (ETF) evaporator, although that evaporator operates at near atmospheric pressure and tests were done under vacuum. For the alkaline test, the target pH was based on the tank farm corrosion control program requirements, and the test protocol and equipment was comparable to that used for routine evaluation of feed compatibility studies for the 242-A evaporator. One of the

  8. Qualification of Innovative High Level Waste Pipeline Unplugging Technologies

    SciTech Connect (OSTI)

    McDaniel, D.; Gokaltun, S.; Varona, J.; Awwad, A.; Roelant, D.; Srivastava, R. [Applied Research Center, Florida International University, Miami, FL (United States)

    2008-07-01T23:59:59.000Z

    In the past, some of the pipelines have plugged during high level waste (HLW) transfers resulting in schedule delays and increased costs. Furthermore, pipeline plugging has been cited by the 'best and brightest' technical review as one of the major issues that can result in unplanned outages at the Waste Treatment Plant causing inconsistent operation. As the DOE moves toward a more active high level waste retrieval, the site engineers will be faced with increasing cross-site pipeline waste slurry transfers that will result in increased probability of a pipeline getting plugged. Hence, availability of a pipeline unplugging tool/technology is crucial to ensure smooth operation of the waste transfers and in ensuring tank farm cleanup milestones are met. FIU had earlier tested and evaluated various unplugging technologies through an industry call. Based on mockup testing, two technologies were identified that could withstand the rigors of operation in a radioactive environment and with the ability to handle sharp 90 elbows. We present results of the second phase of detailed testing and evaluation of pipeline unplugging technologies and the objective is to qualify these pipeline unplugging technologies for subsequent deployment at a DOE facility. The current phase of testing and qualification comprises of a heavily instrumented 3-inch diameter (full-scale) pipeline facilitating extensive data acquisition for design optimization and performance evaluation, as it applies to three types of plugs atypical of the DOE HLW waste. Furthermore, the data from testing at three different lengths of pipe in conjunction with the physics of the process will assist in modeling the unplugging phenomenon that will then be used to scale-up process parameters and system variables for longer and site typical pipe lengths, which can extend as much as up to 19,000 ft. Detailed information resulting from the testing will provide the DOE end-user with sufficient data and understanding of the technology, and its limitations to aid in the benefit-cost analysis for management decision whether to deploy the technology or to abandon the pipeline as has been done in the past. In conclusion: The ultimate objective of this study is to qualify NuVision's unplugging technology for use at Hanford. Experimental testing has been conducted using three pipeline lengths and three types of blockages. Erosion rates have been obtained and pressure data is being analyzed. An amplification of the inlet pressure has been observed along the pipeline and is the key to determining up to what pipe lengths the technology can be used without surpassing the site pressure limit. In addition, we will attempt to establish what the expected unplugging rates will be at the longer pipe lengths for each of the three blockages tested. Detailed information resulting from the testing will provide the DOE end-user with sufficient data and understanding of the technology, and its limitations so that management decisions can be made whether the technology has a reasonable chance to successfully unplug a pipeline, such as a cross site transfer line or process transfer pipeline at the Waste Treatment Plant. (authors)

  9. The High-Level Radioactive Waste Act (Manitoba, Canada)

    Broader source: Energy.gov [DOE]

    Manitoba bars the storage of high-level radioactive wastes from spent nuclear fuel, not intended for research purposes, that was produced at a nuclear facility or in a nuclear reactor outside the...

  10. Characterization of selected waste tanks from the active LLLW system

    SciTech Connect (OSTI)

    Keller, J.M.; Giaquinto, J.M.; Griest, W.H.

    1996-08-01T23:59:59.000Z

    From September 1989 through January of 1990, there was a major effort to sample and analyze the Active Liquid-Low Level Waste (LLLW) tanks at ORNL which include the Melton Valley Storage Tanks (MVST) and the Bethel Valley Evaporator Service Tanks (BVEST). The purpose of this report is to summarize additional analytical data collected from some of the active waste tanks from November 1993 through February 1996. The analytical data for this report was collected for several unrelated projects which had different data requirements. The overall analyte list was similar for these projects and the level of quality assurance was the same for all work reported. the new data includes isotopic ratios for uranium and plutonium and an evaluation of the denature ratios to address criticality concerns. Also, radionuclides not previously measured in these waste tanks, including 99Tc and 237Np, are provided in this report.

  11. High-Level Waste System Process Interface Description

    SciTech Connect (OSTI)

    d'Entremont, P.D.

    1999-01-14T23:59:59.000Z

    The High-Level Waste System is a set of six different processes interconnected by pipelines. These processes function as one large treatment plant that receives, stores, and treats high-level wastes from various generators at SRS and converts them into forms suitable for final disposal. The three major forms are borosilicate glass, which will be eventually disposed of in a Federal Repository, Saltstone to be buried on site, and treated water effluent that is released to the environment.

  12. Progress in resolving Hanford Site high-level waste tank safety issues

    SciTech Connect (OSTI)

    Babad, H.; Eberlein, S.J.; Johnson, G.D.; Meacham, J.E.; Osborne, J.W.; Payne, M.A.; Turner, D.A.

    1995-02-01T23:59:59.000Z

    Interim storage of alkaline, high-level radioactive waste, from two generations of spent fuel reprocessing and waste management activities, has resulted in the accumulation of 238 million liters of waste in Hanford Site single and double-shell tanks. Before the 1990`s, the stored waste was believed to be: (1) chemically unreactive under its existing storage conditions and plausible accident scenarios; and (2) chemically stable. This paradigm was proven incorrect when detailed evaluation of tank contents and behavior revealed a number of safety issues and that the waste was generating flammable and noxious gases. In 1990, the Waste Tank Safety Program was formed to focus on identifying safety issues and resolving the ferrocyanide, flammable gas, organic, high heat, noxious vapor, and criticality issues. The tanks of concern were placed on Watch Lists by safety issue. This paper summarizes recent progress toward resolving Hanford Site high-level radioactive waste tank safety issues, including modeling, and analyses, laboratory experiments, monitoring upgrades, mitigation equipment, and developing a strategy to screen tanks for safety issues.

  13. Water borne transport of high level nuclear waste in very deep borehole disposal of high level nuclear waste

    E-Print Network [OSTI]

    Cabeche, Dion Tunick

    2011-01-01T23:59:59.000Z

    The purpose of this report is to examine the feasibility of the very deep borehole experiment and to determine if it is a reasonable method of storing high level nuclear waste for an extended period of time. The objective ...

  14. EIS-0062: Double-Shell Tanks for Defense High Level Waste Storage, Savannah River Site, Aiken, SC

    Broader source: Energy.gov [DOE]

    This EIS analyzes the impacts of the various design alternatives for the construction of fourteen 1.3 million gallon high-activity radioactive waste tanks. The EIS further evaluates the effects of these alternative designs on tank durability, on the ease of waste retrieval from such tanks, and the choice of technology and timing for long-term storage or disposal of the wastes.

  15. Alternatives Generation and Analysis for Heat Removal from High Level Waste Tanks

    SciTech Connect (OSTI)

    WILLIS, W.L.

    2000-06-15T23:59:59.000Z

    This document addresses the preferred combination of design and operational configurations to provide heat removal from high-level waste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. An interim decision for the preferred method to remove the heat from the high-level waste tanks during waste feed delivery operations is presented herein.

  16. Tank waste remediation system high-level waste feed processability assessment report

    SciTech Connect (OSTI)

    Lambert, S.L. [Westinghouse Hanford Co., Richland, WA (United States); Kim, D.S. [Pacific Northwest Lab., Richland, WA (United States)

    1994-12-01T23:59:59.000Z

    This study evaluates the effect of feed composition on the performance of the high-level vitrification process. It is assumed in this study that the tank wastes are retrieved and blended by tank farms, producing 12 different blends from the single-shell tank farms, two blends of double-shell tank waste, and a separately defined all-tank blend. This blending scenario was chosen only for evaluating the impact of composition on the volume of high- level waste glass produced. Special glass compositions were formulated for each waste blend based on glass property models and the properties of similar glasses. These glasses were formulated to meet the applicable viscosity, electrical conductivity, and liquidus temperature constraints for the identified candidate melters. Candidate melters in this study include the low-temperature stirred melter, which operates at 1050{degrees}C; the reference Hanford Waste Vitrification Plant liquid-fed ceramic melter, which operates at 1150{degrees}C; and the high-temperature, joule-heated melter and the cold-crucible melter, which operate over a temperature range of 1150{degrees}C to 1400{degrees}C. In the most conservative case, it is estimated that 61,000 MT of glass will be produced if the Site`s high-level wastes are retrieved by tank farms and processed in the reference joule-heated melter. If an all-tank blend was processed under the same conditions, the reference melter would produce 21,250 MT of glass. If cross-tank blending were used, it is anticipated that $2.0 billion could be saved in repository disposal costs (based on an average disposal cost of $217,000 per canister) by blending the S, SX, B, and T Tank Farm wastes with other wastes prior to vitrification. General blending among all the tank farms is expected to produce great potential benefit.

  17. Development of an Integrated Raman and Turbidity Fiber Optic Sensor for the In-Situ Analysis of High Level Nuclear Waste - 13532

    SciTech Connect (OSTI)

    Gasbarro, Christina; Bello, Job [EIC Laboratories, Inc., 111 Downey St., Norwood, MA, 02062 (United States)] [EIC Laboratories, Inc., 111 Downey St., Norwood, MA, 02062 (United States); Bryan, Samuel; Lines, Amanda; Levitskaia, Tatiana [Pacific Northwest National Laboratory, PO Box 999, Richland, WA, 99352 (United States)] [Pacific Northwest National Laboratory, PO Box 999, Richland, WA, 99352 (United States)

    2013-07-01T23:59:59.000Z

    Stored nuclear waste must be retrieved from storage, treated, separated into low- and high-level waste streams, and finally put into a disposal form that effectively encapsulates the waste and isolates it from the environment for a long period of time. Before waste retrieval can be done, waste composition needs to be characterized so that proper safety precautions can be implemented during the retrieval process. In addition, there is a need for active monitoring of the dynamic chemistry of the waste during storage since the waste composition can become highly corrosive. This work describes the development of a novel, integrated fiber optic Raman and light scattering probe for in situ use in nuclear waste solutions. The dual Raman and turbidity sensor provides simultaneous chemical identification of nuclear waste as well as information concerning the suspended particles in the waste using a common laser excitation source. (authors)

  18. Development of an Integrated Raman and Turbidity Fiber Optic Sensor for the In-Situ Analysis of High Level Nuclear Waste

    SciTech Connect (OSTI)

    Gasbarro, Christina; Bello, Job M.; Bryan, Samuel A.; Lines, Amanda M.; Levitskaia, Tatiana G.

    2013-02-24T23:59:59.000Z

    Stored nuclear waste must be retrieved from storage, treated, separated into low- and high-level waste streams, and finally put into a disposal form that effectively encapsulates the waste and isolates it from the environment for a long period of time. Before waste retrieval can be done, waste composition needs to be characterized so that proper safety precautions can be implemented during the retrieval process. In addition, there is a need for active monitoring of the dynamic chemistry of the waste during storage since the waste composition can become highly corrosive. This work describes the development of a novel, integrated fiber optic Raman and light scattering probe for in situ use in nuclear waste solutions. The dual Raman and turbidity sensor provides simultaneous chemical identification of nuclear waste as well as information concerning the suspended particles in the waste using a common laser excitation source.

  19. Disposal Activities and the Unique Waste Streams at the Nevada National Security Site (NNSS)

    SciTech Connect (OSTI)

    Arnold, P.

    2012-10-31T23:59:59.000Z

    This slide show documents waste disposal at the Nevada National Security Site. Topics covered include: radionuclide requirements for waste disposal; approved performance assessment (PA) for depleted uranium disposal; requirements; program approval; the Waste Acceptance Review Panel (WARP); description of the Radioactive Waste Acceptance Program (RWAP); facility evaluation; recent program accomplishments, nuclear facility safety changes; higher-activity waste stream disposal; and, large volume bulk waste streams.

  20. Demonstrating Reliable High Level Waste Slurry Sampling Techniques to Support Hanford Waste Processing

    SciTech Connect (OSTI)

    Kelly, Steven E.

    2013-11-11T23:59:59.000Z

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capability using simulated Hanford High-Level Waste (HL W) formulations. This work represents one of the remaining technical issues with the high-level waste treatment mission at Hanford. The TOC must demonstrate the ability to adequately mix and sample high-level waste feed to meet the WTP Waste Acceptance Criteria and Data Quality Objectives. The sampling method employed must support both TOC and WTP requirements. To facilitate information transfer between the two facilities the mixing and sampling demonstrations are led by the One System Integrated Project Team. The One System team, Waste Feed Delivery Mixing and Sampling Program, has developed a full scale sampling loop to demonstrate sampler capability. This paper discusses the full scale sampling loops ability to meet precision and accuracy requirements, including lessons learned during testing. Results of the testing showed that the Isolok(R) sampler chosen for implementation provides precise, repeatable results. The Isolok(R) sampler accuracy as tested did not meet test success criteria. Review of test data and the test platform following testing by a sampling expert identified several issues regarding the sampler used to provide reference material used to judge the Isolok's accuracy. Recommendations were made to obtain new data to evaluate the sampler's accuracy utilizing a reference sampler that follows good sampling protocol.

  1. Vitrification and testing of a Hanford high-level waste sample. Part 1: Glass fabrication, and chemical and radiochemical analysis

    SciTech Connect (OSTI)

    Hrma, Pavel R.; Crum, Jarrod V.; Bates, Derrick J.; Bredt, Paul; Greenwood, Lawrence R.; Smith, H D.

    2005-10-01T23:59:59.000Z

    The Hanford radioactive tank waste will be separated into low-activity waste and high-level waste that will both be vitrified into borosilicate glasses. To demonstrate the feasibility of vitrification and the durability of the high-level waste glass, a high-level waste sample from Tank AZ-101 was processed to glass in a hot cell and analyzed with respect to chemical composition, radionuclide content, waste loading, and the presence of crystalline phases and then tested for leachability. The glass was analyzed with inductively coupled plasma-atomic emission spectroscopy, inductively coupled plasma-mass spectrometry, ? energy spectrometry, ? spectrometry, and liquid scintillation counting. The WISE Uranium Project calculator was used to calculate the main sources of radioactivity to the year 3115. The observed crystallinity and the results of leachability testing of the glass will be reported in Part 2 of this paper.

  2. Spanish high level radioactive waste management system issues

    SciTech Connect (OSTI)

    Ulibarri, A.; Veganzones, A. [ENRESA, Madrid (Spain)

    1993-12-31T23:59:59.000Z

    The Empresa Nacional de Residuous Radiactivos, S.A. (ENRESA) was set up in 1984 as a state-owned limited liability company to be responsible for the management of all kinds of radioactive wastes in Spain. This paper provides an overview of the strategy and main lines of action stated in the third General Radioactive Waste Plan, currently in force, for the management of spent nuclear fuel and high-level wastes, as well as an outline of the main related projects, either being developed or foreseen. Aspects concerning the organizational structure, the economic and financing system and the international co-operational are also included.

  3. High level waste facilities -- Continuing operation or orderly shutdown

    SciTech Connect (OSTI)

    Decker, L.A.

    1998-04-01T23:59:59.000Z

    Two options for Environmental Impact Statement No action alternatives describe operation of the radioactive liquid waste facilities at the Idaho Chemical Processing Plant at the Idaho National Engineering and Environmental Laboratory. The first alternative describes continued operation of all facilities as planned and budgeted through 2020. Institutional control for 100 years would follow shutdown of operational facilities. Alternatively, the facilities would be shut down in an orderly fashion without completing planned activities. The facilities and associated operations are described. Remaining sodium bearing liquid waste will be converted to solid calcine in the New Waste Calcining Facility (NWCF) or will be left in the waste tanks. The calcine solids will be stored in the existing Calcine Solids Storage Facilities (CSSF). Regulatory and cost impacts are discussed.

  4. Preliminary Glass Development and Testing for In-Container Vitrification of Hanford Low-Activity Waste

    SciTech Connect (OSTI)

    Vienna, John D.; Kim, Dong-Sang; Schweiger, Michael J.; Hrma, Pavel R.; Matyas, Josef; Crum, Jarrod V.; Smith, Donald E.

    2004-01-01T23:59:59.000Z

    Roughly 50 million gallons of high-level waste (HLW) are stored at the Hanford site. This waste will be separated into HLW and low-activity waste (LAW) fractions and each fraction will be immobilized for final storage/disposal. The US Department of Energy (DOE) Office of River Protection (ORP) is constructing a Waste Treatment and Immobilization Plant (WTP) which will be capable of separating the waste, vitrifying the entire HLW fraction of the waste and vitrifying roughly 50% the LAW fraction. The remaining fraction of LAW will be immobilized by one of a number of possible technologies. ORP is currently evaluating options for LAW immobilization. One possible option is In-Container Vitrification (ICV) of the LAW. ICV is a technology developed by AMEC, GeoMelt Division, for treatment of hazardous, radioactive, and mixed wastes. The ICV process, as applied to Hanford LAW, includes the blending of liquid waste with additives (primarily composed of local soil) and drying to a granular state. The dried material is loaded into a refractory lined steel box and melted by passing a current through the material between two graphite electrodes. The box containing the molten waste/additive mixture is cooled, backfilled, and disposed of. The purpose of the study was to develop a glass composition suitable for the demonstration of ICV on Hanford LAW at full scale. Testing included crucible-scale tests with simulants and actual Hanford LAW. Following the crucible-scale tests, engineering-scale and large-scale melts were performed with LAW simulants. This paper discusses the formulation and testing of glass compositions for ICV of Hanford LAW at crucible scale. The results from process scale-up test are reported elsewhere.

  5. Effects of Globally Waste Disturbing Activities on Gas Generation, Retention, and Release in Hanford Waste Tanks

    SciTech Connect (OSTI)

    Stewart, Charles W.; Fountain, Matthew S.; Huckaby, James L.; Mahoney, Lenna A.; Meyer, Perry A.; Wells, Beric E.

    2005-08-02T23:59:59.000Z

    Various operations are authorized in Hanford single- and double-shell tanks that disturb all or a large fraction of the waste. These globally waste-disturbing activities have the potential to release a large fraction of the retained flammable gas and to affect future gas generation, retention, and release behavior. This report presents analyses of the expected flammable gas release mechanisms and the potential release rates and volumes resulting from these activities. The background of the flammable gas safety issue at Hanford is summarized, as is the current understanding of gas generation, retention, and release phenomena. Considerations for gas monitoring and assessment of the potential for changes in tank classification and steady-state flammability are given.

  6. CHARACTERIZATION OF HIGH PHOSPHATE RADIOACTIVE TANK WASTE AND SIMULANT DEVELOPMENT

    SciTech Connect (OSTI)

    Lumetta, Gregg J.; McNamara, Bruce K.; Buck, Edgar C.; Fiskum, Sandra K.; Snow, Lanee A.

    2009-10-15T23:59:59.000Z

    A sample of high-level radioactive tank waste was characterized to provide a basis for developing a waste simulant. The simulant is required for engineered-scaled testing of pretreatment processes in a non-radiological facility. The waste material examined was derived from the bismuth phosphate process, which was the first industrial process implemented to separate plutonium from irradiated nuclear fuel. The bismuth phosphate sludge is a complex mixture rich in bismuth, iron, sodium, phosphorus, silicon, and uranium. The form of phosphorus in this particular tank waste material is of specific importance because that is the primary component (other than water-soluble sodium salts) that must be removed from the high-level waste solids by pretreatment. This work shows unequivocally that the phosphorus present in this waste material is not present as bismuth phosphate. Rather, the phosphorus appears to be incorporated mostly into an amorphous iron(III) phosphate species. The bismuth in the sludge solids is best described as bismuth ferrite, BiFeO3. Infrared spectral data, microscopy, and thermal analysis data are presented to support these conclusions. The behavior of phosphorus during caustic leaching of the bismuth phosphate sludge solids is also discussed.

  7. Survey of waste package designs for disposal of high-level waste/spent fuel in selected foreign countries

    SciTech Connect (OSTI)

    Schneider, K.J.; Lakey, L.T.; Silviera, D.J.

    1989-09-01T23:59:59.000Z

    This report presents the results of a survey of the waste package strategies for seven western countries with active nuclear power programs that are pursuing disposal of spent nuclear fuel or high-level wastes in deep geologic rock formations. Information, current as of January 1989, is given on the leading waste package concepts for Belgium, Canada, France, Federal Republic of Germany, Sweden, Switzerland, and the United Kingdom. All but two of the countries surveyed (France and the UK) have developed design concepts for their repositories, but none of the countries has developed its final waste repository or package concept. Waste package concepts are under study in all the countries surveyed, except the UK. Most of the countries have not yet developed a reference concept and are considering several concepts. Most of the information presented in this report is for the current reference or leading concepts. All canisters for the wastes are cylindrical, and are made of metal (stainless steel, mild steel, titanium, or copper). The canister concepts have relatively thin walls, except those for spent fuel in Sweden and Germany. Diagrams are presented for the reference or leading concepts for canisters for the countries surveyed. The expected lifetimes of the conceptual canisters in their respective disposal environment are typically 500 to 1,000 years, with Sweden's copper canister expected to last as long as one million years. Overpack containers that would contain the canisters are being considered in some of the countries. All of the countries surveyed, except one (Germany) are currently planning to utilize a buffer material (typically bentonite) surrounding the disposal package in the repository. Most of the countries surveyed plan to limit the maximum temperature in the buffer material to about 100{degree}C. 52 refs., 9 figs.

  8. International program to study subseabed disposal of high-level radioactive wastes

    SciTech Connect (OSTI)

    Carlin, E.M.; Hinga, K.R.; Knauss, J.A.

    1984-01-01T23:59:59.000Z

    This report provides an overview of the international program to study seabed disposal of nuclear wastes. Its purpose is to inform legislators, other policy makers, and the general public as to the history of the program, technological requirements necessary for feasibility assessment, legal questions involved, international coordination of research, national policies, and research and development activities. Each of these major aspects of the program is presented in a separate section. The objective of seabed burial, similar to its continental counterparts, is to contain and to isolate the wastes. The subseabed option should not be confuesed with past practices of ocean dumping which have introduced wastes into ocean waters. Seabed disposal refers to the emplacement of solidified high-level radioactive waste (with or without reprocessing) in certain geologically stable sediments of the deep ocean floor. Specially designed surface ships would transport waste canisters from a port facility to the disposal site. Canisters would be buried from a few tens to a few hundreds of meters below the surface of ocean bottom sediments, and hence would not be in contact with the overlying ocean water. The concept is a multi-barrier approach for disposal. Barriers, including waste form, canister, ad deep ocean sediments, will separate wastes from the ocean environment. High-level wastes (HLW) would be stabilized by conversion into a leach-resistant solid form such as glass. This solid would be placed inside a metallic canister or other type of package which represents a second barrier. The deep ocean sediments, a third barrier, are discussed in the Feasibility Assessment section. The waste form and canister would provide a barrier for several hundred years, and the sediments would be relied upon as a barrier for thousands of years. 62 references, 3 figures, 2 tables.

  9. Briefing book on environmental and waste management activities

    SciTech Connect (OSTI)

    Quayle, T.A.

    1993-04-01T23:59:59.000Z

    The purpose of the Briefing Book is to provide current information on Environmental Restoration and Waste Management Activities at the Hanford Site. Each edition updates the information in the previous edition by deleting those sections determined not to be of current interest and adding new topics to keep up to date with the changing requirements and issues. This edition covers the period from October 15, 1992 through April 15, 1993.

  10. Activated carbon adsorbents from waste tires for air quality control

    SciTech Connect (OSTI)

    Lehmann, C.M.B.; Rostam-Abadi, M.; Rood, M.J.; Hsi, H.C.

    1999-07-01T23:59:59.000Z

    This study evaluates methodologies for utilizing waste tire rubber to produce carbonaceous adsorbents for use in air quality control operations. Such an approach provides a two-fold environmental and economic benefit. A recycling path is developed for waste tire rubber and new adsorbents are produced from a low cost feedstock for use in environmentally-related operations. Bench-scale and pilot-scale quantities of tire-derived activated carbon (TDAC) were produced from waste tire rubber. Raw tire rubber samples and devolatilized tire char were obtained from several US vendors. The raw samples were analyzed using proximate, ultimate, and elemental analyses. Batches of activated carbon samples were prepared using a bench-scale fixed-tubular reactor to prepare {approximately}10 g samples and a fluidized-bed reactor to prepare {approximately}100 g quantities. About 25 kg of activated carbon was also produced at a pilot-scale commercial facility. The resulting TDACs were then characterized by nitrogen adsorption at 77K. The sample surface areas were determined by the BET method, and the pore size distribution (PSD) was evaluated using the BJH model, and a 3-D PSD model. Performance of the TDACs was evaluated in their ability to remove gaseous mercury species from simulated power-plant flue-gas streams, and for the removal of organic compounds (e.g., acetone and 1,1,1-trichloroethane) from flowing gas streams.

  11. Waste Minimization Plans and activities in the MFD Plating Shop

    SciTech Connect (OSTI)

    Dini, J.W.; Steffani, C.P.

    1991-02-01T23:59:59.000Z

    The Plating Shop (Building 322), provides processes such as electroplating, electroless plating, anodizing, cleaning, etching, electroforming and chemical milling. We in MFD are committed to an active program on waste minimization, and the purpose of this document is to outline the plan of action. Our short range minimization goals are threefold: (1) Reduce our major waste stream by 90%, (2) Minimize discharge of rinse water to sewer system, and (3) Eliminate vapor degreasing in the shop. The intermediate goals consist of characterizing the waste streams and evaluating recovery processes. To do this, we first need to have the distillation unit operational and time to determine its effectiveness. If it proves to be as effective as we anticipate, we will perhaps purchase a second unit. Regardless, the streams that we can identify include: nickel, copper, rinses, acids, alkalies, electropolish and miscellaneous. Our goal is to utilize electrolytic processes to recover metals such as nickel and copper and processes such as ion exchange for some of the other streams. We intend to evaluate the full gamut of recycling processes available for these streams. We anticipate completing this phase of the minimization program by January 1993. The long range goal is zero discharge or since this could prove extremely difficult, development of processes that will allow us to produce a sludge cake that could be handled by our Hazardous Waste Management Group.

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

    SciTech Connect (OSTI)

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

    2013-07-01T23:59:59.000Z

    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)

  13. RADIOACTIVE DEMONSTRATIONS OF FLUIDIZED BED STEAM REFORMING WITH ACUTAL HANFORD LOW ACTIVITY WASTES VERIFYING FBSR AS A SUPPLEMENTARY TREATMENT

    SciTech Connect (OSTI)

    Jantzen, C.; Crawford, C.; Burket, P.; Bannochie, C.; Daniel, G.; Nash, C.; Cozzi, A.; Herman, C.

    2012-01-12T23:59:59.000Z

    The U.S. Department of Energy's Office of River Protection is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level waste (HLW) and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the cleanup mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA). Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. Fluidized Bed Steam Reforming (FBSR) is one of the supplementary treatments being considered. FBSR offers a moderate temperature (700-750 C) continuous method by which LAW and other secondary wastes can be processed irrespective of whether they contain organics, nitrates/nitrites, sulfates/sulfides, chlorides, fluorides, and/or radio-nuclides like I-129 and Tc-99. Radioactive testing of Savannah River LAW (Tank 50) shimmed to resemble Hanford LAW and actual Hanford LAW (SX-105 and AN-103) have produced a ceramic (mineral) waste form which is the same as the non-radioactive waste simulants tested at the engineering scale. The radioactive testing demonstrated that the FBSR process can retain the volatile radioactive components that cannot be contained at vitrification temperatures. The radioactive and nonradioactive mineral waste forms that were produced by co-processing waste with kaolin clay in an FBSR process are shown to be as durable as LAW glass.

  14. Mercury Reduction and Removal from High Level Waste at the Defense Waste Processing Facility - 12511

    SciTech Connect (OSTI)

    Behrouzi, Aria [Savannah River Remediation, LLC (United States); Zamecnik, Jack [Savannah River National Laboratory, Aiken, South Carolina, 29808 (United States)

    2012-07-01T23:59:59.000Z

    The Defense Waste Processing Facility processes legacy nuclear waste generated at the Savannah River Site during production of enriched uranium and plutonium required 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 is a safe, effective way to reduce the volume of the waste and stabilize the radionuclides. One of the constituents in the nuclear waste is mercury, which is present because it served as a catalyst in the dissolution of uranium-aluminum alloy fuel rods. At high temperatures mercury is corrosive to off-gas equipment, this poses a major challenge to the overall vitrification process in separating mercury from the waste stream prior to feeding the high temperature melter. Mercury is currently removed during the chemical process via formic acid reduction followed by steam stripping, which allows elemental mercury to be evaporated with the water vapor generated during boiling. The vapors are then condensed and sent to a hold tank where mercury coalesces and is recovered in the tank's sump via gravity settling. Next, mercury is transferred from the tank sump to a purification cell where it is washed with water and nitric acid and removed from the facility. Throughout the chemical processing cell, compounds of mercury exist in the sludge, condensate, and off-gas; all of which present unique challenges. Mercury removal from sludge waste being fed to the DWPF melter is required to avoid exhausting it to the environment or any negative impacts to the Melter Off-Gas system. The mercury concentration must be reduced to a level of 0.8 wt% or less before being introduced to the melter. Even though this is being successfully accomplished, the material balances accounting for incoming and collected mercury are not equal. In addition, mercury has not been effectively purified and collected in the Mercury Purification Cell (MPC) since 2008. A significant cleaning campaign aims to bring the MPC back up to facility housekeeping standards. Two significant investigations are being undertaken to restore mercury collection. The SMECT mercury pump has been removed from the tank and will be functionally tested. Also, research is being conducted by the Savannah River National Laboratory to determine the effects of antifoam addition on the behavior of mercury. These path forward items will help us better understand what is occurring in the mercury collection system and ultimately lead to an improved DWPF production rate and mercury recovery rate. (authors)

  15. Development and deployment of advanced corrosion monitoring systems for high-level waste tanks.

    SciTech Connect (OSTI)

    Terry, M. T. (Michael T.); Edgemon, G. L. (Glenn L.); Mickalonis, J. I. (John I.); Mizia, R. E. (Ronald E.)

    2002-01-01T23:59:59.000Z

    This paper describes the results of a collaborative technology development program, sponsored by the Tanks Focus Area, to use electrochemical noise (EN) for corrosion monitoring in underground storage tanks. These tanks, made of carbon or stainless steels, contain high-level radioactive liquid waste (HLW) generated by weapons production or radioactive liquid waste from nuclear fuel reprocessing activities at several Department of Energy (DOE) sites. The term EN is used to describe low frequency fluctuations in current and voltage measurements associated with corrosion. In their most basic form, EN-based corrosion monitoring systems measure and record these fluctuations over time from electrodes immersed in the environment of interest - in this case, radioactive tank waste. The resulting EN signals have characteristic patterns for different corrosion mechanisms. In recent years, engineers and scientists from several DOE sites, in collaboration with several private companies, have conducted laboratory studies and field applications to correlate the EN signals with corrosion mechanisms active in the radioactive waste tanks. The participating DOE sites are Hanford, Savannah River, Oak Ridge Reservation and the Idaho National Engineering and Environmental Laboratory. The commercial vendors have included HiLine Engineering and Fabrication, Inc., EIC Laboratories, Inc., and M A Technologies. Successful deployment of the EN technology will yield improved information of waste tank corrosion conditions, better tank management, and lower overall cost.

  16. Development and Deployment of Advanced Corrosion Monitoring Systems for High-Level Waste Tanks

    SciTech Connect (OSTI)

    Terry, M. T.; Edgemon, G. L.; Mickalonis, J. I.; Mizia, R. E.

    2002-02-26T23:59:59.000Z

    This paper describes the results of a collaborative technology development program, sponsored by the Tanks Focus Area, to use electrochemical noise (EN) for corrosion monitoring in underground storage tanks. These tanks, made of carbon or stainless steels, contain high-level radioactive liquid waste (HLW) generated by weapons production or radioactive liquid waste from nuclear fuel reprocessing activities at several Department of Energy (DOE) sites. The term EN is used to describe low frequency fluctuations in current and voltage measurements associated with corrosion. In their most basic form, EN-based corrosion monitoring systems measure and record these fluctuations over time from electrodes immersed in the environment of interest--in this case, radioactive tank waste. The resulting EN signals have characteristic patterns for different corrosion mechanisms. In recent years, engineers and scientists from several DOE sites, in collaboration with several private companies, have conducted laboratory studies and field applications to correlate the EN signals with corrosion mechanisms active in the radioactive waste tanks. The participating DOE sites are Hanford, Savannah River, Oak Ridge Reservation and the Idaho National Engineering and Environmental Laboratory. The commercial vendors have included HiLine Engineering and Fabrication, Inc., EIC Laboratories, Inc., and AEA Technologies. Successful deployment of the EN technology will yield improved information of waste tank corrosion conditions, better tank management, and lower overall cost.

  17. Glass Formulations for Immobilizing Hanford Low-Activity Wastes

    SciTech Connect (OSTI)

    Kim, D.S.; Elliott, M.L.; Smith, H.D.; Bagaasen, L.M.; Hrma, P.R. [Pacific Northwest National Laboratory, P.O. Box 999, Richland, WA 99352 (United States)

    2006-07-01T23:59:59.000Z

    Researchers at Pacific Northwest National Laboratory (PNNL) are developing and testing glasses for immobilizing low-activity wastes (LAW) for the full Hanford mission. PNNL is performing testing for low-activity waste glasses for both the Hanford Waste Treatment Plant (WTP) and the Bulk Vitrification Plant. The objective of this work is to increase the waste content of the glasses and ultimately increase the waste throughput of the LAW vitrification plants. This paper focuses on PNNL's development and testing of glasses for the Bulk Vitrification process. Collaborative studies are also being conducted with the Khlopin Radium Institute in St. Petersburg, Russia, to increase the solubility of sulfur in WTP glasses through the addition of trace chemicals to alter the glass chemistry. That research will be presented in a separate paper at this conference. Bulk Vitrification was selected as a potential supplemental treatment to accelerate the cleanup of LAW at Hanford. Also known as In-Container Vitrification{sup TM} (ICV{sup TM}), the Bulk Vitrification process combines soil, LAW, and chemical amendments; dries the mixture; and then vitrifies the material in a batch process in a refractory lined box. The process was developed by AMEC Earth and Environmental, Inc. (AMEC). Working with AMEC, PNNL developed a glass formulation that could incorporate a broad range of Hanford LAW. The initial glass development involved a 'nominal' waste composition, and a baseline glass was formulated and tested at crucible, engineering, and full scales. The performance of the baseline glass was then verified using a battery of laboratory tests as well as engineering-scale and full-scale ICV{sup TM} tests. Continued testing has focused on developing an acceptable operating envelope for the baseline glass. The current glass constraints are: - 17 {<=} Na{sub 2}O {<=} 22 mass%; - 3 {<=} B{sub 2}O{sub 3} {<=} 5 mass%; - 8 {<=} Al{sub 2}O{sub 3} {<=} 12.5 mass%; - 5.5 {<=} ZrO{sub 2} {<=} 8 mass%; 6.4 {<=} ZrO{sub 2} {<=} 8 mass% if Al{sub 2}O{sub 3} {>=} 9.5 mass%; - 40 {<=} SiO{sub 2} {<=} 48.5 mass%. Multiple samples from engineering-scale and full-scale ICV{sup TM} tests performed with a baseline glass formulation developed from crucible tests were analyzed for chemical composition, Product Consistency Test, Vapor Hydration Test, and the Toxicity Characteristic Leaching Procedure. The results show good agreement between glasses prepared in a crucible in the laboratory and the glasses from the larger scale tests. The results also show that the glass in the ICV{sup TM} box is homogeneous. Future testing is planned for optimizing the glass waste loading and qualifying a broader range of waste streams for treatment in the Bulk Vitrification process. This paper reviews the glass development and qualification process completed to date. This includes several series of crucible studies as well as confirmation testing at engineering-scale and full-scale. This formulation paper complements information presented by AMEC in an ICV{sup TM} processing paper. (authors)

  18. ROAD MAP FOR DEVELOPMENT OF CRYSTAL-TOLERANT HIGH LEVEL WASTE GLASSES

    SciTech Connect (OSTI)

    Fox, K.; Peeler, D.; Herman, C.

    2014-05-15T23:59:59.000Z

    The U.S. Department of Energy (DOE) is building a Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is being temporarily stored in 177 underground tanks. Efforts are being made to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. This road map guides the research and development for formulation and processing of crystaltolerant glasses, identifying near- and long-term activities that need to be completed over the period from 2014 to 2019. The primary objective is to maximize waste loading for Hanford waste glasses without jeopardizing melter operation by crystal accumulation in the melter or melter discharge riser. The potential applicability to the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) will also be addressed in this road map. The planned research described in this road map is motivated by the potential for substantial economic benefits (significant reductions in glass volumes) that will be realized if the current constraints (T1% for WTP and TL for DWPF) are approached in an appropriate and technically defensible manner for defense waste and current melter designs. 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 high waste loadings while still meeting process related limits and melter lifetime expectancies. The modeling effort will be an iterative process, where model form and a broader range of conditions, e.g., glass composition and temperature, will evolve as additional data on crystal accumulation are gathered. Model validation steps will be included to guide the development process and ensure the value of the effort (i.e., increased waste loading and waste throughput). A summary of the stages of the road map for developing the crystal-tolerant glass approach, their estimated durations, and deliverables is provided.

  19. High Flux Isotopes Reactor (HFIR) Cooling Towers Demolition Waste Management

    SciTech Connect (OSTI)

    Pudelek, R. E.; Gilbert, W. C.

    2002-02-26T23:59:59.000Z

    This paper describes the results of a joint initiative between Oak Ridge National Laboratory, operated by UT-Battelle, and Bechtel Jacobs Company, LLC (BJC) to characterize, package, transport, treat, and dispose of demolition waste from the High Flux Isotope Reactor (HFIR), Cooling Tower. The demolition and removal of waste from the site was the first critical step in the planned HFIR beryllium reflector replacement outage scheduled. The outage was scheduled to last a maximum of six months. Demolition and removal of the waste was critical because a new tower was to be constructed over the old concrete water basin. A detailed sampling and analysis plan was developed to characterize the hazardous and radiological constituents of the components of the Cooling Tower. Analyses were performed for Resource Conservation and Recovery Act (RCRA) heavy metals and semi-volatile constituents as defined by 40 CFR 261 and radiological parameters including gross alpha, gross beta, gross gamma, alpha-emitting isotopes and beta-emitting isotopes. Analysis of metals and semi-volatile constituents indicated no exceedances of regulatory limits. Analysis of radionuclides identified uranium and thorium and associated daughters. In addition 60Co, 99Tc, 226Rm, and 228Rm were identified. Most of the tower materials were determined to be low level radioactive waste. A small quantity was determined not to be radioactive, or could be decontaminated. The tower was dismantled October 2000 to January 2001 using a detailed step-by-step process to aid waste segregation and container loading. The volume of waste as packaged for treatment was approximately 1982 cubic meters (70,000 cubic feet). This volume was comprised of plastic ({approx}47%), wood ({approx}38%) and asbestos transite ({approx}14%). The remaining {approx}1% consisted of the fire protection piping (contaminated with lead-based paint) and incidental metal from conduit, nails and braces/supports, and sludge from the basin. The waste, except for the asbestos, was volume reduced via a private contract mechanism established by BJC. After volume reduction, the waste was packaged for rail shipment. This large waste management project successfully met cost and schedule goals.

  20. active liquid wastes: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Assessment Institute Rethinking the Waste Hierarchy March 2005 Recommendations A number of specific recommendations for achieving cost-effective waste policies can be made...

  1. active liquid waste: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Assessment Institute Rethinking the Waste Hierarchy March 2005 Recommendations A number of specific recommendations for achieving cost-effective waste policies can be made...

  2. Symmetric Active/Active HighSymmetric Active/Active High Availability for HighAvailability for High--PerformancePerformance

    E-Print Network [OSTI]

    Engelmann, Christian

    Symmetric Active/Active HighSymmetric Active/Active High Availability for HighAvailability for High reliability and availability is deceasing rapidlyHPC system reliability and availability is deceasing rapidly memory)bit errors in ECC memory) High availability as well as high performance isHigh availability

  3. Corrosion assessment of refractory materials for high temperature waste vitrification

    SciTech Connect (OSTI)

    Marra, J.C.; Congdon, J.W.; Kielpinski, A.L. [and others

    1995-11-01T23:59:59.000Z

    A variety of vitrification technologies are being evaluated to immobilize radioactive and hazardous wastes following years of nuclear materials production throughout the Department of Energy (DOE) complex. The compositions and physical forms of these wastes are diverse ranging from inorganic sludges to organic liquids to heterogeneous debris. Melt and off-gas products can be very corrosive at the high temperatures required to melt many of these waste streams. Ensuring material durability is required to develop viable treatment processes. Corrosion testing of materials in some of the anticipated severe environments is an important aspect of the materials identification and selection process. Corrosion coupon tests on typical materials used in Joule heated melters were completed using glass compositions with high salt contents. The presence of chloride in the melts caused the most severe attack. In the metal alloys, oxidation was the predominant corrosion mechanism, while in the tested refractory material enhanced dissolution of the refractory into the glass was observed. Corrosion testing of numerous different refractory materials was performed in a plasma vitrification system using a surrogate heterogeneous debris waste. Extensive corrosion was observed in all tested materials.

  4. Long-term management of high-level radioactive waste (HLW) and...

    Office of Environmental Management (EM)

    Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF)...

  5. Coupled Model for Heat and Water Transport in a High Level Waste...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Model for Heat and Water Transport in a High Level Waste Repository in Salt Coupled Model for Heat and Water Transport in a High Level Waste Repository in Salt This report...

  6. Scenarios for the Hanford Immobilized Low-Activity Waste (ILAW) performance assessment

    SciTech Connect (OSTI)

    MANN, F.M.

    1999-03-17T23:59:59.000Z

    Scenarios describing representative exposure cases associated with the disposal of low activity waste from the Hanford Waste Tanks have been defined. These scenarios are based on guidance from the Department of Energy, the U.S. Nuclear Regulatory Commission, and previous Hanford waste disposal performance assessments.

  7. Performance objectives for the Hanford Immobilized Low-Activity Waste (ILAW) performance assessment

    SciTech Connect (OSTI)

    MANN, F.M.

    1999-03-17T23:59:59.000Z

    Performance objectives for the disposal of low activity waste from Hanford Waste Tanks have been developed. These objectives have been based on DOE requirements, programmatic requirements, and public involvement. The DOE requirements include regulations that direct the performance assessment and are cited within the Radioactive Waste Management Order (DOE Order 435.1). Performance objectives for other DOE complex performance assessments have been included.

  8. Conceptual design statement of work for the immobilized low-activity waste interim storage facility project

    SciTech Connect (OSTI)

    Carlson, T.A., Fluor Daniel Hanford

    1997-02-06T23:59:59.000Z

    The Immobilized Low-Activity Waste Interim Storage subproject will provide storage capacity for immobilized low-activity waste product sold to the U.S. Department of Energy by the privatization contractor. This statement of work describes the work scope (encompassing definition of new installations and retrofit modifications to four existing grout vaults), to be performed by the Architect-Engineer, in preparation of a conceptual design for the Immobilized Low-Activity Waste Interim Storage Facility.

  9. Deep borehole disposal of high-level radioactive waste.

    SciTech Connect (OSTI)

    Stein, Joshua S.; Freeze, Geoffrey A.; Brady, Patrick Vane; Swift, Peter N.; Rechard, Robert Paul; Arnold, Bill Walter; Kanney, Joseph F.; Bauer, Stephen J.

    2009-07-01T23:59:59.000Z

    Preliminary evaluation of deep borehole disposal of high-level radioactive waste and spent nuclear fuel indicates the potential for excellent long-term safety performance at costs competitive with mined repositories. Significant fluid flow through basement rock is prevented, in part, by low permeabilities, poorly connected transport pathways, and overburden self-sealing. Deep fluids also resist vertical movement because they are density stratified. Thermal hydrologic calculations estimate the thermal pulse from emplaced waste to be small (less than 20 C at 10 meters from the borehole, for less than a few hundred years), and to result in maximum total vertical fluid movement of {approx}100 m. Reducing conditions will sharply limit solubilities of most dose-critical radionuclides at depth, and high ionic strengths of deep fluids will prevent colloidal transport. For the bounding analysis of this report, waste is envisioned to be emplaced as fuel assemblies stacked inside drill casing that are lowered, and emplaced using off-the-shelf oilfield and geothermal drilling techniques, into the lower 1-2 km portion of a vertical borehole {approx}45 cm in diameter and 3-5 km deep, followed by borehole sealing. Deep borehole disposal of radioactive waste in the United States would require modifications to the Nuclear Waste Policy Act and to applicable regulatory standards for long-term performance set by the US Environmental Protection Agency (40 CFR part 191) and US Nuclear Regulatory Commission (10 CFR part 60). The performance analysis described here is based on the assumption that long-term standards for deep borehole disposal would be identical in the key regards to those prescribed for existing repositories (40 CFR part 197 and 10 CFR part 63).

  10. Nondestructive examination of DOE high-level waste storage tanks

    SciTech Connect (OSTI)

    Bush, S.; Bandyopadhyay, K.; Kassir, M.; Mather, B.; Shewmon, P.; Streicher, M.; Thompson, B.; van Rooyen, D.; Weeks, J.

    1995-05-01T23:59:59.000Z

    A number of DOE sites have buried tanks containing high-level waste. Tanks of particular interest am double-shell inside concrete cylinders. A program has been developed for the inservice inspection of the primary tank containing high-level waste (HLW), for testing of transfer lines and for the inspection of the concrete containment where possible. Emphasis is placed on the ultrasonic examination of selected areas of the primary tank, coupled with a leak-detection system capable of detecting small leaks through the wall of the primary tank. The NDE program is modelled after ASME Section XI in many respects, particularly with respects to the sampling protocol. Selected testing of concrete is planned to determine if there has been any significant degradation. The most probable failure mechanisms are corrosion-related so that the examination program gives major emphasis to possible locations for corrosion attack.

  11. Glass formulation for phase 1 high-level waste vitrification

    SciTech Connect (OSTI)

    Vienna, J.D.; Hrma, P.R.

    1996-04-01T23:59:59.000Z

    The purpose of this study is to provide potential glass formulations for prospective Phase 1 High-Level Waste (HLW) vitrification at Hanford. The results reported here will be used to aid in developing a Phase 1 HLW vitrification request for proposal (RFP) and facilitate the evaluation of ensuing proposals. The following factors were considered in the glass formulation effort: impact on total glass volume of requiring the vendor to process each of the tank compositions independently versus as a blend; effects of imposing typical values of B{sub 2}O{sub 3} content and waste loading in HLW borosilicate glasses as restrictions on the vendors (according to WAPS 1995, the typical values are 5--10 wt% B{sub 2}O{sub 3} and 20--40 wt% waste oxide loading); impacts of restricting the processing temperature to 1,150 C on eventual glass volume; and effects of caustic washing on any of the selected tank wastes relative to glass volume.

  12. INTERNATIONAL STUDIES OF ENHANCED WASTE LOADING AND IMPROVED MELT RATE FOR HIGH ALUMINA CONCENTRATION NUCLEAR WASTE GLASSES

    SciTech Connect (OSTI)

    Fox, K; David Peeler, D; James Marra, J

    2008-09-11T23:59:59.000Z

    The goal of this study was to determine the impacts of glass compositions with high aluminum concentrations on melter performance, crystallization and chemical durability for Savannah River Site (SRS) and Hanford waste streams. Glass compositions for Hanford targeted both high aluminum concentrations in waste sludge and a high waste loading in the glass. Compositions for SRS targeted Sludge Batch 5, the next sludge batch to be processed in the Defense Waste Processing Facility (DWPF), which also has a relatively high aluminum concentration. Three frits were selected for combination with the SRS waste to evaluate their impact on melt rate. The glasses were melted in two small-scale test melters at the V. G. Khlopin Radium Institute. The results showed varying degrees of spinel formation in each of the glasses. Some improvements in melt rate were made by tailoring the frit composition for the SRS feeds. All of the Hanford and SRS compositions had acceptable chemical durability.

  13. CHARACTERIZATION OF DEFENSE NUCLEAR WASTE USING HAZARDOUS WASTE GUIDANCE. APPLICATIONS TO HANFORD SITE ACCELERATED HIGH-LEVEL WASTE TREATMENT AND DISPOSAL MISSION0

    SciTech Connect (OSTI)

    Hamel, William; Huffman, Lori; Lerchen, Megan; Wiemers, Karyn

    2003-02-27T23:59:59.000Z

    Federal hazardous waste regulations were developed for management of industrial waste. These same regulations are also applicable for much of the nation's defense nuclear wastes. At the U.S. Department of Energy's (DOE) Hanford Site in southeast Washington State, one of the nation's largest inventories of nuclear waste remains in storage in large underground tanks. The waste's regulatory designation and its composition and form constrain acceptable treatment and disposal options. Obtaining detailed knowledge of the tank waste composition presents a significant portion of the many challenges in meeting the regulatory-driven treatment and disposal requirements for this waste. Key in applying the hazardous waste regulations to defense nuclear wastes is defining the appropriate and achievable quality for waste feed characterization data and the supporting evidence demonstrating that applicable requirements have been met at the time of disposal. Application of a performance-based approach to demonstrating achievable quality standards will be discussed in the context of the accelerated high-level waste treatment and disposal mission at the Hanford Site.

  14. Assessment of national systems for obtaining local acceptance of waste management siting and routing activities

    SciTech Connect (OSTI)

    Paige, H.W.; Lipman, D.S.; Owens, J.E.

    1980-07-01T23:59:59.000Z

    There is a rich mixture of formal and informal approaches being used in our sister nuclear democracies in their attempts to deal with the difficulties of obtaining local acceptance for siting of waste management facilities and activities. Some of these are meeting with a degree of success not yet achieved in the US. Although this survey documents and assesses many of these approaches, time did not permit addressing in any detail their relevance to common problems in the US. It would appear the US could benefit from a periodic review of the successes and failures of these efforts, including analysis of their applicability to the US system. Of those countries (Germany, Sweden, Switzerland, Japan, Belgium, and the US) who are working to a time table for the preparation of a high-level waste (HLW) repository, Germany is the only country to have gained local siting acceptance for theirs. With this (the most difficult of siting problems) behind them they appear to be in the best overall condition relative to waste management progress and plans. This has been achieved without a particularly favorable political structure, made up for by determination on the part of the political leadership. Of the remaining three countries studied (France, UK and Canada) France, with its AVM production facility, is clearly the world leader in the HLW immobilization aspect of waste management. France, Belgium and the UK appear to have the least favorable political structures and environments for arriving at waste management decisions. US, Switzerland and Canada appear to have the least favorable political structures and environments for arriving at waste management decisions.

  15. HIGH LEVEL WASTE SLUDGE BATCH 4 VARIABILITY STUDY

    SciTech Connect (OSTI)

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

    2006-10-02T23:59:59.000Z

    The Defense Waste Processing Facility (DWPF) is preparing for vitrification of High Level Waste (HLW) Sludge Batch 4 (SB4) in early FY2007. To support this process, the Savannah River National Laboratory (SRNL) has provided a recommendation to utilize Frit 503 for vitrifying this sludge batch, based on the composition projection provided by the Liquid Waste Organization on June 22, 2006. Frit 418 was also recommended for possible use during the transition from SB3 to SB4. A critical step in the SB4 qualification process is to demonstrate the applicability of the durability models, which are used as part of the DWPF's process control strategy, to the glass system of interest via a variability study. A variability study is an experimentally-driven assessment of the predictability and acceptability of the quality of the vitrified waste product that is anticipated from the processing of a sludge batch. At the DWPF, the durability of the vitrified waste product is not directly measured. Instead, the durability is predicted using a set of models that relate the Product Consistency Test (PCT) response of a glass to the chemical composition of that glass. In addition, a glass sample is taken during the processing of that sludge batch, the sample is transmitted to SRNL, and the durability is measured to confirm acceptance. The objective of a variability study is to demonstrate that these models are applicable to the glass composition region anticipated during the processing of the sludge batch - in this case the Frit 503 - SB4 compositional region. The success of this demonstration allows the DWPF to confidently rely on the predictions of the durability/composition models as they are used in the control of the DWPF process.

  16. Hanford Immobilized Low Activity Waste (ILAW) Performance Assessment 2001 Version [Formerly DOE/RL-97-69] [SEC 1 & 2

    SciTech Connect (OSTI)

    MANN, F.M.

    2000-08-01T23:59:59.000Z

    The Hanford Immobilized Low-Activity Waste Performance Assessment examines the long-term environmental and human health effects associated with the planned disposal of the vitrified low-activity fraction of waste presently contained in Hanford Site tanks. The tank waste is the byproduct of separating special nuclear materials from irradiated nuclear fuels over the past 50 years. This waste is stored in underground single- and double-shell tanks. The tank waste is to be retrieved, separated into low-activity and high-level fractions, and then immobilized by vitrification. The US. Department of Energy (DOE) plans to dispose of the low-activity fraction in the Hanford Site 200 East Area. The high-level fraction will be stored at the Hanford Site until a national repository is approved. This report provides the site-specific long-term environmental information needed by the DOE to modify the current Disposal Authorization Statement for the Hanford Site that would allow the following: construction of disposal trenches; and filling of these trenches with ILAW containers and filler material with the intent to dispose of the containers.

  17. High-Level Waste Corporate Board Meeting Agenda

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM Flash2011-12Approved on 24 July 2008 1 Office ofHighHigh-Level Waste

  18. High-Level Waste Corporate Board Performance Assessment Subcommittee

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM Flash2011-12Approved on 24 July 2008 1 Office ofHighHigh-Level WasteLevel

  19. Foreign programs for the storage of spent nuclear power plant fuels, high-level waste canisters and transuranic wastes

    SciTech Connect (OSTI)

    Harmon, K.M.; Johnson, A.B. Jr.

    1984-04-01T23:59:59.000Z

    The various national programs for developing and applying technology for the interim storage of spent fuel, high-level radioactive waste, and TRU wastes are summarized. Primary emphasis of the report is on dry storage techniques for uranium dioxide fuels, but data are also provided concerning pool storage.

  20. EIS-0063: Waste Management Operations, Double-Shell Tanks for Defense High Level Radioactive Waste Storage, Hanford Site, Richland, Washington

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy developed this statement to evaluate the existing tank design and consider additional specific design and safety feature alternatives for the thirteen tanks being constructed for storage of defense high-level radioactive liquid waste at the Hanford Site in Richland, Washington. This statement supplements ERDA-1538, "Final Environmental Statement on Waste Management Operation."

  1. High-level waste tank farm set point document

    SciTech Connect (OSTI)

    Anthony, J.A. III

    1995-01-15T23:59:59.000Z

    Setpoints for nuclear safety-related instrumentation are required for actions determined by the design authorization basis. Minimum requirements need to be established for assuring that setpoints are established and held within specified limits. This document establishes the controlling methodology for changing setpoints of all classifications. The instrumentation under consideration involve the transfer, storage, and volume reduction of radioactive liquid waste in the F- and H-Area High-Level Radioactive Waste Tank Farms. The setpoint document will encompass the PROCESS AREA listed in the Safety Analysis Report (SAR) (DPSTSA-200-10 Sup 18) which includes the diversion box HDB-8 facility. In addition to the PROCESS AREAS listed in the SAR, Building 299-H and the Effluent Transfer Facility (ETF) are also included in the scope.

  2. 4.5 Meter high level waste canister study

    SciTech Connect (OSTI)

    Calmus, R. B.

    1997-10-01T23:59:59.000Z

    The Tank Waste Remediation System (TWRS) Storage and Disposal Project has established the Immobilized High-Level Waste (IBLW) Storage Sub-Project to provide the capability to store Phase I and II BLW products generated by private vendors. A design/construction project, Project W-464, was established under the Sub-Project to provide the Phase I capability. Project W-464 will retrofit the Hanford Site Canister Storage Building (CSB) to accommodate the Phase I I-ILW products. Project W-464 conceptual design is currently being performed to interim store 3.0 m-long BLW stainless steel canisters with a 0.61 in diameter, DOE is considering using a 4.5 in canister of the same diameter to reduce permanent disposal costs. This study was performed to assess the impact of replacing the 3.0 in canister with the 4.5 in canister. The summary cost and schedule impacts are described.

  3. An Investigation into the Oxidation State of Molybdenum in Simplified High Level Nuclear Waste Glass Compositions

    E-Print Network [OSTI]

    Sheffield, University of

    An Investigation into the Oxidation State of Molybdenum in Simplified High Level Nuclear Waste of Mo in glasses containing simplified simulated high level nuclear waste (HLW) streams has been originating from the reprocessing of spent nuclear fuel. Experiments using simulated nuclear waste streams

  4. Characteristics Data Base: Programmer's guide to the High-Level Waste Data Base

    SciTech Connect (OSTI)

    Jones, K.E. (DataPhile, Inc., Knoxville, TN (USA)); Salmon, R. (Oak Ridge National Lab., TN (USA))

    1990-08-01T23:59:59.000Z

    The High-Level Waste Data Base is a menu-driven PC data base developed as part of OCRWM's technical data base on the characteristics of potential repository wastes, which also includes spent fuel and other materials. This programmer's guide completes the documentation for the High-Level Waste Data Base, the user's guide having been published previously. 3 figs.

  5. Conversion of Waste CO2 and Shale Gas to High-Value Chemicals

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Conversion of Waste CO 2 and Shale Gas to High-Value Chemicals Enabling high-yield, low-cost, low- temperature production of chemical intermediates Chemical intermediates,...

  6. Issues and design concepts for high-activity liquid packaging

    SciTech Connect (OSTI)

    Meinert, N.M. [Westinghouse Hanford Co., Richland, WA (United States); Riley, D. [Walla Walla College of Engineering, Walla Walla, WA (United States); Wells, A.H. [Nuclear Assurance Corp., Norcross, GA (United States)

    1994-02-01T23:59:59.000Z

    The tank waste pretreatment process involves the separation of low-level and high-level constituents. The liquid high-level defense production waste will be vitrified into thousands of glass logs at the US DOE sites and then transported to a high-level repository for final disposal. Pretreatment and vitrification technology will need to be developed and tested to assess cost-effectiveness. The appropriate pretreatment strategy for complex high-activity liquid will depend on proving a competent process. As technology development matures, actual liquid will be substituted for simulants, and pilot scale plants will replace laboratory scale process demonstrations. Development of this strategy depends on tank waste sample analyses and a high-activity liquid supply for process testing. However, high-activity liquid transportation beyond DOE site boundaries is limited to Type B quantities in volumes less than 50 mL; no licensed packaging exists for greater than 50 mL quantities. The following paper summarizes the need for a high-activity liquid packaging, and identifies the agencies effecting packaging design and transportation. The high-activity liquid packaging concept retrofits licensed spent fuel casks by replacing the spent fuel basket with a sturdy containment vessel appropriate for the chemical nature of the liquid. A Nuclear Packaging (Pacific Nuclear`s NuPat{trademark} 125-B) spent fuel cask was hypothetically retrofitted with a containment vessel filled with liquid source term, the radionuclide inventory contained in the liquid. The structural, thermal, dose rate, and criticality consequences of retrofitting the cask body were evaluated based on data in the 125-B Cask Safety Analysis Report for Packaging. In addition, future packaging development work is discussed.

  7. active waste solutions: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    solutions Computer Technologies and Information Sciences Websites Summary: -mining or radioactive waste storage. Despite the constraints imposed by geological considerations,...

  8. FLUIDIZED BED STEAM REFORMING FOR TREATMENT AND IMMOBILIZATION OF LOW-ACTIVITY WASTE

    SciTech Connect (OSTI)

    HEWITT WM

    2011-04-08T23:59:59.000Z

    This report is one of four reports written to provide background information regarding immobilization technologies remaining under consideration for supplemental immobilization of Hanford's low-activity waste. This paper provides the reader a general understanding of fluidized bed steam reforming and its possible application to treat and immobilize Hanford low-activity waste.

  9. A JOULE-HEATED MELTER TECHNOLOGY FOR THE TREATMENT AND IMMOBILIZATION OF LOW-ACTIVITY WASTE

    SciTech Connect (OSTI)

    KELLY SE

    2011-04-07T23:59:59.000Z

    This report is one of four reports written to provide background information regarding immobilization technologies remaining under consideration for supplemental immobilization of Hanford's low-activity waste. This paper provides the reader a general understanding of joule-heated ceramic lined melters and their application to Hanford's low-activity waste.

  10. BULK VITRIFICATION TECHNOLOGY FOR THE TREATMENT AND IMMOBILIZATION OF LOW-ACTIVITY WASTE

    SciTech Connect (OSTI)

    ARD KE

    2011-04-11T23:59:59.000Z

    This report is one of four reports written to provide background information regarding immobilization technologies under consideration for supplemental immobilization of Hanford's low-activity waste. This paper is intended to provide the reader with general understanding of Bulk Vitrification and how it might be applied to immobilization of Hanford's low-activity waste.

  11. Defense High-Level Waste Leaching Mechanisms Program. Final report

    SciTech Connect (OSTI)

    Mendel, J.E. (compiler)

    1984-08-01T23:59:59.000Z

    The Defense High-Level Waste Leaching Mechanisms Program brought six major US laboratories together for three years of cooperative research. The participants reached a consensus that solubility of the leached glass species, particularly solubility in the altered surface layer, is the dominant factor controlling the leaching behavior of defense waste glass in a system in which the flow of leachant is constrained, as it will be in a deep geologic repository. Also, once the surface of waste glass is contacted by ground water, the kinetics of establishing solubility control are relatively rapid. The concentrations of leached species reach saturation, or steady-state concentrations, within a few months to a year at 70 to 90/sup 0/C. Thus, reaction kinetics, which were the main subject of earlier leaching mechanisms studies, are now shown to assume much less importance. The dominance of solubility means that the leach rate is, in fact, directly proportional to ground water flow rate. Doubling the flow rate doubles the effective leach rate. This relationship is expected to obtain in most, if not all, repository situations.

  12. High Level Waste Management Division . H L W System Plan

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC) EnvironmentalGyroSolé(tm) Harmonicbet WhenHiggs Boson May| ArgonneHigh Level Waste.

  13. 3-D MAPPING TECHNOLOGIES FOR HIGH LEVEL WASTE TANKS

    SciTech Connect (OSTI)

    Marzolf, A.; Folsom, M.

    2010-08-31T23:59:59.000Z

    This research investigated four techniques that could be applicable for mapping of solids remaining in radioactive waste tanks at the Savannah River Site: stereo vision, LIDAR, flash LIDAR, and Structure from Motion (SfM). Stereo vision is the least appropriate technique for the solids mapping application. Although the equipment cost is low and repackaging would be fairly simple, the algorithms to create a 3D image from stereo vision would require significant further development and may not even be applicable since stereo vision works by finding disparity in feature point locations from the images taken by the cameras. When minimal variation in visual texture exists for an area of interest, it becomes difficult for the software to detect correspondences for that object. SfM appears to be appropriate for solids mapping in waste tanks. However, equipment development would be required for positioning and movement of the camera in the tank space to enable capturing a sequence of images of the scene. Since SfM requires the identification of distinctive features and associates those features to their corresponding instantiations in the other image frames, mockup testing would be required to determine the applicability of SfM technology for mapping of waste in tanks. There may be too few features to track between image frame sequences to employ the SfM technology since uniform appearance may exist when viewing the remaining solids in the interior of the waste tanks. Although scanning LIDAR appears to be an adequate solution, the expense of the equipment ($80,000-$120,000) and the need for further development to allow tank deployment may prohibit utilizing this technology. The development would include repackaging of equipment to permit deployment through the 4-inch access ports and to keep the equipment relatively uncontaminated to allow use in additional tanks. 3D flash LIDAR has a number of advantages over stereo vision, scanning LIDAR, and SfM, including full frame time-of-flight data (3D image) collected with a single laser pulse, high frame rates, direct calculation of range, blur-free images without motion distortion, no need for precision scanning mechanisms, ability to combine 3D flash LIDAR with 2D cameras for 2D texture over 3D depth, and no moving parts. The major disadvantage of the 3D flash LIDAR camera is the cost of approximately $150,000, not including the software development time and repackaging of the camera for deployment in the waste tanks.

  14. CEMENTITIOUS GROUT FOR CLOSING SRS HIGH LEVEL WASTE TANKS - #12315

    SciTech Connect (OSTI)

    Langton, C.; Burns, H.; Stefanko, D.

    2012-01-10T23:59:59.000Z

    In 1997, the first two United States Department of Energy (US DOE) high level waste tanks (Tanks 17-F and 20-F: Type IV, single shell tanks) were taken out of service (permanently closed) at the Savannah River Site (SRS). In 2012, the DOE plans to remove from service two additional Savannah River Site (SRS) Type IV high-level waste tanks, Tanks 18-F and 19-F. These tanks were constructed in the late 1950's and received low-heat waste and do not contain cooling coils. Operational closure of Tanks 18-F and 19-F is intended to be consistent with the applicable requirements of the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and will be performed in accordance with South Carolina Department of Health and Environmental Control (SCDHEC). The closure will physically stabilize two 4.92E+04 cubic meter (1.3 E+06 gallon) carbon steel tanks and isolate and stabilize any residual contaminants left in the tanks. The closure will also fill, physically stabilize and isolate ancillary equipment abandoned in the tanks. A Performance Assessment (PA) has been developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closure of the F-Area Tank Farm (FTF) waste tanks. Next generation flowable, zero-bleed cementitious grouts were designed, tested, and specified for closing Tanks 18-F and 19-F and for filling the abandoned equipment. Fill requirements were developed for both the tank and equipment grouts. All grout formulations were required to be alkaline with a pH of 12.4 and chemically reduction potential (Eh) of -200 to -400 to stabilize selected potential contaminants of concern. This was achieved by including Portland cement and Grade 100 slag in the mixes, respectively. Ingredients and proportions of cementitious reagents were selected and adjusted, respectively, to support the mass placement strategy developed by closure operations. Subsequent down selection was based on compressive strength and saturated hydraulic conductivity results. Fresh slurry property results were used as the first level of screening. A high range water reducing admixture and a viscosity modifying admixture were used to adjust slurry properties to achieve flowable grouts. Adiabatic calorimeter results were used as the second level screening. The third level of screening was used to design mixes that were consistent with the fill material parameters used in the F-Tank Farm Performance Assessment which was developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closures.

  15. Sequential Thermo-Hydraulic Modeling of Variably Saturated Flow in High-Level Radioactive Waste Repository

    E-Print Network [OSTI]

    Boyer, Edmond

    Sequential Thermo-Hydraulic Modeling of Variably Saturated Flow in High-Level Radioactive Waste-Malabry, France Key words: waste repository, geological disposal, thermo- hydraulic modeling Introduction The most long-lived radioactive wastes must be managed in a safe way for human health and for the environment

  16. Independent Oversight Activity Report, Hanford Waste Tank Farms...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    6, 2013 February 2014 Follow-up on Previously Identified Items Regarding Positive Ventilation of Hanford Underground Waste Tanks HIAR-HANFORD-2013-10-28 This Independent...

  17. activity waste storage: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    of distributed storage systems Engelmann, Christian 13 Waste-Lithium-Liquid (WLL) Flow Battery for Stationary Energy Storage Applications Youngsik Kim* and Nina MahootcheianAsl...

  18. Independent Activity Report, Office of River Protection Waste...

    Broader source: Energy.gov (indexed) [DOE]

    and Introduction of New Office of Safety and Emergency Management Evaluations Site Lead for the Office of River Protection Waste Treatment Plant and Tank Farms...

  19. activity waste vitrification: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    and characterization alternatives of mixed waste soil and debris at disposal area remedial action DARA solids storage facility (SSF) University of California eScholarship...

  20. Processing constraints on high-level nuclear waste glasses for Hanford Waste Vitrification Plant

    SciTech Connect (OSTI)

    Hrma, P. [Pacific Northwest Lab., Richland, WA (United States)

    1993-12-31T23:59:59.000Z

    The work presented in this paper is a part of a major technology program supported by the US Department of Energy (DOE) in preparation for the planned operation of the Hanford Waste Vitrification Plant (HWVP). Because composition of Hanford waste varies greatly, processability is a major concern for successful vitrification. This paper briefly surveys general aspects of waste glass processability and then discusses their ramifications for specific examples of Hanford waste streams.

  1. Improved Alumina Loading in High-Level Waste Glasses

    SciTech Connect (OSTI)

    Kim, D.; Vienna, J.D. [Pacific Northwest National Laboratory, Richland, WA (United States); Peeler, D.K.; Fox, K.M. [Savannah River National Laboratory, Aiken, SC (United States); Aloy, A.; Trofimenko, A.V. [V.G. Khlopin Radium Institute, St. Petersburg (Russian Federation); Gerdes, K.D. [EM-21, Office of Waste Processing, U.S. Department of Energy, Washington, DC (United States)

    2008-07-01T23:59:59.000Z

    Recent tank retrieval, blending, and treatment strategies at both the Savannah River Site (SRS) and Hanford have identified increased amounts of high-Al{sub 2}O{sub 3} waste streams that are scheduled to be processed through their respective high-level waste (HLW) vitrification facilities. It is well known that the addition of small amounts of Al{sub 2}O{sub 3} to borosilicate glasses generally enhances the durability of the waste glasses. However, at higher Al{sub 2}O{sub 3} concentrations nepheline (NaAlSiO{sub 4}) formation can result in a severe deterioration of the chemical durability of the slowly cooled glass near the center of the canister. Additionally, higher concentrations of Al{sub 2}O{sub 3} generally increase the liquidus temperature of the melt and decrease the processing rate. Pacific Northwest National Laboratory (PNNL), Savannah River National Laboratory (SRNL), and Khlopin Radium Institute (KRI) are jointly performing laboratory and scaled-melter tests, through US Department of Energy, EM-21 Office of Waste Processing program, to develop glass formulations with increased Al{sub 2}O{sub 3} concentrations. These glasses are formulated for specific DOE waste compositions at Hanford and Savannah River Site. The objectives are to avoid nepheline formation while maintaining or meeting waste loading and/or waste throughput expectations as well as satisfying critical process and product performance related constraints such as viscosity, liquidus temperature, and glass durability. This paper summarizes the results of recent tests of simulated Hanford HLW glasses containing up to 26 wt% Al{sub 2}O{sub 3} in glass. In summary: Glasses with Al{sub 2}O{sub 3} loading ranging from 25 to 27 wt% were formulated and tested at a crucible scale. Successful glass formulations with up to 26 wt% Al{sub 2}O{sub 3} that do not precipitate nepheline during CCC treatment and had spinel crystals 1 vol% or less after 24 hr heat treatment at 950 deg. C were obtained. The selected glass, HAL-17 with 26 wt% Al{sub 2}O{sub 3}, had viscosity and electrical conductivity within the boundaries for adequate processing in the Joule heated melters operated at 1150 deg. C. This HAL-17 glass was successfully processed using small-scale (SMK) and larger scale (EP-5) melters. There was no indication of spinel settling during processing. The product glass samples from these melter tests contained 1 to 4 vol% spinel crystals that are likely formed during cooling. The PCT tests on the product glasses are underway. The present study demonstrated that it is possible to formulate the glasses with up to 26 wt% Al{sub 2}O{sub 3} that satisfy the property requirements and is processable with Joule-heated melters operated at 1150 deg. C. The 'nepheline discriminator' for HAL-17 glass is 0.45, which supports that claim that the current rule ('nepheline discriminator' < 0.62) is too restrictive. Considering that the cost of HLW treatment is highly dependent on loading of waste in glass, this result provides a potential for significant cost saving for Hanford. The maximum Al{sub 2}O{sub 3} loading that can be achieved will also depend on concentrations of other components in wastes. For example, the loading of waste used in this study was also limited by the spinel crystallization after 950 deg. C 24 hr heat treatment, which suggests that the concentrations of spinel-forming components such as Fe{sub 2}O{sub 3}, Cr{sub 2}O{sub 3}, NiO, ZnO, and MnO would be critical in addition to Al{sub 2}O{sub 3} for the maximum Al{sub 2}O{sub 3} loading achievable. The observed glass production rate per unit melter surface area of 0.75 MT/(d.m{sup 2}) for SMK test is comparable to the design capacity of WTP HLW melters at 0.8 MT/(d.m{sup 2}). However, the test with EP-5 melter achieved 0.38 MT/(d.m{sup 2}), which is roughly a half of the WTP design capacity. This result may imply that the glass with 26 wt% Al{sub 2}O{sub 3} may not achieve the WTP design production rate. However, this hypothesis is not conclusive because of unknown effects of melter size and operation

  2. Long-Term Waste Package Degradation Studies at the Yucca Mountain Potential High-Level Nuclear Waste Repository

    SciTech Connect (OSTI)

    Mon, K. G.; Bullard, B. E.; Longsine, D. E.; Mehta, S.; Lee, J. H.; Monib, A. M.

    2002-02-26T23:59:59.000Z

    The Site Recommendation (SR) process for the potential repository for spent nuclear fuel (SNF) and high-level nuclear waste (HLW) at Yucca Mountain, Nevada is underway. Fulfillment of the requirements for substantially complete containment of the radioactive waste emplaced in the potential repository and subsequent slow release of radionuclides from the Engineered Barrier System (EBS) into the geosphere will rely on a robust waste container design, among other EBS components. Part of the SR process involves sensitivity studies aimed at elucidating which model parameters contribute most to the drip shield and waste package degradation characteristics. The model parameters identified included (a) general corrosion rate model parameters (temperature-dependence and uncertainty treatment), and (b) stress corrosion cracking (SCC) model parameters (uncertainty treatment of stress and stress intensity factor profiles in the Alloy 22 waste package outer barrier closure weld regions, the SCC initiation stress threshold, and the fraction of manufacturing flaws oriented favorably for through-wall penetration by SCC). These model parameters were reevaluated and new distributions were generated. Also, early waste package failures due to improper heat treatment were added to the waste package degradation model. The results of these investigations indicate that the waste package failure profiles are governed by the manufacturing flaw orientation model parameters and models used.

  3. High Level Waste Tank Closure Project at the Idaho National Engineering and Environmental Laboratory

    SciTech Connect (OSTI)

    Wessman, D. L.; Quigley, K. D.

    2002-02-27T23:59:59.000Z

    The Department of Energy, Idaho Operations Office (DOE-ID) is making preparations to close two underground high-level waste (HLW) storage tanks at the Idaho National Engineering and Environmental Laboratory (INEEL) to meet Resource Conservation and Recovery Act (RCRA) regulations and Department of Energy orders. Closure of these two tanks is scheduled for 2004 as the first phase in closure of the eleven 300,000 gallon tanks currently in service at the Idaho Nuclear Technology and Engineering Center (INTEC). The INTEC Tank Farm Facility (TFF) Closure sequence consists of multiple steps to be accomplished through the existing tank riser access points. Currently, the tank risers contain steam and process waste lines associated with the steam jets, corrosion coupons, and liquid level indicators. As necessary, this equipment will be removed from the risers to allow adequate space for closure equipment and activities.

  4. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-2000 Status Report

    SciTech Connect (OSTI)

    Herbst, Alan Keith; Mc Cray, John Alan; Kirkham, Robert John; Pao, Jenn Hai; Argyle, Mark Don; Lauerhass, Lance; Bendixsen, Carl Lee; Hinckley, Steve Harold

    2000-11-01T23:59:59.000Z

    The Low-Activity Waste Process Technology Program anticipated that grouting will be used for disposal of low-level and transuranic wastes generated at the Idaho Nuclear Technology Engineering Center (INTEC). During fiscal year 2000, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed using silica gel and other absorbents to solidify sodium-bearing wastes. A feasibility study and conceptual design were completed for the construction of a grout pilot plant for simulated wastes and demonstration facility for actual wastes.

  5. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-2000 Status Report

    SciTech Connect (OSTI)

    Herbst, A.K.; McCray, J.A.; Kirkham, R.J.; Pao, J.; Argyle, M.D.; Lauerhass, L.; Bendixsen, C.L.; Hinckley, S.H.

    2000-10-31T23:59:59.000Z

    The Low-Activity Waste Process Technology Program anticipated that grouting will be used for disposal of low-level and transuranic wastes generated at the Idaho Nuclear Technology Engineering Center (INTEC). During fiscal year 2000, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed using silica gel and other absorbents to solidify sodium-bearing wastes. A feasibility study and conceptual design were completed for the construction of a grout pilot plant for simulated wastes and demonstration facility for actual wastes.

  6. PLUTONIUM/HIGH-LEVEL VITRIFIED WASTE BDBE DOSE CALCULATION

    SciTech Connect (OSTI)

    D.C. Richardson

    2003-03-19T23:59:59.000Z

    In accordance with the Nuclear Waste Policy Amendments Act of 1987, Yucca Mountain was designated as the site to be investigated as a potential repository for the disposal of high-level radioactive waste. The Yucca Mountain site is an undeveloped area located on the southwestern edge of the Nevada Test Site (NTS), about 100 miles northwest of Las Vegas. The site currently lacks rail service or an existing right-of-way. If the Yucca Mountain site is found suitable for the repository, rail service is desirable to the Office of Civilian Waste Management (OCRWM) Program because of the potential of rail transportation to reduce costs and to reduce the number of shipments relative to highway transportation. A Preliminary Rail Access Study evaluated 13 potential rail spur options. Alternative routes within the major options were also developed. Each of these options was then evaluated for potential land use conflicts and access to regional rail carriers. Three potential routes having few land use conflicts and having access to regional carriers were recommended for further investigation. Figure 1-1 shows these three routes. The Jean route is estimated to be about 120 miles long, the Carlin route to be about 365 miles long, and Caliente route to be about 365 miles long. The remaining ten routes continue to be monitored and should any of the present conflicts change, a re-evaluation of that route will be made. Complete details of the evaluation of the 13 routes can be found in the previous study. The DOE has not identified any preferred route and recognizes that the transportation issues need a full and open treatment under the National Environmental Policy Act. The issue of transportation will be included in public hearings to support development of the Environmental Impact Statement (EIS) proceedings for either the Monitored Retrievable Storage Facility or the Yucca Mountain Project or both.

  7. Advanced and developmental technologies for treatment and volume reduction of dry active wastes

    SciTech Connect (OSTI)

    Kohout, R. [R. Kohout & Associates, Ltd., Toronto (Canada)

    1994-12-31T23:59:59.000Z

    The nuclear power industry processes Dry Active Wastes (DAW) to achieve cost-effective volume reduction and/or to produce a residue that is more compatible with final disposal criteria. The two principal processes currently used by the industry are compaction and incineration. Although incineration is often considered the process of choice, capital and operating cost are often high, and in some countries, public opposition and lengthy permitting processes result in expensive delays to bringing the process to operation. Therefore, alternative treatment options (mechanical, thermal, chemical, and biological) are being investigated to provide timely, cost-effective options for industry use. An overview of those developmental processes considered applicable to processing DAW is presented. In each category, {open_quotes}established{close_quotes} processes are mentioned and/or referenced, but the focus is on {open_quotes}potential{close_quotes} technologies and the status of their development. The emphasis is on processing DAW, and therefore, those developmental processes that primarily treat solids in aqueous streams and melting/sintering technologies, both of lesser applicability to nuclear utility wastes, have been omitted. Included are those developmental technologies that appear to have a potential for radioactive waste application based on development on demonstration programs.

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

    SciTech Connect (OSTI)

    Not Available

    1983-06-01T23:59:59.000Z

    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.

  9. EIS-0120: Waste Management Activities for Groundwater Protection, Savannah River Plant, Aiken, South Carolina

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy has prepared this environmental impact statement to assess the environmental consequences of the implementation of modified waste management activities for hazardous, low-level radioactive, and mixed wastes for the protection of groundwater, human health, and the environment at its Savannah River Plant in Aiken, South Carolina.

  10. EIS-0120: Waste Management Activities for Groundwater Protection, Savannah River Plant, Aiken, SC

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy has prepared this environmental impact statement to assess the environmental consequences of the implementation of modified waste management activities for hazardous, low-level radioactive, and mixed wastes for the protection of groundwater, human health, and the environment at its Savannah River Plant in Aiken, South Carolina.

  11. JET MIXING ANALYSIS FOR SRS HIGH-LEVEL WASTE RECOVERY

    SciTech Connect (OSTI)

    Lee, S.

    2011-07-05T23:59:59.000Z

    The process of recovering the waste in storage tanks at the Savannah River Site (SRS) typically requires mixing the contents of the tank to ensure uniformity of the discharge stream. Mixing is accomplished with one to four slurry pumps located within the tank liquid. The slurry pump may be fixed in position or they may rotate depending on the specific mixing requirements. The high-level waste in Tank 48 contains insoluble solids in the form of potassium tetraphenyl borate compounds (KTPB), monosodium titanate (MST), and sludge. Tank 48 is equipped with 4 slurry pumps, which are intended to suspend the insoluble solids prior to transfer of the waste to the Fluidized Bed Steam Reformer (FBSR) process. The FBSR process is being designed for a normal feed of 3.05 wt% insoluble solids. A chemical characterization study has shown the insoluble solids concentration is approximately 3.05 wt% when well-mixed. The project is requesting a Computational Fluid Dynamics (CFD) mixing study from SRNL to determine the solids behavior with 2, 3, and 4 slurry pumps in operation and an estimate of the insoluble solids concentration at the suction of the transfer pump to the FBSR process. The impact of cooling coils is not considered in the current work. The work consists of two principal objectives by taking a CFD approach: (1) To estimate insoluble solids concentration transferred from Tank 48 to the Waste Feed Tank in the FBSR process and (2) To assess the impact of different combinations of four slurry pumps on insoluble solids suspension and mixing in Tank 48. For this work, several different combinations of a maximum of four pumps are considered to determine the resulting flow patterns and local flow velocities which are thought to be associated with sludge particle mixing. Two different elevations of pump nozzles are used for an assessment of the flow patterns on the tank mixing. Pump design and operating parameters used for the analysis are summarized in Table 1. The baseline pump orientations are chosen by the previous work [Lee et. al, 2008] and the initial engineering judgement for the conservative flow estimate since the modeling results for the other pump orientations are compared with the baseline results. As shown in Table 1, the present study assumes that each slurry pump has 900 gpm flowrate for the tank mixing analysis, although the Standard Operating Procedure for Tank 48 currently limits the actual pump speed and flowrate to a value less than 900 gpm for a 29 inch liquid level. Table 2 shows material properties and weight distributions for the solids to be modeled for the mixing analysis in Tank 48.

  12. Phosphate bonded structural products from high volume wastes

    DOE Patents [OSTI]

    Singh, D.; Wagh, A.S.

    1998-12-08T23:59:59.000Z

    A method to produce structural products from benign waste is provided comprising mixing pretreated oxide with phosphoric acid to produce an acid solution, mixing the acid solution with waste particles to produce a slurry, and allowing the slurry to cure. The invention also provides for a structural material comprising waste particles enveloped by an inorganic binder. 1 fig.

  13. EIS-0303: Savannah River Site High-Level Waste Tank Closure

    Broader source: Energy.gov [DOE]

    This EIS evaluates alternatives for closing 49 high-level radioactive waste tanks and associated equipment such as evaporator systems, transfer pipelines, diversion boxes, and pump pits. DOE...

  14. EIS-0287: Idaho High-Level Waste and Facilities Disposition Final...

    Office of Environmental Management (EM)

    EIS-0287 (September 2002) This EIS analyzes the potential environmental consequences of alternatives for managing high-level waste (HLW) calcine, mixed transuranic wastesodium...

  15. High level waste tank farm setpoint document. Revision 1

    SciTech Connect (OSTI)

    Anthony, J.A. III

    1995-01-31T23:59:59.000Z

    Revision 1 modifies Attachment I of this Technical Report as a result of a meeting which was held Friday, January 27, 1994 between Maintenance, Work Control, and Engineering to discuss report contents. Upon completion of the meeting, the Flow Chart was edited accordingly. Attachment 2 is modified for clerical reasons. Setpoints for nuclear safety-related instrumentation are required for actions determined by the design authorization basis. Minimum requirements need to be established for assuring that setpoints are established and held within specified limits. This document establishes the controlling methodology for changing setpoints of all classifications. The instrumentation under consideration involve the transfer, storage, and volume reduction of radioactive liquid waste in the F- and H-Area High-Level Radioactive Waste Tank Fanns. The setpoint document (Appendix 2) will encompass the PROCESS AREA listed in the Safety Analysis Report (SAR) (DSTSA-200-10 Sup 18) which includes the diversion box HDB-8 facility. In addition to the PROCESS AREAS listed in the SAR, Building 299-H and the Effluent Transfer Facility (ETF) are also included in the scope.

  16. High specific activity silicon-32

    DOE Patents [OSTI]

    Phillips, D.R.; Brzezinski, M.A.

    1996-06-11T23:59:59.000Z

    A process for preparation of silicon-32 is provided and includes contacting an irradiated potassium chloride target, including spallation products from a prior irradiation, with sufficient water, hydrochloric acid or potassium hydroxide to form a solution, filtering the solution, adjusting pH of the solution from about 5.5 to about 7.5, admixing sufficient molybdate-reagent to the solution to adjust the pH of the solution to about 1.5 and to form a silicon-molybdate complex, contacting the solution including the silicon-molybdate complex with a dextran-based material, washing the dextran-based material to remove residual contaminants such as sodium-22, separating the silicon-molybdate complex from the dextran-based material as another solution, adding sufficient hydrochloric acid and hydrogen peroxide to the solution to prevent reformation of the silicon-molybdate complex and to yield an oxidation state of the molybdate adapted for subsequent separation by an anion exchange material, contacting the solution with an anion exchange material whereby the molybdate is retained by the anion exchange material and the silicon remains in solution, and optionally adding sufficient alkali metal hydroxide to adjust the pH of the solution to about 12 to 13. Additionally, a high specific activity silicon-32 product having a high purity is provided.

  17. High specific activity silicon-32

    DOE Patents [OSTI]

    Phillips, Dennis R. (Los Alamos, NM); Brzezinski, Mark A. (Santa Barbara, CA)

    1996-01-01T23:59:59.000Z

    A process for preparation of silicon-32 is provided and includes contacting an irradiated potassium chloride target, including spallation products from a prior irradiation, with sufficient water, hydrochloric acid or potassium hydroxide to form a solution, filtering the solution, adjusting pH of the solution to from about 5.5 to about 7.5, admixing sufficient molybdate-reagent to the solution to adjust the pH of the solution to about 1.5 and to form a silicon-molybdate complex, contacting the solution including the silicon-molybdate complex with a dextran-based material, washing the dextran-based material to remove residual contaminants such as sodium-22, separating the silicon-molybdate complex from the dextran-based material as another solution, adding sufficient hydrochloric acid and hydrogen peroxide to the solution to prevent reformation of the silicon-molybdate complex and to yield an oxidization state of the molybdate adapted for subsequent separation by an anion exchange material, contacting the solution with an anion exchange material whereby the molybdate is retained by the anion exchange material and the silicon remains in solution, and optionally adding sufficient alkali metal hydroxide to adjust the pH of the solution to about 12 to 13. Additionally, a high specific activity silicon-32 product having a high purity is provided.

  18. I-NERI Annual Technical Progress Report 2007-004-K Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing

    SciTech Connect (OSTI)

    S. Frank

    2010-09-01T23:59:59.000Z

    The current method for the immobilization of fission products that accumulate in electrorefiner salt during the electrochemical processing of used metallic nuclear fuel is to encapsulate the electrorefiner salt in a glass-bonded sodalite ceramic waste form. This process was developed by Argonne National Laboratory in the USA and is currently performed at the Idaho National Laboratory for the treatment of Experimental Breeder Reactor-II (EBR-II) used fuel. This process utilizes a “once-through” option for the disposal of spent electrorefiner salt; where, after the treatment of the EBR-II fuel, the electrorefiner salt containing the active fission products will be disposed of in the ceramic waste form (CWF). The CWF produced will have low fission product loading of approximately 2 to 5 weight percent due to the limited fuel inventory currently being processed. However; the design and implementation of advanced electrochemical processing facilities to treat used fuel would process much greater quantities fuel. With an advanced processing facility, it would be necessary to selectively remove fission products from the electrorefiner salt for salt recycle and to concentrate the fission products to reduce the volume of high-level waste from the treatment facility. The Korean Atomic Energy Research Institute and the Idaho National Laboratory have been collaborating on I-NERI research projects for a number of years to investigate both aspects of selective fission product separation from electrorefiner salt, and to develop advanced waste forms for the immobilization of the collected fission products. The first joint KAERI/INL I-NERI project titled: 2006-002-K, Separation of Fission Products from Molten LiCl-KCl Salt Used for Electrorefining of Metal Fuels, was successfully completed in 2009 by concentrating and isolating fission products from actual electrorefiner salt used for the treated used EBR-II fuel. Two separation methods were tested and from these tests were produced concentrated salt products that acted as the feed material for development of advanced waste forms investigated in this proposal. Accomplishments from the first year activities associated with this I-NERI project included the down selection of candidate waste forms to immobilize fission products separated from electrorefiner salt, and the design of equipment to fabricate actual waste forms in the Hot Fuels Examination Facility (HFEF) at the INL. Reported in this document are accomplishments from the second year (FY10) work performed at the INL, and includes the testing of waste form fabrication equipment, repeating the fission product precipitation experiment, and initial waste form fabrication efforts.

  19. High Level Waste System Impacts from Acid Dissolution of Sludge

    SciTech Connect (OSTI)

    KETUSKY, EDWARD

    2006-04-20T23:59:59.000Z

    This research evaluates the ability of OLI{copyright} equilibrium based software to forecast Savannah River Site High Level Waste system impacts from oxalic acid dissolution of Tank 1-15 sludge heels. Without further laboratory and field testing, only the use of oxalic acid can be considered plausible to support sludge heel dissolution on multiple tanks. Using OLI{copyright} and available test results, a dissolution model is constructed and validated. Material and energy balances, coupled with the model, identify potential safety concerns. Overpressurization and overheating are shown to be unlikely. Corrosion induced hydrogen could, however, overwhelm the tank ventilation. While pH adjustment can restore the minimal hydrogen generation, resultant precipitates will notably increase the sludge volume. OLI{copyright} is used to develop a flowsheet such that additional sludge vitrification canisters and other negative system impacts are minimized. Sensitivity analyses are used to assess the processability impacts from variations in the sludge/quantities of acids.

  20. Symmetric Active/Active High Availability for High-Performance Computing System Services

    E-Print Network [OSTI]

    He, Xubin "Ben"

    Symmetric Active/Active High Availability for High-Performance Computing System Services Christian to pave the way for high avail- ability in high-performance computing (HPC) by focusing on efficient symmetric active/active high availability for multiple redundant head and service nodes running in virtual

  1. Determination of total cyanide in Hanford Site high-level wastes

    SciTech Connect (OSTI)

    Winters, W.I. [Westinghouse Hanford Co., Richland, WA (United States); Pool, K.H. [Pacific Northwest Lab., Richland, WA (United States)

    1994-05-01T23:59:59.000Z

    Nickel ferrocyanide compounds (Na{sub 2-x}Cs{sub x}NiFe (CN){sub 6}) were produced in a scavenging process to remove {sup 137}Cs from Hanford Site single-shell tank waste supernates. Methods for determining total cyanide in Hanford Site high-level wastes are needed for the evaluation of potential exothermic reactions between cyanide and oxidizers such as nitrate and for safe storage, processing, and management of the wastes in compliance with regulatory requirements. Hanford Site laboratory experience in determining cyanide in high-level wastes is summarized. Modifications were made to standard cyanide methods to permit improved handling of high-level waste samples and to eliminate interferences found in Hanford Site waste matrices. Interferences and associated procedure modifications caused by high nitrates/nitrite concentrations, insoluble nickel ferrocyanides, and organic complexants are described.

  2. Vit Plant receives and sets key air filtration equipment for Low Activity Waste Facility

    Broader source: Energy.gov [DOE]

    WTP lifted a nearly 100-ton carbon bed absorber into the Low-Activity Waste Facility. This key piece of air-filtration equipment will remove mercury and acidic gases before air is channeled through...

  3. Alternatives Generation and Analysis for Phase 1 High Level Waste Feed Tanks Selection

    SciTech Connect (OSTI)

    CRAWFORD, T.W.

    1999-08-16T23:59:59.000Z

    A recent revision of the US. Department of Energy privatization contract for the immobilization of high-level waste (HLW) at Hanford necessitates the investigation of alternative waste feed sources to meet contractual feed requirements. This analysis identifies wastes to be considered as HLW feeds and develops and conducts alternative analyses to comply with established criteria. A total of 12,426 cases involving 72 waste streams are evaluated and ranked in three cost-based alternative models. Additional programmatic criteria are assessed against leading alternative options to yield an optimum blended waste feed stream.

  4. RECENT PROCESS AND EQUIPMENT IMPROVEMENTS TO INCREASE HIGH LEVEL WASTE THROUGHPUT AT THE DEFENSE WASTE PROCESSING FACILITY

    SciTech Connect (OSTI)

    Odriscoll, R; Allan Barnes, A; Jim Coleman, J; Timothy Glover, T; Robert Hopkins, R; Dan Iverson, D; Jeff Leita, J

    2008-01-15T23:59:59.000Z

    The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF) began stabilizing high level waste (HLW) in a glass matrix in 1996. Over the past few years, there have been several process and equipment improvements at the DWPF to increase the rate at which the high level waste can be stabilized. These improvements have either directly increased waste processing rates or have desensitized the process to upsets, thereby minimizing downtime and increasing production. Improvements due to optimization of waste throughput with increased HLW loading of the glass resulted in a 6% waste throughput increase based upon operational efficiencies. Improvements in canister production include the pour spout heated bellows liner (5%), glass surge (siphon) protection software (2%), melter feed pump software logic change to prevent spurious interlocks of the feed pump with subsequent dilution of feed stock (2%) and optimization of the steam atomized scrubber (SAS) operation to minimize downtime (3%) for a total increase in canister production of 12%. A number of process recovery efforts have allowed continued operation. These include the off gas system pluggage and restoration, slurry mix evaporator (SME) tank repair and replacement, remote cleaning of melter top head center nozzle, remote melter internal inspection, SAS pump J-Tube recovery, inadvertent pour scenario resolutions, dome heater transformer bus bar cooling water leak repair and new Infra-red camera for determination of glass height in the canister are discussed.

  5. SETTLING OF SPINEL IN A HIGH-LEVEL WASTE GLASS MELTER

    SciTech Connect (OSTI)

    Pavel Hrma; Pert Schill; Lubomir Nemec

    2002-01-07T23:59:59.000Z

    High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors called melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150 decrees C to create a melt that becomes glass on cooling.

  6. Immobilized high-level waste interim storage alternatives generation and analysis and decision report

    SciTech Connect (OSTI)

    CALMUS, R.B.

    1999-05-18T23:59:59.000Z

    This report presents a study of alternative system architectures to provide onsite interim storage for the immobilized high-level waste produced by the Tank Waste Remediation System (TWRS) privatization vendor. It examines the contract and program changes that have occurred and evaluates their impacts on the baseline immobilized high-level waste (IHLW) interim storage strategy. In addition, this report documents the recommended initial interim storage architecture and implementation path forward.

  7. STATUS OF THE DEVELOPMENT OF IN-TANK/AT-TANK SEPARATIONS TECHNOLOGIES FOR FOR HIGH-LEVEL WASTE PROCESSING FOR THE U.S. DEPARTMENT OF ENERGY

    SciTech Connect (OSTI)

    Aaron, G.; Wilmarth, B.

    2011-09-19T23:59:59.000Z

    Within the U.S. Department of Energy's (DOE) Office of Technology Innovation and Development, the Office of Waste Processing manages a research and development program related to the treatment and disposition of radioactive waste. At the Savannah River (South Carolina) and Hanford (Washington) Sites, approximately 90 million gallons of waste are distributed among 226 storage tanks (grouped or collocated in 'tank farms'). This waste may be considered to contain mixed and stratified high activity and low activity constituent waste liquids, salts and sludges that are collectively managed as high level waste (HLW). A large majority of these wastes and associated facilities are unique to the DOE, meaning many of the programs to treat these materials are 'first-of-a-kind' and unprecedented in scope and complexity. As a result, the technologies required to disposition these wastes must be developed from basic principles, or require significant re-engineering to adapt to DOE's specific applications. Of particular interest recently, the development of In-tank or At-Tank separation processes have the potential to treat waste with high returns on financial investment. The primary objective associated with In-Tank or At-Tank separation processes is to accelerate waste processing. Insertion of the technologies will (1) maximize available tank space to efficiently support permanent waste disposition including vitrification; (2) treat problematic waste prior to transfer to the primary processing facilities at either site (i.e., Hanford's Waste Treatment and Immobilization Plant (WTP) or Savannah River's Salt Waste Processing Facility (SWPF)); and (3) create a parallel treatment process to shorten the overall treatment duration. This paper will review the status of several of the R&D projects being developed by the U.S. DOE including insertion of the ion exchange (IX) technologies, such as Small Column Ion Exchange (SCIX) at Savannah River. This has the potential to align the salt and sludge processing life cycle, thereby reducing the Defense Waste Processing Facility (DWPF) mission by 7 years. Additionally at the Hanford site, problematic waste streams, such as high boehmite and phosphate wastes, could be treated prior to receipt by WTP and thus dramatically improve the capacity of the facility to process HLW. Treatment of boehmite by continuous sludge leaching (CSL) before receipt by WTP will dramatically reduce the process cycle time for the WTP pretreatment facility, while treatment of phosphate will significantly reduce the number of HLW borosilicate glass canisters produced at the WTP. These and other promising technologies will be discussed.

  8. Road Map for Development of Crystal-Tolerant High Level Waste Glasses

    SciTech Connect (OSTI)

    Matyas, Josef; Vienna, John D.; Peeler, David; Fox, Kevin; Herman, Connie; Kruger, Albert A.

    2014-05-31T23:59:59.000Z

    This road map guides the research and development for formulation and processing of crystal-tolerant glasses, identifying near- and long-term activities that need to be completed over the period from 2014 to 2019. The primary objective is to maximize waste loading for Hanford waste glasses without jeopardizing melter operation by crystal accumulation in the melter or melter discharge riser. The potential applicability to the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) is also addressed in this road map.

  9. Data Packages for the Hanford Immobilized Low Activity Tank Waste Performance Assessment 2001 Version [SEC 1 THRU 5

    SciTech Connect (OSTI)

    MANN, F.M.

    2000-03-02T23:59:59.000Z

    Data package supporting the 2001 Immobilized Low-Activity Waste Performance Analysis. Geology, hydrology, geochemistry, facility, waste form, and dosimetry data based on recent investigation are provided. Verification and benchmarking packages for selected software codes are provided.

  10. Parametric Analyses of Heat Removal from High Level Waste Tanks

    SciTech Connect (OSTI)

    TRUITT, J.B.

    2000-06-05T23:59:59.000Z

    The general thermal hydraulics program GOTH-SNF was used to predict the thermal response of the waste in tanks 241-AY-102 and 241-AZ-102 when mixed by two 300 horsepower mixer pumps. This mixing was defined in terms of a specific waste retrieval scenario. Both dome and annulus ventilation system flow are necessary to maintain the waste within temperature control limits during the mixing operation and later during the sludge-settling portion of the scenario are defined.

  11. High-level waste issues and resolutions document

    SciTech Connect (OSTI)

    Not Available

    1994-05-01T23:59:59.000Z

    The High-Level Waste (HLW) Issues and Resolutions Document recognizes US Department of Energy (DOE) complex-wide HLW issues and offers potential corrective actions for resolving these issues. Westinghouse Management and Operations (M&O) Contractors are effectively managing HLW for the Department of Energy at four sites: Idaho National Engineering Laboratory (INEL), Savannah River Site (SRS), West Valley Demonstration Project (WVDP), and Hanford Reservation. Each site is at varying stages of processing HLW into a more manageable form. This HLW Issues and Resolutions Document identifies five primary issues that must be resolved in order to reach the long-term objective of HLW repository disposal. As the current M&O contractor at DOE`s most difficult waste problem sites, Westinghouse recognizes that they have the responsibility to help solve some of the complexes` HLW problems in a cost effective manner by encouraging the M&Os to work together by sharing expertise, eliminating duplicate efforts, and sharing best practices. Pending an action plan, Westinghouse M&Os will take the initiative on those corrective actions identified as the responsibility of an M&O. This document captures issues important to the management of HLW. The proposed resolutions contained within this document set the framework for the M&Os and DOE work cooperatively to develop an action plan to solve some of the major complex-wide problems. Dialogue will continue between the M&Os, DOE, and other regulatory agencies to work jointly toward the goal of storing, treating, and immobilizing HLW for disposal in an environmentally sound, safe, and cost effective manner.

  12. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-99 Status Report

    SciTech Connect (OSTI)

    Herbst, Alan Keith; Mc Cray, John Alan; Kirkham, Robert John; Pao, Jenn Hai; Hinckley, Steve Harold

    1999-10-01T23:59:59.000Z

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1999, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed on radionuclide leaching, microbial degradation, waste neutralization, and a small mockup for grouting the INTEC underground storage tank residual heels.

  13. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-99 Status Report

    SciTech Connect (OSTI)

    A. K. Herbst; J. A. McCray; R. J. Kirkham; J. Pao; S. H. Hinckley

    1999-09-30T23:59:59.000Z

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1999, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed on radionuclide leaching, microbial degradation, waste neutralization, and a small mockup for grouting the INTEC underground storage tank residual heels.

  14. Optimisation of the Management of Higher Activity Waste in the UK - 13537

    SciTech Connect (OSTI)

    Walsh, Ciara; Buckley, Matthew [Nuclear Decommissioning Authority, Building 587, Curie Avenue, Harwell Oxford, Didcot, Oxfordshire, OX11 0RH (United Kingdom)] [Nuclear Decommissioning Authority, Building 587, Curie Avenue, Harwell Oxford, Didcot, Oxfordshire, OX11 0RH (United Kingdom)

    2013-07-01T23:59:59.000Z

    The Upstream Optioneering project was created in the Nuclear Decommissioning Authority (UK) to support the development and implementation of significant opportunities to optimise activities across all the phases of the Higher Activity Waste management life cycle (i.e. retrieval, characterisation, conditioning, packaging, storage, transport and disposal). The objective of the Upstream Optioneering project is to work in conjunction with other functions within NDA and the waste producers to identify and deliver solutions to optimise the management of higher activity waste. Historically, optimisation may have occurred on aspects of the waste life cycle (considered here to include retrieval, conditioning, treatment, packaging, interim storage, transport to final end state, which may be geological disposal). By considering the waste life cycle as a whole, critical analysis of assumed constraints may lead to cost savings for the UK Tax Payer. For example, it may be possible to challenge the requirements for packaging wastes for disposal to deliver an optimised waste life cycle. It is likely that the challenges faced in the UK are shared in other countries. It is therefore likely that the opportunities identified may also apply elsewhere, with the potential for sharing information to enable value to be shared. (authors)

  15. Kinetics of Conversion of High-level Waste to Glass

    SciTech Connect (OSTI)

    Izak, Pavel (ASSOC WESTERN UNIVERSITY); Hrma, Pavel R. (BATTELLE (PACIFIC NW LAB)); Schweiger, Michael J. (BATTELLE (PACIFIC NW LAB)); Heineman, W.R.; Eller, P.G.

    2001-01-01T23:59:59.000Z

    The kinetics of the conversion of high-level waste (HLW) feed to glass controls the rate of HLW processing. Simulated HLW feed and low silica - high sodium (LSHS) feed with co-precipitated Fe, Ni, Cr, and Mn hydroxides (to simulate the chemical and physical makeup of these components in the melter feed) were heated at constant temperature increase rates (0.4, 4, and 14?C/min), quenched at different stages of conversion, and analyzed with optical microscope, scanning electron microscope, and x-ray diffraction (XRD). Quartz, sodium nitrate, carnegieite (Na8Al4Si4O18), sodalite (Na8(AlSiO4)6(NO2)2), and spinel were identified in the samples. Mass fractions of these phases were determined as functions of the temperature and the heating rate. The fractions of nitrates and quartz decreased with increasing temperature, starting above 550?C and dropping to zero at 850?C. Spinel was present in the feed within the temperature interval from 350?C to 1050?C, peaking between 550 and 700?C. Sodalite (in HLW feed) and carnegieite (in LSHS feed) formed at temperatures above 600?C and then began to dissolve. TGA and DSC were use to determine the mass loss and the conversion heat as functions of temperature and heating rate and were compared with the reaction progress reached in quenched samples.

  16. West Valley demonstration project: alternative processes for solidifying the high-level wastes

    SciTech Connect (OSTI)

    Holton, L.K.; Larson, D.E.; Partain, W.L.; Treat, R.L.

    1981-10-01T23:59:59.000Z

    In 1980, the US Department of Energy (DOE) established the West Valley Solidification Project as the result of legislation passed by the US Congress. The purpose of this project was to carry out a high level nuclear waste management demonstration project at the Western New York Nuclear Service Center in West Valley, New York. The DOE authorized the Pacific Northwest Laboratory (PNL), which is operated by Battelle Memorial Institute, to assess alternative processes for treatment and solidification of the WNYNSC high-level wastes. The Process Alternatives Study is the suject of this report. Two pretreatment approaches and several waste form processes were selected for evaluation in this study. The two waste treatment approaches were the salt/sludge separation process and the combined waste process. Both terminal and interim waste form processes were studied.

  17. HIGH LEVEL WASTE TANK CLOSURE PROJECT AT THE IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LABORATORY

    SciTech Connect (OSTI)

    Quigley, K.D.; Wessman, D

    2003-02-27T23:59:59.000Z

    The Department of Energy, Idaho Operations Office (DOE-ID) is in the process of closing two underground high-level waste (HLW) storage tanks at the Idaho National Engineering and Environmental Laboratory (INEEL) to meet Resource Conservation and Recovery Act (RCRA) regulations and Department of Energy orders. Closure of these two tanks is scheduled for 2004 as the first phase in closure of the eleven 1.14 million liter (300,000 gallon) tanks currently in service at the Idaho Nuclear Technology and Engineering Center (INTEC). The INTEC Tank Farm Facility (TFF) Closure sequence consists of multiple steps to be accomplished through the existing tank riser access points. Currently, the tank risers contain steam and process waste lines associated with the steam jets, corrosion coupons, and liquid level indicators. As necessary, this equipment will be removed from the risers to allow adequate space for closure equipment and activities. The basic tank closure sequence is as follows: Empty the tank to the residual heel using the existing jets; Video and sample the heel; Replace steam jets with new jet at a lower position in the tank, and remove additional material; Flush tank, piping and secondary containment with demineralized water; Video and sample the heel; Evaluate decontamination effectiveness; Displace the residual heel with multiple placements of grout; and Grout piping, vaults and remaining tank volume. Design, development, and deployment of a remotely operated tank cleaning system were completed in June 2002. The system incorporates many commercially available components, which have been adapted for application in cleaning high-level waste tanks. The system is cost-effective since it also utilizes existing waste transfer technology (steam jets), to remove tank heel solids from the tank bottoms during the cleaning operations. Remotely operated directional spray nozzles, automatic rotating wash balls, video monitoring equipment, decontamination spray-rings, and tank -specific access interface devices have been integrated to provide a system that efficiently cleans tank walls and heel solids in an acidic, radioactive environment. Through the deployment of the tank cleaning system, the INEEL High Level Waste Program has cleaned tanks to meet RCRA clean closure standards and DOE closure performance measures. Design, development, and testing of tank grouting delivery equipment were completed in October 2002. The system incorporates lessons learned from closures at other DOE facilities. The grout will be used to displace the tank residuals remaining after the cleaning is complete. To maximize heel displacement to the discharge pump, grout was placed in a sequence of five positions utilizing two riser locations. The project is evaluating the use of six positions to optimize the residuals removed. After the heel has been removed and the residuals stabilized, the tank, piping, and secondary containment will be grouted.

  18. STATUS & DIRECTION OF THE BULK VITRIFICATION PROGRAM FOR THE SUPPLEMENTAL TREATMENT OF LOW ACTIVITY TANK WASTE AT HANFORD

    SciTech Connect (OSTI)

    RAYMOND, R.E.

    2005-01-12T23:59:59.000Z

    The DOE Office of River Protection (ORP) is managing a program at the Hanford site that will retrieve and treat more than 200 million liters (53 million gal.) of radioactive waste stored in underground storage tanks. The waste was generated over the past 50 years as part of the nation's defense programs. The project baseline calls for the waste to be retrieved from the tanks and partitioned to separate the highly radioactive constituents from the large volumes of chemical waste. These highly radioactive components will be vitrified into glass logs in the Waste Treatment Plant (WTP), temporarily stored on the Hanford Site, and ultimately disposed of as high-level waste in the offsite national repository. The less radioactive chemical waste, referred to as low-activity waste (LAW), is also planned to be vitrified by the WTP, and then disposed of in approved onsite trenches. However, additional treatment capacity is required in order to complete the pretreatment and immobilization of the tank waste by 2028, which represents a Tri-Party Agreement milestone. To help ensure that the treatment milestones will be met, the Supplemental Treatment Program was undertaken. The program, managed by CH2M HILL Hanford Group, Inc., involves several sub-projects each intended to supplement part of the treatment of waste being designed into the WTP. This includes the testing, evaluation, design, and deployment of supplemental LAW treatment and immobilization technologies, retrieval and treatment of mixed TRU waste stored in the Hanford Tanks, and supplemental pre-treatment. Applying one or more supplemental treatment technologies to the LAW has several advantages, including providing additional processing capacity, reducing the planned loading on the WTP, and reducing the need for double-shell tank space for interim storage of LAW. In fiscal year 2003, three potential supplemental treatment technologies were evaluated including grout, steam reforming and bulk vitrification using AMEC's In-Container Vitrification{trademark} process. As an outcome of this work, the hulk vitrification process was recommended for further evaluation. In fiscal year 2004, a follow-on bulk vitrification project was initiated to design, procure, assemble and operate a full-scale bulk vitrification pilot-plant to treat low activity tank waste from Hanford tank 241-S-109 under a Research, Development and Demonstration permit. That project is referred to as the Demonstration Bulk Vitrification System (or DBVS). The DBVS project will provide a full-scale bulk vitrification demonstration facility that can be used to assess the effectiveness of the bulk vitrification process under actual operating conditions. The pilot-plant is scheduled to commence operations in late 2005. The Supplemental Treatment Program represents a major element of the ORP's strategy to complete the pretreatment and immobilization of tank wastes by 2028. This paper will provide an overview of the bulk vitrification process and the progress in establishing the pilot-plant.

  19. Evaluation of West Valley High-Level Waste Tank Lay-Up Strategies

    SciTech Connect (OSTI)

    McClure, L. W.; Henderson, J. C.; Elmore, M. R.

    2002-02-25T23:59:59.000Z

    The primary objective of the task summarized in this paper was to demonstrate a methodology for evaluating alternative strategies for preclosure lay-up of the two high-level waste (HLW) storage tanks at the West Valley Demonstration Project (WVDP). Lay-up is defined as the period between operational use of tanks for waste storage and final closure. The U.S. Department of Energy (DOE) is planning to separate the environmental impact statement (EIS) for completion of closure of the WVDP into two separate EISs. The first EIS will cover only waste management and decontamination. DOE expects to complete this EIS in about 18 months. The second EIS will cover final decommissioning and closure and may take up to five years to complete. This approach has been proposed to expedite continued management of the waste and decontamination activities in advance of the final EIS and its associated Record of Decision on final site closure. Final closure of the WVDP site may take 10 to 15 years; therefore, the tanks need to be placed in a safe, stable condition with minimum surveillance during an extended lay-up period. The methodology developed for ranking the potential strategies for lay-up of the WVDP tanks can be used to provide a basis for a decision on the preferred path forward. The methodology is also applicable to determining preferred lay-up approaches at other DOE sites. Some of the alternative strategies identified for the WVDP should also be considered for implementation at the other DOE sites. Each site has unique characteristics that would require unique considerations for lay-up.

  20. Evaluation of West Valley High-Level Waste Tank Lay-up Strategies

    SciTech Connect (OSTI)

    Mcclure, Lloyd W.; Henderson, J C.; Elmore, Monte R.

    2002-06-28T23:59:59.000Z

    The primary objective of this task was to demonstrate a methodology for evaluating alternative strategies for preclosure lay-up of the two high-level waste (HLW) storage tanks at the West Valley Demonstration Project (WVDP). Lay-up is defined as the period between operational use of tanks for waste storage and final closure. The U.S. Department of Energy (DOE) is planning to separate the environmental impact statement (EIS) for completion of closure of the WVDP into two separate EISs. The first EIS will cover only waste management and decontamination. DOE expects to complete this EIS in about 18 months. The second EIS will cover final decommissioning and closure, and may take up to five years to complete. This approach has been proposed to expedite continued management of the waste and decontamination activities in advance of the final EIS and Record of Decision on final site closure. Final closure of the WVDP site may take 10 to 15 years. Therefore, the tanks need to be placed in a safe, stable condition with minimum surveillance during an extended lay-up period. The methodology developed for ranking the potential strategies for lay-up of the WVDP tanks can be used to provide a basis for a decision on the preferred path forward. The methodology is also applicable to determining preferred lay-up approaches at other DOE sites. Some of the alternative strategies identified for West Valley should also be considered for implementation at the other sites. Each site has unique characteristics that would require unique considerations for lay-up.

  1. Hanford Site River Protection Project High-Level Waste Safe Storage and Retrieval

    SciTech Connect (OSTI)

    Aromi, E. S.; Raymond, R. E.; Allen, D. I.; Payne, M. A.; DeFigh-Price, C.; Kristofzski, J. G.; Wiegman, S. A.

    2002-02-25T23:59:59.000Z

    This paper provides an update from last year and describes project successes and issues associated with the management and work required to safely store, enhance readiness for waste feed delivery, and prepare for treated waste receipts for the approximately 53 million gallons of mixed and high-level waste currently in aging tanks at the Hanford Site. The Hanford Site is a 560 square-mile area in southeastern Washington State near Richland, Washington.

  2. Integrated High-Level Waste System Planning - Utilizing an Integrated Systems Planning Approach to Ensure End-State Definitions are Met and Executed - 13244

    SciTech Connect (OSTI)

    Ling, Lawrence T. [URS-Savannah River Remediation, Savannah River Site, Building 766-H Room 2205, Aiken, SC 29808 (United States)] [URS-Savannah River Remediation, Savannah River Site, Building 766-H Room 2205, Aiken, SC 29808 (United States); Chew, David P. [URS-Savannah River Remediation, Savannah River Site, Building 766-H Room 2426, Aiken, SC 29808 (United States)] [URS-Savannah River Remediation, Savannah River Site, Building 766-H Room 2426, Aiken, SC 29808 (United States)

    2013-07-01T23:59:59.000Z

    The Savannah River Site (SRS) is a Department of Energy site which has produced nuclear materials for national defense, research, space, and medical programs since the 1950's. As a by-product of this activity, approximately 37 million gallons of high-level liquid waste containing approximately 292 million curies of radioactivity is stored on an interim basis in 45 underground storage tanks. Originally, 51 tanks were constructed and utilized to support the mission. Four tanks have been closed and taken out of service and two are currently undergoing the closure process. The Liquid Waste System is a highly integrated operation involving safely storing liquid waste in underground storage tanks; removing, treating, and dispositioning the low-level waste fraction in grout; vitrifying the higher activity waste at the Defense Waste Processing Facility; and storing the vitrified waste in stainless steel canisters until permanent disposition. After waste removal and processing, the storage and processing facilities are decontaminated and closed. A Liquid Waste System Plan (hereinafter referred to as the Plan) was developed to integrate and document the activities required to disposition legacy and future High-Level Waste and to remove from service radioactive liquid waste tanks and facilities. It establishes and records a planning basis for waste processing in the liquid waste system through the end of the program mission. The integrated Plan which recognizes the challenges of constrained funding provides a path forward to complete the liquid waste mission within all regulatory and legal requirements. The overarching objective of the Plan is to meet all Federal Facility Agreement and Site Treatment Plan regulatory commitments on or ahead of schedule while preserving as much life cycle acceleration as possible through incorporation of numerous cost savings initiatives, elimination of non-essential scope, and deferral of other scope not on the critical path to compliance. There is currently a premium on processing and storage space in the radioactive liquid waste tank system. To enable continuation of risk reduction initiatives, the Plan establishes a processing strategy that provides tank space required to meet, or minimizes the impacts to meeting, programmatic objectives. The Plan also addresses perturbations in funding and schedule impacts. (authors)

  3. Observation and Measurement of Se-79 in SRS High-Level Tank Fission Product Waste

    SciTech Connect (OSTI)

    Dewberry, R.A.

    2000-08-21T23:59:59.000Z

    The authors report the first observation of confirmed Se-79 activity in Savannah River Site high level fission product waste. Se-79 was measured after a seven step chemical treatment to remove interfering activity from Cs-137, Sr-90, and plutonium at levels 105 times higher than the observed Se-79 content and to remove Tc-99 at levels 300 times higher than observed Se-79. Se-79 was measured by liquid scintillation beta-decay counting after specific tests to eliminate uncertainties from possible contributions from Tc-99, Pm-147, Sm-151, Zr-93, or Pu-241, whose beta-decay spectra could appear similar to that of Se-79, and whose content would be expected at levels near or greater than Se-79.

  4. Laboratory Scoping Tests Of Decontamination Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

    SciTech Connect (OSTI)

    Taylor-Pashow, Kathryn M.; Nash, Charles A.; Crawford, Charles L.; McCabe, Daniel J.; Wilmarth, William R.

    2014-01-21T23:59:59.000Z

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task seeks to examine the potential treatment of this stream to remove radionuclides and subsequently disposition the decontaminated stream elsewhere, such as the Effluent Treatment Facility (ETF), for example. The treatment process envisioned is very similar to that used for the Actinide Removal Process (ARP) that has been operating for years at the Savannah River Site (SRS), and focuses on using mature radionuclide removal technologies that are also compatible with longterm tank storage and immobilization methods. For this new application, testing is needed to demonstrate acceptable treatment sorbents and precipitating agents and measure decontamination factors for additional radionuclides in this unique waste stream. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet and will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. One of the radionuclides that is volatile and expected to be in high concentration in this LAW Off-Gas Condensate stream is Technetium-99 ({sup 99}Tc). Technetium will not be removed from the aqueous waste in the Hanford WTP, and will primarily end up immobilized in the LAW glass by repeated recycle of the off-gas condensate into the LAW melter. Other radionuclides that are also expected to be in appreciable concentration in the LAW Off-Gas Condensate are {sup 129}I, {sup 90}Sr, {sup 137}Cs, and {sup 241}Am. This report discusses results of preliminary radionuclide decontamination testing of the simulant. Testing examined use of Monosodium Titanate (MST) to remove {sup 90}Sr and actinides, inorganic reducing agents for {sup 99}Tc, and zeolites for {sup 137}Cs. Test results indicate that excellent removal of {sup 99}Tc was achieved using Sn(II)Cl{sub 2} as a reductant, coupled with sorption onto hydroxyapatite, even in the presence of air and at room temperature. This process was very effective at neutral pH, with a Decontamination Factor (DF) >577 in two hours. It was less effective at alkaline pH. Conversely, removal of the cesium was more effective at alka

  5. Cementitious Grout for Closing SRS High Level Waste Tanks - 12315

    SciTech Connect (OSTI)

    Langton, C.A.; Stefanko, D.B.; Burns, H.H. [Savannah River National Laboratory (United States); Waymer, J.; Mhyre, W.B. [URS Quality and Testing (United States); Herbert, J.E.; Jolly, J.C. Jr. [Savannah River Remediation, LLC, Savannah River Site, Aiken, SC 29808 (United States)

    2012-07-01T23:59:59.000Z

    In 1997, the first two United States Department of Energy (US DOE) high level waste tanks (Tanks 17-F and 20-F: Type IV, single shell tanks) were taken out of service (permanently closed) at the Savannah River Site (SRS). In 2012, the DOE plans to remove from service two additional Savannah River Site (SRS) Type IV high-level waste tanks, Tanks 18-F and 19-F. These tanks were constructed in the late 1950's and received low-heat waste and do not contain cooling coils. Operational closure of Tanks 18-F and 19-F is intended to be consistent with the applicable requirements of the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and will be performed in accordance with South Carolina Department of Health and Environmental Control (SCDHEC). The closure will physically stabilize two 4.92E+04 cubic meter (1.3 E+06 gallon) carbon steel tanks and isolate and stabilize any residual contaminants left in the tanks. Ancillary equipment abandoned in the tanks will also be filled to the extent practical. A Performance Assessment (PA) has been developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closure of the F-Area Tank Farm (FTF) waste tanks. Next generation flowable, zero-bleed cementitious grouts were designed, tested, and specified for closing Tanks 18-F and 19-F and for filling the abandoned equipment. Fill requirements were developed for both the tank and equipment grouts. All grout formulations were required to be alkaline with a pH of 12.4 and to be chemically reducing with a reduction potential (Eh) of -200 to -400. Grouts with this chemistry stabilize potential contaminants of concern. This was achieved by including Portland cement and Grade 100 slag in the mixes, respectively. Ingredients and proportions of cementitious reagents were selected and adjusted to support the mass placement strategy developed by Savannah River Remediation (SRR) Closure Operations. Subsequent down selection was based on compressive strength and saturated hydraulic conductivity results. Fresh slurry property results were used as the first level of screening. A high range water reducing admixture and a viscosity modifying admixture were used to adjust slurry properties to achieve flowable grouts. Adiabatic calorimeter results were used as the second level screening. The third level of screening was used to design mixes that were consistent with the fill material parameters used in the F-Tank Farm Performance Assessment which was developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closures. The cement and slag contents of a mix selected for filling Tanks 18-F and 19-F should be limited to no more than 125 and 210 lbs/cyd, respectively, to limit the heat generated as the result of hydration reaction during curing and thereby enable mass pour placement. Trial mixes with water to total cementitious materials ratios of 0.550 to 0.580 and 125 lbs/cyd of cement and 210 lbs/cyd of slag met the strength and permeability requirements. Mix LP no.8-16 was selected for closing SRS Tanks 18-F and 19-F because it meets or exceeds the design requirements with the least amount of Portland cement and blast furnace slag. This grout is expected to flow at least 45 feet. A single point of discharge should be sufficient for unrestricted flow conditions. However, additional entry points should be identified as back-up in case restrictions in the tank impede flow. The LP no.8 series of trial mixes had surprisingly high design compressive strengths (2000 to 4000/5000 psi) which were achieved at extended curing times (28 to 90 days, respectively) given the small amount of Portland cement in the mixes (100 to 185 lbs/cyd). The grouts were flowable structural fills containing 3/8 inch gravel and concrete sand aggregate. These grouts did not segregate and require no compaction. They have low permeabilities (? 10{sup -9} cm/s) and are consequen

  6. Safety-Related Activities of the IAEA for Radioactive Waste, Decommissioning and Remediation - 13473

    SciTech Connect (OSTI)

    Hahn, Pil-Soo; Vesterlind, Magnus [Division of Radiation, Transport and Waste Safety, International Atomic Energy Agency, PO Box 100, A-1400 Vienna (Austria)] [Division of Radiation, Transport and Waste Safety, International Atomic Energy Agency, PO Box 100, A-1400 Vienna (Austria)

    2013-07-01T23:59:59.000Z

    To fulfil its mandate and serve the needs of its Member States, the IAEA is engaged in a wide range of safety-related activities pertaining to radioactive waste management, decommissioning and remediation. One of the statutory obligations of the IAEA is to establish safety standards and to provide for the application of these standards. The present paper describes recent developments in regard to the IAEA's waste safety standards, and some of the ways the IAEA makes provision for their application. The safety standards and supporting safety demonstration projects seek to establish international consensus on methodologies and approaches for dealing with particular subject areas, for example, safety assessment for radioactive waste disposal. (authors)

  7. X-RAY FLUORESCENCE ANALYSIS OF HANFORD LOW ACTIVITY WASTE SIMULANTS METHOD DEVELOPMENT

    SciTech Connect (OSTI)

    Jurgensen, A; David Missimer, D; Ronny Rutherford, R

    2007-08-08T23:59:59.000Z

    The x-ray fluorescence laboratory (XRF) in the Analytical Development Directorate (ADD) of the Savannah River National Laboratory (SRNL) was requested to develop an x-ray fluorescence spectrometry method for elemental characterization of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) pretreated low activity waste (LAW) stream to the LAW Vitrification Plant. The WTP is evaluating the potential for using XRF as a rapid turnaround technique to support LAW product compliance and glass former batching. The overall objective of this task was to develop an XRF analytical method that provides rapid turnaround time (<8 hours), while providing sufficient accuracy and precision to determine variations in waste.

  8. Preliminary Waste Form Compliance Plan for the Idaho National Engineering and Environmental Laboratory High-Level Waste

    SciTech Connect (OSTI)

    B. A. Staples; T. P. O'Holleran

    1999-05-01T23:59:59.000Z

    The Department of Energy (DOE) has specific technical and documentation requirements for high-level waste (HLW) that is to be placed in a federal repository. This document describes in general terms the strategy to be used at the Idaho National Engineering and Environmental Laboratory (INEEL) to demonstrate that vitrified HLW, if produced at the INEEL, meets these requirements. Waste form, canister, quality assurance, and documentation specifications are discussed. Compliance strategy is given, followed by an overview of how this strategy would be implemented for each specification.

  9. Secondary Waste Cast Stone Waste Form Qualification Testing Plan

    SciTech Connect (OSTI)

    Westsik, Joseph H.; Serne, R. Jeffrey

    2012-09-26T23:59:59.000Z

    The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the 56 million gallons of radioactive waste stored in 177 underground tanks at the Hanford Site. The WTP includes a pretreatment facility to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions for vitrification and disposal. The LAW will be converted to glass for final disposal at the Integrated Disposal Facility (IDF). Cast Stone – a cementitious waste form, has been selected for solidification of this secondary waste stream after treatment in the ETF. The secondary-waste Cast Stone waste form must be acceptable for disposal in the IDF. This secondary waste Cast Stone waste form qualification testing plan outlines the testing of the waste form and immobilization process to demonstrate that the Cast Stone waste form can comply with the disposal requirements. Specifications for the secondary-waste Cast Stone waste form have not been established. For this testing plan, Cast Stone specifications are derived from specifications for the immobilized LAW glass in the WTP contract, the waste acceptance criteria for the IDF, and the waste acceptance criteria in the IDF Permit issued by the State of Washington. This testing plan outlines the testing needed to demonstrate that the waste form can comply with these waste form specifications and acceptance criteria. The testing program must also demonstrate that the immobilization process can be controlled to consistently provide an acceptable waste form product. This testing plan also outlines the testing needed to provide the technical basis for understanding the long-term performance of the waste form in the disposal environment. These waste form performance data are needed to support performance assessment analyses of the long-term environmental impact of the secondary-waste Cast Stone waste form in the IDF

  10. Site Selection and Geological Research Connected with High Level Waste Disposal Programme in the Czech Republic

    SciTech Connect (OSTI)

    Tomas, J.

    2003-02-25T23:59:59.000Z

    Attempts to solve the problem of high-level waste disposal including the spent fuel from nuclear power plants have been made in the Czech Republic for over the 10 years. Already in 1991 the Ministry of Environment entitled The Czech Geological Survey to deal with the siting of the locality for HLW disposal and the project No. 3308 ''The geological research of the safe disposal of high level waste'' had started. Within this project a sub-project ''A selection of perspective HLW disposal sites in the Bohemian Massif'' has been elaborated and 27 prospective areas were identified in the Czech Republic. This selection has been later narrowed to 8 areas which are recently studied in more detail. As a parallel research activity with siting a granitic body Melechov Massif in Central Moldanubian Pluton has been chosen as a test site and the 1st stage of research i.e. evaluation and study of its geological, hydrogeological, geophysical, tectonic and structural properties has been already completed. The Melechov Massif was selected as a test site after the recommendation of WATRP (Waste Management Assessment and Technical Review Programme) mission of IAEA (1993) because it represents an area analogous with the host geological environment for the future HLW and spent fuel disposal in the Czech Republic, i.e. variscan granitoids. It is necessary to say that this site would not be in a locality where the deep repository will be built, although it is a site suitable for oriented research for the sampling and collection of descriptive data using up to date and advanced scientific methods. The Czech Republic HLW and spent fuel disposal programme is now based on The Concept of Radioactive Waste and Spent Nuclear Fuel Management (''Concept'' hereinafter) which has been prepared in compliance with energy policy approved by Government Decree No. 50 of 12th January 2000 and approved by the Government in May 2002. Preparation of the Concept was required, amongst other reasons in connection with preparations for the Czech Republic's accession to the European Union and in connection with the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management adopted under the auspices of the International Atomic Energy Agency, which was signed by the Czech Republic in 1997. According to the approved Concept it is expected that a deep geological repository in the Czech Republic will be built in granitic rocks.

  11. DEVELOPMENT QUALIFICATION AND DISPOSAL OF AN ALTERNATIVE IMMOBILIZED LOW-ACTIVITY WASTE FORM AT THE HANFORD SITE

    SciTech Connect (OSTI)

    SAMS TL; EDGE JA; SWANBERG DJ; ROBBINS RA

    2011-01-13T23:59:59.000Z

    Demonstrating that a waste form produced by a given immobilization process is chemically and physically durable as well as compliant with disposal facility acceptance criteria is critical to the success of a waste treatment program, and must be pursued in conjunction with the maturation of the waste processing technology. Testing of waste forms produced using differing scales of processing units and classes of feeds (simulants versus actual waste) is the crux of the waste form qualification process. Testing is typically focused on leachability of constituents of concern (COCs), as well as chemical and physical durability of the waste form. A principal challenge regarding testing immobilized low-activity waste (ILAW) forms is the absence of a standard test suite or set of mandatory parameters against which waste forms may be tested, compared, and qualified for acceptance in existing and proposed nuclear waste disposal sites at Hanford and across the Department of Energy (DOE) complex. A coherent and widely applicable compliance strategy to support characterization and disposal of new waste forms is essential to enhance and accelerate the remediation of DOE tank waste. This paper provides a background summary of important entities, regulations, and considerations for nuclear waste form qualification and disposal. Against this backdrop, this paper describes a strategy for meeting and demonstrating compliance with disposal requirements emphasizing the River Protection Project (RPP) Integrated Disposal Facility (IDF) at the Hanford Site and the fluidized bed steam reforming (FBSR) mineralized low-activity waste (LAW) product stream.

  12. Yucca Mountain, Nevada - A Proposed Geologic Repository for High-Level Radioactive Waste (Volume 1) Introduction

    SciTech Connect (OSTI)

    R.A. Levich; J.S. Stuckless

    2006-09-25T23:59:59.000Z

    Yucca Mountain in Nevada represents the proposed solution to what has been a lengthy national effort to dispose of high-level radioactive waste, waste which must be isolated from the biosphere for tens of thousands of years. This chapter reviews the background of that national effort and includes some discussion of international work in order to provide a more complete framework for the problem of waste disposal. Other chapters provide the regional geologic setting, the geology of the Yucca Mountain site, the tectonics, and climate (past, present, and future). These last two chapters are integral to prediction of long-term waste isolation.

  13. Annotated bibliography for the design of waste packages for geologic disposal of spent fuel and high-level waste

    SciTech Connect (OSTI)

    Wurm, K.J.; Miller, N.E.

    1982-11-01T23:59:59.000Z

    This bibliography identifies documents that are pertinent to the design of waste packages for geologic disposal of nuclear waste. The bibliography is divided into fourteen subject categories so that anyone wishing to review the subject of leaching, for example, can turn to the leaching section and review the abstracts of reports which are concerned primarily with leaching. Abstracts are also cross referenced according to secondary subject matter so that one can get a complete list of abstracts for any of the fourteen subject categories. All documents which by their title alone appear to deal with the design of waste packages for the geologic disposal of spent fuel or high-level waste were obtained and reviewed. Only those documents which truly appear to be of interest to a waste package designer were abstracted. The documents not abstracted are listed in a separate section. There was no beginning date for consideration of a document for review. About 1100 documents were reviewed and about 450 documents were abstracted.

  14. activity waste simulants: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    KEYWORDS Active Database, Active Rules of Database Management Systems (DBMS) is concerned with an augmentation of DBMS by rules. Passive rules Franek, Frantisek 75...

  15. activity waste simulant: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    KEYWORDS Active Database, Active Rules of Database Management Systems (DBMS) is concerned with an augmentation of DBMS by rules. Passive rules Franek, Frantisek 75...

  16. Glass Property Data and Models for Estimating High-Level Waste Glass Volume

    SciTech Connect (OSTI)

    Vienna, John D.; Fluegel, Alexander; Kim, Dong-Sang; Hrma, Pavel R.

    2009-10-05T23:59:59.000Z

    This report describes recent efforts to develop glass property models that can be used to help estimate the volume of high-level waste (HLW) glass that will result from vitrification of Hanford tank waste. The compositions of acceptable and processable HLW glasses need to be optimized to minimize the waste-form volume and, hence, to save cost. A database of properties and associated compositions for simulated waste glasses was collected for developing property-composition models. This database, although not comprehensive, represents a large fraction of data on waste-glass compositions and properties that were available at the time of this report. Glass property-composition models were fit to subsets of the database for several key glass properties. These models apply to a significantly broader composition space than those previously publised. These models should be considered for interim use in calculating properties of Hanford waste glasses.

  17. A COMPARISON OF HANFORD AND SAVANNAH RIVER SITE HIGH-LEVEL WASTES

    SciTech Connect (OSTI)

    HILL RC PHILIP; REYNOLDS JG; RUTLAND PL

    2011-02-23T23:59:59.000Z

    This study is a simple comparison of high-level waste from plutonium production stored in tanks at the Hanford and Savannah River sites. Savannah River principally used the PUREX process for plutonium separation. Hanford used the PUREX, Bismuth Phosphate, and REDOX processes, and reprocessed many wastes for recovery of uranium and fission products. Thus, Hanford has 55 distinct waste types, only 17 of which could be at Savannah River. While Hanford and Savannah River wastes both have high concentrations of sodium nitrate, caustic, iron, and aluminum, Hanford wastes have higher concentrations of several key constituents. The factors by which average concentrations are higher in Hanford salt waste than in Savannah River waste are 67 for {sup 241}Am, 4 for aluminum, 18 for chromium, 10 for fluoride, 8 for phosphate, 6 for potassium, and 2 for sulfate. The factors by which average concentrations are higher in Hanford sludges than in Savannah River sludges are 3 for chromium, 19 for fluoride, 67 for phosphate, and 6 for zirconium. Waste composition differences must be considered before a waste processing method is selected: A method may be applicable to one site but not to the other.

  18. Waste Form Release Data Package for the 2001 Immobilized Low-Activity Waste Performance Assessment

    SciTech Connect (OSTI)

    McGrail, B. Peter; Icenhower, Jonathan P.; Martin, Paul F.; Schaef, Herbert T.; O'Hara, Matthew J.; Rodriguez, Eugenio; Steele, Jackie L.

    2001-02-01T23:59:59.000Z

    This data package documents the experimentally derived input data on the representative waste glasses LAWABP1 and HLP-31 that will be used for simulations of the immobilized lowactivity waste disposal system with the Subsurface Transport Over Reactive Multiphases (STORM) code. The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in March of 2001. Documented in this data package are data related to 1) kinetic rate law parameters for glass dissolution, 2) alkali-H ion exchange rate, 3) chemical reaction network of secondary phases that form in accelerated weathering tests, and 4) thermodynamic equilibrium constants assigned to these secondary phases. The kinetic rate law and Na+-H+ ion exchange rate were determined from single-pass flow-through experiments. Pressurized unsaturated flow and vapor hydration experiments were used for accelerated weathering or aging of the glasses. The majority of the thermodynamic data were extracted from the thermodynamic database package shipped with the geochemical code EQ3/6. However, several secondary reaction products identified from laboratory tests with prototypical LAW glasses were not included in this database, nor are the thermodynamic data available in the open literature. One of these phases, herschelite, was determined to have a potentially significant impact on the release calculations and so a solubility product was estimated using a polymer structure model developed for zeolites. Although this data package is relatively complete, final selection of ILAW glass compositions has not been done by the waste treatment plant contractor. Consequently, revisions to this data package to address new ILAW glass formulations are to be regularly expected.

  19. High-Level Waste Systems Plan. Revision 7

    SciTech Connect (OSTI)

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

    1996-10-01T23:59:59.000Z

    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.

  20. Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

    DOE Patents [OSTI]

    Boatner, Lynn A. (Oak Ridge, TN); Sales, Brian C. (Oak Ridge, TN)

    1989-01-01T23:59:59.000Z

    Lead-iron phosphate glasses containing a high level of Fe.sub.2 O.sub.3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90.degree. C., with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10.sup.2 to 10.sup.3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe.sub.2 O.sub.3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800.degree. C., since they exhibit very low melt viscosities in the 800.degree. to 1050.degree. C. temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550.degree. C. and are not adversely affected by large doses of gamma radiation in H.sub.2 O at 135.degree. C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms.

  1. Calculations of Induced Activity in the ATLAS Experiment for Nuclear Waste Zoning.

    E-Print Network [OSTI]

    Morev, M N

    2007-01-01T23:59:59.000Z

    Extensive calculations were performed with the general activation formula using the fluxes of high-energy hadrons and low-energy neutrons previously obtained from simulations with the GCALOR code of the ATLAS detector. Three sets of proton cross-sections were used for hadrons energy above 20 MeV: (a) one set calculated with the YIELDX code (i.e., the Silberberg-Tsao formula of partial proton spallation cross-sections), (b) one set calculated with the Rudstam formula, and (c) the â??best-estimate' dataset which was a compilation of the available experimental and calculated data. In the energy region below 20 MeV, neutron activation cross-sections were taken from evaluated nuclear data files. The activity of each nuclide for a predefined operation scenario (i.e., number and duration of irradiation and shutdown cycles) was normalized to reference values taken from the European or Swiss legislations, to obtain an aggregate estimate of the radiological hazard comparable with a nuclear waste zoning definition cr...

  2. Nuclear Waste Fund Activities Management Team | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page onYouTube YouTube Note: Since the.pdfBreaking ofOilNEWResponse to Time-Based Rates fromNuclear Security ConferenceWaste

  3. Development of Crystal-Tolerant High-Level Waste Glasses

    SciTech Connect (OSTI)

    Matyas, Josef; Vienna, John D.; Schaible, Micah J.; Rodriguez, Carmen P.; Crum, Jarrod V.; Arrigoni, Alyssa L.; Tate, Rachel M.

    2010-12-17T23:59:59.000Z

    Twenty five glasses were formulated. They were batched from HLW AZ-101 simulant or raw chemicals and melted and tested with a series of tests to elucidate the effect of spinel-forming components (Ni, Fe, Cr, Mn, and Zn), Al, and noble metals (Rh2O3 and RuO2) on the accumulation rate of spinel crystals in the glass discharge riser of the high-level waste (HLW) melter. In addition, the processing properties of glasses, such as the viscosity and TL, were measured as a function of temperature and composition. Furthermore, the settling of spinel crystals in transparent low-viscosity fluids was studied at room temperature to access the shape factor and hindered settling coefficient of spinel crystals in the Stokes equation. The experimental results suggest that Ni is the most troublesome component of all the studied spinel-forming components producing settling layers of up to 10.5 mm in just 20 days in Ni-rich glasses if noble metals or a higher concentration of Fe was not introduced in the glass. The layer of this thickness can potentially plug the bottom of the riser, preventing glass from being discharged from the melter. The noble metals, Fe, and Al were the components that significantly slowed down or stopped the accumulation of spinel at the bottom. Particles of Rh2O3 and RuO2, hematite and nepheline, acted as nucleation sites significantly increasing the number of crystals and therefore decreasing the average crystal size. The settling rate of ?10-?m crystal size around the settling velocity of crystals was too low to produce thick layers. The experimental data for the thickness of settled layers in the glasses prepared from AZ-101 simulant were used to build a linear empirical model that can predict crystal accumulation in the riser of the melter as a function of concentration of spinel-forming components in glass. The developed model predicts the thicknesses of accumulated layers quite well, R2 = 0.985, and can be become an efficient tool for the formulation of the crystal-tolerant HLW glasses for higher waste loading. A physical modeling effort revealed that the Stokes and Richardson-Zaki equations can be used to adequately predict the accumulation rate of spinel crystals of different sizes and concentrations in the glass discharge riser of HLW melters. The determined shape factor for the glass beads was only 0.73% lower than the theoretical shape factor for a perfect sphere. The shape factor for the spinel crystals matched the theoretically predicted value to within 10% and was smaller than that of the beads, given the larger drag force caused by the larger surface area-to-volume ratio of the octahedral crystals. In the hindered settling experiments, both the glass bead and spinel suspensions were found to follow the predictions of the Richardson-Zaki equation with the exponent n = 3.6 and 2.9 for glass beads and spinel crystals, respectively.

  4. aging high-level waste: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Corrosion Issues in High-level Nuclear Waste Containers. Open Access Theses and Dissertations Summary: ??In this...

  5. Feasibility of lateral emplacement in very deep borehole disposal of high level nuclear waste

    E-Print Network [OSTI]

    Gibbs, Jonathan Sutton

    2010-01-01T23:59:59.000Z

    The U.S. Department of Energy recently filed a motion to withdraw the Nuclear Regulatory Commission license application for the High Level Waste Repository at Yucca Mountain in Nevada. As the U.S. has focused exclusively ...

  6. Design of a high-level waste repository system for the United States

    E-Print Network [OSTI]

    Driscoll, Michael J.

    1988-01-01T23:59:59.000Z

    This report presents a conceptual design for a High Level Waste disposal system for fuel discharged by U.S. commercial power reactors, using the Yucca Mountain repository site recently designated by federal legislation. ...

  7. Risk-informing decisions about high-level nuclear waste repositories

    E-Print Network [OSTI]

    Ghosh, Suchandra Tina, 1973-

    2004-01-01T23:59:59.000Z

    Performance assessments (PAs) are important sources of information for societal decisions in high-level radioactive waste (HLW) management, particularly in evaluating safety cases for proposed HLW repository development. ...

  8. Demonstration of Small Tank Tetraphenylborate Precipitation Process Using Savannah River Site High Level Waste

    SciTech Connect (OSTI)

    Peters, T.B.

    2001-09-10T23:59:59.000Z

    This report details the experimental effort to demonstrate the continuous precipitation of cesium from Savannah River Site High Level Waste using sodium tetraphenylborate. In addition, the experiments examined the removal of strontium and various actinides through addition of monosodium titanate.

  9. Preliminary Technology Maturation Plan for Immobilization of High-Level Waste in Glass Ceramics

    SciTech Connect (OSTI)

    Vienna, John D.; Crum, Jarrod V.; Sevigny, Gary J.; Smith, G L.

    2012-09-30T23:59:59.000Z

    A technology maturation plan (TMP) was developed for immobilization of high-level waste (HLW) raffinate in a glass ceramics waste form using a cold-crucible induction melter (CCIM). The TMP was prepared by the following process: 1) define the reference process and boundaries of the technology being matured, 2) evaluate the technology elements and identify the critical technology elements (CTE), 3) identify the technology readiness level (TRL) of each of the CTE’s using the DOE G 413.3-4, 4) describe the development and demonstration activities required to advance the TRLs to 4 and 6 in order, and 5) prepare a preliminary plan to conduct the development and demonstration. Results of the technology readiness assessment identified five CTE’s and found relatively low TRL’s for each of them: • Mixing, sampling, and analysis of waste slurry and melter feed: TRL-1 • Feeding, melting, and pouring: TRL-1 • Glass ceramic formulation: TRL-1 • Canister cooling and crystallization: TRL-1 • Canister decontamination: TRL-4 Although the TRL’s are low for most of these CTE’s (TRL-1), the effort required to advance them to higher values. The activities required to advance the TRL’s are listed below: • Complete this TMP • Perform a preliminary engineering study • Characterize, estimate, and simulate waste to be treated • Laboratory scale glass ceramic testing • Melter and off-gas testing with simulants • Test the mixing, sampling, and analyses • Canister testing • Decontamination system testing • Issue a requirements document • Issue a risk management document • Complete preliminary design • Integrated pilot testing • Issue a waste compliance plan A preliminary schedule and budget were developed to complete these activities as summarized in the following table (assuming 2012 dollars). TRL Budget Year MSA FMP GCF CCC CD Overall $M 2012 1 1 1 1 4 1 0.3 2013 2 2 1 1 4 1 1.3 2014 2 3 1 1 4 1 1.8 2015 2 3 2 2 4 2 2.6 2016 2 3 2 2 4 2 4.9 2017 2 3 3 2 4 2 9.8 2018 3 3 3 3 4 3 7.9 2019 3 3 3 3 4 3 5.1 2020 3 3 3 3 4 3 14.6 2021 3 3 3 3 4 3 7.3 2022 3 3 3 3 4 3 8.8 2023 4 4 4 4 4 4 9.1 2024 5 5 5 5 5 5 6.9 2025 6 6 6 6 6 6 6.9 CCC = canister cooling and crystallization; FMP = feeding, melting, and pouring; GCF = glass ceramic formulation; MSA = mixing, sampling, and analyses. This TMP is intended to guide the development of the glass ceramics waste form and process to the point where it is ready for industrialization.

  10. Hanford Waste Vitrification Plant Project Waste Form Qualification Program Plan

    SciTech Connect (OSTI)

    Randklev, E.H.

    1993-06-01T23:59:59.000Z

    The US Department of Energy has created a waste acceptance process to help guide the overall program for the disposal of high-level nuclear waste in a federal repository. This Waste Form Qualification Program Plan describes the hierarchy of strategies used by the Hanford Waste Vitrification Plant Project to satisfy the waste form qualification obligations of that waste acceptance process. A description of the functional relationship of the participants contributing to completing this objective is provided. The major activities, products, providers, and associated scheduling for implementing the strategies also are presented.

  11. Annual Summary of Immobilized Low Activity Tank Waste (ILAW) Performance Assessment

    SciTech Connect (OSTI)

    MANN, F M

    2000-05-01T23:59:59.000Z

    As required by the Department of Energy (DOE) order on radioactive waste management (DOE 1999a) as implemented by the Maintenance Plan for the Hanford Immobilized Low-Activity Tank Waste Performance Assessment (Mann 2000a), an annual summary of the adequacy of the Hanford Immobilized Low-Activity Tank Waste Performance Assessment (ILAW PA) must be submitted to DOE headquarters each year that a performance assessment is not submitted. Considering the results of data collection and analysis, the conclusions of the 1998 version of the ILAW PA (Mann 1998) as conditionally approved (DOE 1999b) remain valid, but new information indicates more conservatism in the results than previously estimated. A white paper (Mann 2000b) is attached as Appendix A to justify this statement. Recent ILAW performance estimates used on the waste form and geochemical data have resulted in increased confidence that the disposal of ILAW will meet performance objectives. The ILAW performance assessment program will continue to interact with science and technology activities, disposal facility design staff, and operations, as well as to continue to collect new waste form and disposal system data to further increase the understanding of the impacts of the disposal of ILAW. The next full performance assessment should be issued in the spring of 2001.

  12. Supplement Analysis for the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement

    SciTech Connect (OSTI)

    N /A

    2005-06-30T23:59:59.000Z

    In October 2002, DOE issued the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (Final EIS) (DOE 2002) that provided an analysis of the potential environmental consequences of alternatives/options for the management and disposition of Sodium Bearing Waste (SBW), High-Level Waste (HL W) calcine, and HLW facilities at the Idaho Nuclear Technology and Engineering Center (INTEC) located at the Idaho National Engineering and Environmental Laboratory (INEEL), now known as the Idaho National Laboratory (INL) and referred to hereafter as the Idaho Site. Subsequent to the issuance of the Final EIS, DOE included the requirement for treatment of SBW in the Request for Proposals for Environmental Management activities on the Idaho Site. The new Idaho Cleanup Project (ICP) Contractor identified Steam Reforming as their proposed method to treat SBW; a method analyzed in the Final EIS as an option to treat SBW. The proposed Steam Reforming process for SBW is the same as in the Final EIS for retrieval, treatment process, waste form and transportation for disposal. In addition, DOE has updated the characterization data for both the HLW Calcine (BBWI 2005a) and SBW (BBWI 2004 and BBWI 2005b) and identified two areas where new calculation methods are being used to determine health and safety impacts. Because of those changes, DOE has prepared this supplement analysis to determine whether there are ''substantial changes in the proposed action that are relevant to environmental concerns'' or ''significant new circumstances or information'' within the meaning of the Council of Environmental Quality and DOE National Environmental Policy Act (NEPA) Regulations (40 CFR 1502.9 (c) and 10 CFR 1021.314) that would require preparation of a Supplemental EIS. Specifically, this analysis is intended to determine if: (1) the Steam Reforming Option identified in the Final EIS adequately bounds impacts from the Steam Reforming Process proposed by the new ICP Contractor using the new characterization data, (2) the new characterization data is significantly different than the data presented in the Final EIS, (3) the new calculation methods present a significant change to the impacts described in the Final EIS, and (4) would the updated characterization data cause significant changes in the environmental impacts for the action alternatives/options presented in the Final EIS. There are no other aspects of the Final EIS that require additional review because DOE has not identified any additional new significant circumstances or information that would warrant such a review.

  13. HIGH-LEVEL WASTE FEED CERTIFICATION IN HANFORD DOUBLE-SHELL TANKS

    SciTech Connect (OSTI)

    THIEN MG; WELLS BE; ADAMSON DJ

    2010-01-14T23:59:59.000Z

    The ability to effectively mix, sample, certify, and deliver consistent batches of High Level Waste (HLW) feed from the Hanford Double Shell Tanks (DST) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. DOE's River Protection Project (RPP) mission modeling and WTP facility modeling assume that individual 3785 cubic meter (l million gallon) HLW feed tanks are homogenously mixed, representatively sampled, and consistently delivered to the WTP. It has been demonstrated that homogenous mixing ofHLW sludge in Hanford DSTs is not likely achievable with the baseline design thereby causing representative sampling and consistent feed delivery to be more difficult. Inconsistent feed to the WTP could cause additional batch-to-batch operational adjustments that reduce operating efficiency and have the potential to increase the overall mission length. The Hanford mixing and sampling demonstration program will identify DST mixing performance capability, will evaluate representative sampling techniques, and will estimate feed batch consistency. An evaluation of demonstration program results will identify potential mission improvement considerations that will help ensure successful mission completion. This paper will discuss the history, progress, and future activities that will define and mitigate the mission risk.

  14. activity waste haw: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    system CERN Preprints Summary: 1.1 This test method covers a system that performs nondestructive assay (NDA) of uranium or plutonium, or both, using the active, differential...

  15. activity tank waste: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    system CERN Preprints Summary: 1.1 This test method covers a system that performs nondestructive assay (NDA) of uranium or plutonium, or both, using the active, differential...

  16. New high-level waste management technology for IFR pyroprocessing wastes

    SciTech Connect (OSTI)

    Ackerman, J.P.; Johnson, T.R.

    1993-09-01T23:59:59.000Z

    The pyrochemical electrorefining process for recovery of actinides in spent fuel from the Integral Fast Reactor accumulates fission product wastes as chlorides dissolved in molten LiCI-KCI and as metals, some of which are in molten cadmium. Pyrochemical processes are being developed to recover uranium and transuranium elements for return to the reactor, and to separate and immobilize fission products in suitable waste forms. Solvent cadmium is recycled within the process. Electrolyte salt is treated in a series of salt/cadmium extraction steps; it is also returned to the process. Salt-borne fission products are concentrated on a zeolite bed that is converted to a stable, leach-resistant mineral. Rare earth fission products from the salt, noble metal fission products, and cladding hulls are dispersed in a metal matrix.

  17. Radioactive Demonstration Of Mineralized Waste Forms Made From Hanford Low Activity Waste (Tank Farm Blend) By Fluidized Bed Steam Reformation (FBSR)

    SciTech Connect (OSTI)

    Jantzen, C. M.; Crawford, C. L.; Bannochie, C. J.; Burket, P. R.; Cozzi, A. D.; Daniel, W. E.; Hall, H. K.; Miller, D. H.; Missimer, D. M.; Nash, C. A.; Williams, M. F.

    2013-08-21T23:59:59.000Z

    The U.S. Department of Energy’s Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford’s tank waste. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Supplemental Treatment is likely to be required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP’s LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750°C) continuous method by which LAW can be processed irrespective of whether the waste contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be comparable to LAW glass, i.e. leaches Tc-99, Re and Na at <2g/m2 during ASTM C1285 (Product Consistency) durability testing. Monolithing of the granular FBSR product was investigated to prevent dispersion during transport or burial/storage. Monolithing in an inorganic geopolymer binder, which is amorphous, macro-encapsulates the granules, and the monoliths pass ANSI/ANS 16.1 and ASTM C1308 durability testing with Re achieving a Leach Index (LI) of 9 (the Hanford Integrated Disposal Facility, IDF, criteria for Tc-99) after a few days and Na achieving an LI of >6 (the Hanford IDF criteria for Na) in the first few hours. The granular and monolithic waste forms also pass the EPA Toxicity Characteristic Leaching Procedure (TCLP) for all Resource Conservation and Recovery Act (RCRA) components at the Universal Treatment Standards (UTS). Two identical Benchscale Steam Reformers (BSR) were designed and constructed at SRNL, one to treat non-radioactive simulants and the other to treat actual radioactive wastes. The results from the non-radioactive BSR were used to determine the parameters needed to operate the radioactive BSR in order to confirm the findings of non-radioactive FBSR pilot scale and engineering scale tests and to qualify an FBSR LAW waste form for applications at Hanford. Radioactive testing commenced using SRS LAW from Tank 50 chemically trimmed to look like Hanford’s blended LAW known as the Rassat simulant as this simulant composition had been tested in the non-radioactive BSR, the non-radioactive pilot scale FBSR at the Science Applications International Corporation-Science and Technology Applications Research (SAIC-STAR) facility in Idaho Falls, ID and in the TTT Engineering Scale Technology Demonstration (ESTD) at Hazen Research Inc. (HRI) in Denver, CO. This provided a “tie back” between radioactive BSR testing and non-radioactive BSR, pilot scale, and engineering scale testing. Approximately six hundred grams of non-radioactive and radioactive BSR product were made for extensive testing and comparison to the non-radioactive pilot scale tests performed in 2004 at SAIC-STAR and the engineering scale test performed in 2008 at HRI with the Rassat simulant. The same mineral phases and off-gas species were found in the radioactive and non-radioactive testing. The granular ESTD and BSR products (radioactive and non-radioactive) were analyzed for to

  18. Increasing High-Level Waste Loading In Glass Without Changing The Baseline Melter Technology

    SciTech Connect (OSTI)

    Hrma, Pavel R.; Alton, Jesse; Plaisted, Trevor J.; Klouzek, Jaroslav; Matyas, Josef; Mika, Martin; Schill, Petr; Trochta, Miroslav; Nemec, Lubomir

    2001-02-25T23:59:59.000Z

    The main factors that determine the cost of high-level waste (HLW) vitrification are the waste loading (which determines the volume of glass) and the melting rate. Product quality should be the only factor determining the waste loading while melter design should provide a rapid melting technology. In reality, the current HLW melters are slow in glass-production rate and are subjected to operational risks that require waste loading to be kept far below its intrinsic level. One of the constraints that decrease waste loading is the liquidus-temperature limit. close inspection reveals that this constraint is probably too severe, even for the current technology. The purpose of the liquidus-temperature constraint is to prevent solids from settling on the melter bottom. It appears that some limited settling would niether interfere with melter operation nor shorten its lifetime and that the rate of settling can be greatly reduced if only small crystals are allowed to form.

  19. Overview of Hanford Site High-Level Waste Tank Gas and Vapor Dynamics

    SciTech Connect (OSTI)

    Huckaby, James L.; Mahoney, Lenna A.; Droppo, James G.; Meacham, Joseph E.

    2004-08-31T23:59:59.000Z

    Hanford Site processes associated with the chemical separation of plutonium from uranium and other fission products produced a variety of volatile, semivolatile, and nonvolatile organic and inorganic waste chemicals that were sent to high-level waste tanks. These chemicals have undergone and continue to undergo radiolytic and thermal reactions in the tanks to produce a wide variety of degradation reaction products. The origins of the organic wastes, the chemical reactions they undergo, and their reaction products have recently been examined by Stock (2004). Stock gives particular attention to explaining the presence of various types of volatile and semivolatile organic species identified in headspace air samples. This report complements the Stock report by examining the storage of volatile and semivolatile species in the waste, their transport through any overburden of waste to the tank headspaces, the physical phenomena affecting their concentrations in the headspaces, and their eventual release into the atmosphere above the tanks.

  20. Liquidus Temperature and Primary Crystallization Phases in High-Zirconia High-Level Waste Borosilicate Glasses

    SciTech Connect (OSTI)

    Plaisted, Trevor J.; Hrma, Pavel R.; Vienna, John D.; Jiricka, Antonin

    1999-12-09T23:59:59.000Z

    Liquidus temperature (TL) studies of high-Zr high-level waste (HLW) borosilicate glasses have identified three primary phases: baddelyite (ZrO2), zircon (ZrSiO4), and alkali-zirconium silicates, such as parakeldyshite (Na2ZrSi2O7). Using published TL data for HLW glasses with these primary phases, we have computed partial specific TLs for major glass components. On the Na2O-SiO2-ZrO2 submixture, we have determined approximate positions of the boundaries between the baddelyite, zircon, and parakeldyshite primary phase fields. The maximum that can dissolve at 1150?C in a borosilicate HLW glass subjected to common processability and acceptability constraints appears to be 16.5 mass% ZrO2.

  1. Hanford Site Secondary Waste Roadmap

    SciTech Connect (OSTI)

    Westsik, Joseph H.

    2009-01-29T23:59:59.000Z

    Summary The U.S. Department of Energy (DOE) is making plans to dispose of 54 million gallons of radioactive tank wastes at the Hanford Site near Richland, Washington. The high-level wastes and low-activity wastes will be vitrified and placed in permanent disposal sites. Processing of the tank wastes will generate secondary wastes, including routine solid wastes and liquid process effluents, and these need to be processed and disposed of also. The Department of Energy Office of Waste Processing sponsored a meeting to develop a roadmap to outline the steps necessary to design the secondary waste forms. Representatives from DOE, the U.S. Environmental Protection Agency, the Washington State Department of Ecology, the Oregon Department of Energy, Nuclear Regulatory Commission, technical experts from the DOE national laboratories, academia, and private consultants convened in Richland, Washington, during the week of July 21-23, 2008, to participate in a workshop to identify the risks and uncertainties associated with the treatment and disposal of the secondary wastes and to develop a roadmap for addressing those risks and uncertainties. This report describes the results of the roadmap meeting in Richland. Processing of the tank wastes will generate secondary wastes, including routine solid wastes and liquid process effluents. The secondary waste roadmap workshop focused on the waste streams that contained the largest fractions of the 129I and 99Tc that the Integrated Disposal Facility risk assessment analyses were showing to have the largest contribution to the estimated IDF disposal impacts to groundwater. Thus, the roadmapping effort was to focus on the scrubber/off-gas treatment liquids with 99Tc to be sent to the Effluent Treatment Facility for treatment and solidification and the silver mordenite and carbon beds with the captured 129I to be packaged and sent to the IDF. At the highest level, the secondary waste roadmap includes elements addressing regulatory and performance requirements, waste composition, preliminary waste form screening, waste form development, process design and support, and validation. The regulatory and performance requirements activity will provide the secondary waste-form performance requirements. The waste-composition activity will provide workable ranges of secondary waste compositions and formulations for simulants and surrogates. Preliminary waste form screening will identify candidate waste forms for immobilizing the secondary wastes. The waste form development activity will mature the waste forms, leading to a selected waste form(s) with a defensible understanding of the long-term release rate and input into the critical decision process for a secondary waste treatment process/facility. The process and design support activity will provide a reliable process flowsheet and input to support a robust facility design. The validation effort will confirm that the selected waste form meets regulatory requirements. The final outcome of the implementation of the secondary waste roadmap is the compliant, effective, timely, and cost-effective disposal of the secondary wastes. The work necessary to address the programmatic, regulatory, and technical risks and uncertainties identified through the Secondary Waste Roadmap Workshop are assembled into several program needs elements. Programmatic/Regulatory needs include: • Select and deploy Hanford tank waste supplemental treatment technology • Provide treatment capability for secondary waste streams from tank waste treatment • Develop consensus on secondary waste form acceptance. Technology needs include: • Define secondary waste composition ranges and uncertainties • Identify and develop waste forms for secondary waste immobilization and disposal • Develop test methods to characterize secondary waste form performance. Details for each of these program elements are provided.

  2. DESIGN OF THE DEMOSNTRATION BULK VITRIFICATION SYSTEM FOR THE SUPPLEMENTAL TREATMENT OF LOW ACTIVITY TANK WASTE AT HANFORD

    SciTech Connect (OSTI)

    VAN BEEK JE

    2008-02-14T23:59:59.000Z

    In June 2004, the Demonstration Bulk Vitrification System (DBVS) was initiated with the intent to design, construct, and operate a full-scale bulk vitrification pilot-plant to treat low-activity tank waste from Hanford Tank 241-S-109. The DBVS facility uses In-Container Vitrification{trademark} (ICV{trademark}) at the core of the treatment process. The basic process steps combine liquid low-activity waste (LAW) and glassformers; dry the mixture; and then vitrify the mixture in a batch feed-while-melt process in a refractory lined steel container. Off-gases are processed through a state-of-the-art air pollution control system including sintered-metal filtration, thermal oxidation, acid gas scrubbing, and high-efficiency particulate air (HEPA) and high-efficiency gas adsorber (HEGA) filtration. Testing has focused on development and validation of the waste dryer, ICV, and sintered-metal filters (SMFs) equipment, operations enhancements, and glass formulation. With a parallel testing and design process, testing has allowed improvements to the DBVS equipment configuration and operating methodology, since its original inception. Design improvements include optimization of refractory panels in the ICV, simplifying glassformer addition equipment, increasing the number of waste feed chutes to the ICV, and adding capability for remote clean-out of piping, In addition, the U.S. Department of Energy (DOE) has provided an independent review of the entire DBVS process. While the review did not find any fatal flaws, some technical issues were identified that required a re-evaluation of the DBVS design and subsequent changes to the design. A 100 percent design package for the pilot plant will be completed and submitted to DOE for review in early 2008 that incorporates process improvements substantiated through testing and reviews. This paper provides a description of the bulk vitrification process and a discussion of major equipment design changes that have occurred based on full-scale testing over the past two years and DOE reviews.

  3. Preliminary Closure Plan for the Immobilized Low Activity Waste (ILAW) Disposal Facility

    SciTech Connect (OSTI)

    BURBANK, D.A.

    2000-08-31T23:59:59.000Z

    This document describes the preliminary plans for closure of the Immobilized Low-Activity Waste (ILAW) disposal facility to be built by the Office of River Protection at the Hanford site in southeastern Washington. The facility will provide near-surface disposal of up to 204,000 cubic meters of ILAW in engineered trenches with modified RCRA Subtitle C closure barriers.

  4. Investigation of EPS Characteristics and their Effects on Waste Activated Sludge Digestion

    E-Print Network [OSTI]

    Mountziaris, T. J.

    Investigation of EPS Characteristics and their Effects on Waste Activated Sludge Digestion Thomas be accomplished through use of aerobic or anaerobic self digestion, but choice of digestion type in practice on digestibility, especially in regards to the composition of extracellular polymeric substances (EPS). Samples

  5. Impact of EPS on Digestion of Waste Activate Sludge Thomas Gostanian

    E-Print Network [OSTI]

    Mountziaris, T. J.

    Impact of EPS on Digestion of Waste Activate Sludge Thomas Gostanian Faculty Mentor: Professor Chul Park, Civil & Environmental Engineering Digestion by bacteria is a widely used method for organic is by either aerobic or anaerobic self-digestion, in which the bacteria consume their own mass. Currently

  6. Annual Summary of Immobilized Low Activity Tank Waste (ILAW) Performance Assessment for 2002

    SciTech Connect (OSTI)

    MANN, F M

    2002-08-01T23:59:59.000Z

    As required by the Department of Energy ( DOE), an annual summary of the adequacy of the Hanford Immobilized Low-Activity Tank Waste Performance Assessment (ILAW PA) is necessary in each year in which a full performance assessment is not issued.

  7. Spacecraft Habitation Systems, Water Recovery and Waste Active Charged Particle and Neutron Radiation Measurement

    E-Print Network [OSTI]

    SBIR SBIR 62 63 I Spacecraft Habitation Systems, Water Recovery and Waste Management Active Charged indicate that secondary neutrons, with energies ranging between 0.5 to >150 MeV, make a significant discriminate between the fraction of dose, which results from secondary neutrons, and that which results from

  8. CAST STONE TECHNOLOGY FOR THE TREATMENT AND IMMOBILIZATION OF LOW-ACTIVITY WASTE

    SciTech Connect (OSTI)

    MINWALL HJ

    2011-04-08T23:59:59.000Z

    Cast stone technology is being evaluated for potential application in the treatment and immobilization of Hanford low-activity waste. The purpose of this document is to provide background information on cast stone technology. The information provided in the report is mainly based on a pre-conceptual design completed in 2003.

  9. Geochemical evolution of highly alkaline and saline tank waste plumes during seepage through vadose zone sediments

    SciTech Connect (OSTI)

    Wan, Jiamin; Tokunaga, Tetsu K.; Larsen, Joern T.; Serne, R JEFFREY.

    2004-02-01T23:59:59.000Z

    Leakage of highly saline and alkaline radioactive waste from storage tanks into underlying sediments is a serious environmental problem at the Hanford Site in Washington State. This study focuses on geochemical evolution of tank waste plumes resulting from interactions between the waste solution and sediment. A synthetic tank waste solution was infused into unsaturated Hanford sediment columns (0.2, 0.6, and 2 m) maintained at 70C to simulate the field contamination process. Spatially and temporally resolved geochemical profiles of the waste plume were obtained. Thorough OH neutralization (from an initial pH 14 down to 6.3) was observed. Three broad zones of pore solutions were identified to categorize the dominant geochemical reactions: the silicate dissolution zone (pH > 10), pH-neutralized zone (pH 10 to 6.5), and displaced native sediment pore water (pH 6.5 to 8). Elevated concentrations of Si, Fe, and K in plume fluids and their depleted concentrations in plume sediments reflected dissolution of primary minerals within the silicate dissolution zone. The very high Na concentrations in the waste solution resulted in rapid and complete cation exchange, reflected in high concentrations of Ca and Mg at the plume front. The plume-sediment profiles also showed deposition of hydrated solids and carbonates. Fair correspondence was obtained between these results and analyses of field borehole samples from a waste plume at the Hanford Site. Results of this study provide a well-defined framework for understanding waste plumes in the more complex field setting and for understanding geochemical factors controlling transport of contaminant species carried in waste solutions that leaked from single-shell storage tanks in the past.

  10. Vitrification and testing of a Hanford high-level waste sample, Part 2: Phase identification and waste form leachability

    SciTech Connect (OSTI)

    Hrma, Pavel R.; Crum, Jarrod V.; Bredt, Paul; Greenwood, Lawrence R.; Smith, H D.

    2005-10-01T23:59:59.000Z

    A sample of Hanford high-level radioactive waste from Tank AZ-101 was vitrified into borosilicate glass and tested to demonstrate its compliance with regulatory requirements. Compositional aspects of this study were reported in Part 1 of this paper. This second and last part presents results of crystallinity and leachability testing. Crystallinity was quantified in a glass sample heat treated according to the cooling curve of glass at the centerline of a Hanford Waste Treatment Plant canister. By quantitative X-ray diffraction analysis and image analysis applied to scanning electron microscopy micrographs, the sample contained 7 mass% of spinel, predominantly trevorite. Glass leachability was measured with the product consistency test and the toxicity characteristic leaching procedure. Measured data and model estimates were in reasonable agreement. Leachability results were close to those obtained for the nonradioactive simulant. Models were used to elucidate the effects of glass composition of spinel formation and to estimate effects of spinel formation on glass leachability.

  11. International fuel cycle and waste management technology exchange activities sponsored by the United States Department of Energy: FY 1982 evaluation report

    SciTech Connect (OSTI)

    Lakey, L.T.; Harmon, K.M.

    1983-02-01T23:59:59.000Z

    In FY 1982, DOE and DOE contractor personnel attended 40 international symposia and conferences on fuel reprocessing and waste management subjects. The treatment of high-level waste was the topic most often covered in the visits, with geologic disposal and general waste management also being covered in numerous visits. Topics discussed less frequently inlcude TRU/LLW treatment, airborne waste treatment, D and D, spent fuel handling, and transportation. The benefits accuring to the US from technology exchange activities with other countries are both tangible, e.g., design of equipment, and intangible, e.g., improved foreign relations. New concepts initiated in other countries, particularly those with sizable nuclear programs, are beginning to appear in US efforts in growing numbers. The spent fuel dry storage concept originating in the FRG is being considered at numerous sites. Similarly, the German handling and draining concepts for the joule-heated ceramic melter used to vitrify wastes are being incorporated in US designs. Other foreigh technologies applicable in the US include the slagging incinerator (Belgium), the SYNROC waste form (Australia), the decontamination experience gained in decommissioning the Eurochemic reprocessing plant (Belgium), the engineered surface storage of low- and intermediate-level waste (Belgium, FRG, France), the air-cooled storage of vitrified high-level waste (France, UK), waste packaging (Canada, FRG, Sweden), disposal in salt (FRG), disposal in granite (Canada, Sweden), and sea dumping (UK, Belgium, The Netherlands, Switzerland). These technologies did not necessarily originated or have been tried in the US but for various reasons are now being applied and extended in other countries. This growing nuclear technological base in other countires reduces the number of technology avenues the US need follow to develop a solid nuclear power program.

  12. Greater-than-Class C low-level radioactive waste characterization: Estimated volumes, radionuclide activities, and other characteristics

    SciTech Connect (OSTI)

    Hulse, R.A.

    1991-08-01T23:59:59.000Z

    Planning for storage or disposal of greater-than-Class C low-level radioactive waste (GTCC LLW) requires characterization of that waste to estimate volumes, radionuclide activities, and waste forms. Data from existing literature, disposal records, and original research were used to estimate the characteristics and project volumes and radionuclide activities to the year 2035. GTCC LLW is categorized as: nuclear utilities waste, sealed sources waste, DOE-held potential GTCC LLW; and, other generator waste. It has been determined that the largest volume of those wastes, approximately 57%, is generated by nuclear power plants. The Other Generator waste category contributes approximately 10% of the total GTCC LLW volume projected to the year 2035. Waste held by the Department of Energy, which is potential GTCC LLW, accounts for nearly 33% of all waste projected to the year 2035; however, no disposal determination has been made for that waste. Sealed sources are less than 0.2% of the total projected volume of GTCC LLW.

  13. Idaho High-Level Waste & Facilities Disposition, Final Environmental Impact Statement

    SciTech Connect (OSTI)

    N /A

    2002-10-11T23:59:59.000Z

    This EIS analyzes the potential environmental consequences of alternatives for managing high-level waste (HLW) calcine, mixed transuranic waste/sodium bearing waste (SBW) and newly generated liquid waste at the Idaho National Engineering and Environmental Laboratory (INEEL) in liquid and solid forms. This EIS also analyzes alternatives for the final disposition of HLW management facilities at the INEEL after their missions are completed. After considering comments on the Draft EIS (DOE/EIS-0287D), as well as information on available treatment technologies, DOE and the State of Idaho have identified separate preferred alternatives for waste treatment. DOE's preferred alternative for waste treatment is performance based with the focus on placing the wastes in forms suitable for disposal. Technologies available to meet the performance objectives may be chosen from the action alternatives analyzed in this EIS. The State of Idaho's Preferred Alternative for treating mixed transuranic waste/SBW and calcine is vitrification, with or without calcine separations. Under both the DOE and State of Idaho preferred alternatives, newly generated liquid waste would be segregated after 2005, stored or treated directly and disposed of as low-level, mixed low-level, or transuranic waste depending on its characteristics. The objective of each preferred alternative is to enable compliance with the legal requirement to have INEEL HLW road ready by a target date of 2035. Both DOE and the State of Idaho have identified the same preferred alternative for facilities disposition, which is to use performance-based closure methods for existing facilities and to design new facilities consistent with clean closure methods.

  14. The dilemma of siting a high-level nuclear waste repository

    SciTech Connect (OSTI)

    Easterline, D.; Kunreuther, H.

    1995-12-31T23:59:59.000Z

    This books presents a siting process that the authors believe will prove successful within the adversarial world that characterizes most attempts to build waste-disposal facilities. They come to the following conclusions: a volunatary siting process stands the best chance of breaking the `not-in-my-backyard` problem; and without public acknowledgement that a facility is needed, any proposal to build a high-level nuclear waste storage facility will meet with opposition.

  15. Separation and Purification and Beta Liquid Scintillation Analysis of Sm-151 in Savannah River Site and Hanford Site DOE High Level Waste

    SciTech Connect (OSTI)

    Dewberry, R.A.

    2001-02-13T23:59:59.000Z

    This paper describes development work to obtain a product phase of Sm-151 pure of any other radioactive species so that it can be determined in US Department of Energy high level liquid waste and low level solid waste by liquid scintillation {beta}-spectroscopy. The technique provides separation from {mu}Ci/ml levels of Cs-137, Pu alpha and Pu-241 {beta}-decay activity, and Sr-90/Y-90 activity. The separation technique is also demonstrated to be useful for the determination of Pm-147.

  16. LABORATORY OPTIMIZATION TESTS OF TECHNETIUM DECONTAMINATION OF HANFORD WASTE TREATMENT PLANT LOW ACTIVITY WASTE OFF-GAS CONDENSATE SIMULANT

    SciTech Connect (OSTI)

    Taylor-Pashow, K.; Nash, C.; McCabe, D.

    2014-09-29T23:59:59.000Z

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task examines the potential treatment of this stream to remove radionuclides and subsequently disposition the decontaminated stream elsewhere, such as the Effluent Treatment Facility (ETF), for example. The treatment process envisioned is very similar to that used for the Actinide Removal Process (ARP) that has been operating for years at the Savannah River Site (SRS), and focuses on using mature radionuclide removal technologies that are also compatible with longterm tank storage and immobilization methods. For this new application, testing is needed to demonstrate acceptable treatment sorbents and precipitating agents and measure decontamination factors for additional radionuclides in this unique waste stream. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet and will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. One of the radionuclides that is volatile and expected to be in greatest abundance in this LAW Off-Gas Condensate stream is Technetium-99 ({sup 99}Tc). Technetium will not be removed from the aqueous waste in the Hanford WTP, and will primarily end up immobilized in the LAW glass by repeated recycle of the off-gas condensate into the LAW melter. Other radionuclides that are low but are also expected to be in measurable concentration in the LAW Off-Gas Condensate are {sup 129}I, {sup 90}Sr, {sup 137}Cs, {sup 241}Pu, and {sup 241}Am. These are present due to their partial volatility and some entrainment in the off-gas system. This report discusses results of optimized {sup 99}Tc decontamination testing of the simulant. Testing examined use of inorganic reducing agents for {sup 99}Tc. Testing focused on minimizing the quantity of sorbents/reactants added, and minimizing mixing time to reach the decontamination targets in this simulant formulation. Stannous chloride and ferrous sulfate were tested as reducing agents to determine the minimum needed to convert soluble pertechnetate to the insoluble technetium dioxide. The reducing agents were tried with and without sorbents.

  17. Summary Of Cold Crucible Vitrification Tests Results With Savannah River Site High Level Waste Surrogates

    SciTech Connect (OSTI)

    Stefanovsky, Sergey; Marra, James; Lebedev, Vladimir

    2014-01-13T23:59:59.000Z

    The cold crucible inductive melting (CCIM) technology successfully applied for vitrification of low- and intermediate-level waste (LILW) at SIA Radon, Russia, was tested to be implemented for vitrification of high-level waste (HLW) stored at Savannah River Site, USA. Mixtures of Sludge Batch 2 (SB2) and 4 (SB4) waste surrogates and borosilicate frits as slurries were vitrified in bench- (236 mm inner diameter) and full-scale (418 mm inner diameter) cold crucibles. Various process conditions were tested and major process variables were determined. Melts were poured into 10L canisters and cooled to room temperature in air or in heat-insulated boxes by a regime similar to Canister Centerline Cooling (CCC) used at DWPF. The products with waste loading from ~40 to ~65 wt.% were investigated in details. The products contained 40 to 55 wt.% waste oxides were predominantly amorphous; at higher waste loadings (WL) spinel structure phases and nepheline were present. Normalized release values for Li, B, Na, and Si determined by PCT procedure remain lower than those from EA glass at waste loadings of up to 60 wt.%.

  18. Unrestricted disposal of minimal activity levels of radioactive wastes: exposure and risk calculations

    SciTech Connect (OSTI)

    Fields, D.E.; Emerson, C.J.

    1984-08-01T23:59:59.000Z

    The US Nuclear Regulatory Commission is currently considering revision of rule 10 CFR Part 20, which covers disposal of solid wastes containing minimal radioactivity. In support of these revised rules, we have evaluated the consequences of disposing of four waste streams at four types of disposal areas located in three different geographic regions. Consequences are expressed in terms of human exposures and associated health effects. Each geographic region has its own climate and geology. Example waste streams, waste disposal methods, and geographic regions chosen for this study are clearly specified. Monetary consequences of minimal activity waste disposal are briefly discussed. The PRESTO methodology was used to evaluate radionuclide transport and health effects. This methodology was developed to assess radiological impacts to a static local population for a 1000-year period following disposal. Pathways and processes of transit from the trench to exposed populations included the following considerations: groundwater transport, overland flow, erosion, surface water dilution, resuspension, atmospheric transport, deposition, inhalation, and ingestion of contaminated beef, milk, crops, and water. 12 references, 2 figures, 8 tables.

  19. Radioactive waste from transmutation of technetium: a model for anticipating characteristics of high level waste from transmutation

    SciTech Connect (OSTI)

    Seitz, M.G. [Booz Allen Hamilton, Washington DC (United States)

    2007-07-01T23:59:59.000Z

    At this early stage in the conceptualization of fuel treatment and radioisotope transmutation for the disposition of nuclear wastes, it is possible to anticipate some characteristics of the waste stream resulting from the deployment of advanced technologies. Fission products and actinides cannot be completely destroyed by transmutation even with continuous purification and recycle. This is demonstrated for technetium in this analysis, but is true for all radioisotopes. Also, some of the reaction products are themselves long-lived radioactive isotopes. The purification and recycle steps produce nuclear wastes that must be planned for geologic disposal. Five radioisotopes have been identified to be produced in abundance by transmutation of technetium using fast neutrons. Four of these isotopes may be more benign than the original technetium-99 because of their longer half lives. However, one isotope, molybdenum-93 with a half life of four thousand years, may be troublesome. All of the isotopes arising from the transmutation process that end up in high level waste must be examined in terms of their behavior in geologic disposal. In selecting goals for chemical separations, the technologists must consider the entire cycle of separation and transmutation before applying the performance expected in a single separation to implications concerning a repository. A separation efficiency of 0.95 can translate into the disposal of as much as 30 to 60 percent of the technetium in the repository if down stream losses are not controlled. In this case, the treatment may have little impact on anticipated off site radiation from technetium. The destruction of technetium through continuous recycle requires the cost of increased neutron dose and increased space in reactors that must be considered in design of fuel treatment systems. (authors)

  20. Steam reforming utilizing high activity catalyst

    SciTech Connect (OSTI)

    Setzer, H. J.

    1985-03-05T23:59:59.000Z

    High activity, sulfur tolerant steam reforming catalysts are described comprising rhodium or nickel supported on lanthanum stabilized alumina or magnesium promoted lanthanum stabilized alumina. The catalysts have improved activity over conventionally used catalysts in the presence of sulfur containing hydrocarbon fuel (such as No. 2 fuel oil) in a steam reforming environment. The material has particular utility in autothermal, tubular, cyclic and adiabatic steam reforming processes.

  1. MINERALIZING, STEAM REFORMING TREATMENT OF HANFORD LOW-ACTIVITY WASTE (a.k.a. INEEL/EXT-05-02526)

    SciTech Connect (OSTI)

    A. L. Olson; N. R. Soelberg; D. W. Marshall; G. L. Anderson

    2005-02-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) documented, in 2002, a plan for accelerating cleanup of the Hanford Site, located in southeastern Washington State, by at least 35 years. A key element of the plan was acceleration of the tank waste program and completion of ''tank waste treatment by 2028 by increasing the capacity of the planned Waste Treatment Plant (WTP) and using supplemental technologies for waste treatment and immobilization.'' The plan identified steam reforming technology as a candidate for supplemental treatment of as much as 70% of the low-activity waste (LAW). Mineralizing steam reforming technology, offered by THOR Treatment Technologies, LLC would produce a denitrated, granular mineral waste form using a high-temperature fluidized bed process. A pilot scale demonstration of the technology was completed in a 15-cm-diameter reactor vessel. The pilot scale facility was equipped with a cyclone separator and heated sintered metal filters for particulate removal, a thermal oxidizer for reduced gas species and NOx destruction, and a packed activated carbon bed for residual volatile species capture. The pilot scale equipment is owned by the DOE, but located at the Science and Technology Applications Research (STAR) Center in Idaho Falls, ID. Pilot scale testing was performed August 2–5, 2004. Flowsheet chemistry and operational parameters were defined through a collaborative effort involving Idaho National Engineering and Environmental Laboratory (INEEL), Savannah River National Laboratory (SRNL), and THOR Treatment Technologies personnel. Science Application International Corporation, owners of the STAR Center, personnel performed actual pilot scale operation. The pilot scale test achieved a total of 68.4 hours of cumulative/continuous processing operation before termination in response to a bed de-fluidization condition. 178 kg of LAW surrogate were processed that resulted in 148 kg of solid product, a mass reduction of about 17%. The process achieved essentially complete bed turnover within approximately 40 hours. Samples of mineralized solid product materials were analyzed for chemical/physical properties. SRNL will report separately the results of product performance testing that were accomplished.

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

    SciTech Connect (OSTI)

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

    2008-01-23T23:59:59.000Z

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

  3. Development Of High Waste-Loading HLW Glasses For High Bismuth Phosphate Wastes, VSL-12R2550-1, Rev 0

    SciTech Connect (OSTI)

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Gan, Hao [The Catholic University of America, Washington, DC (United States); Kot, Wing K. [The Catholic University of America, Washington, DC (United States)

    2012-12-13T23:59:59.000Z

    This report presents results from tests with new glass formulations that have been developed for several high Bi-P HLW compositions that are expected to be processed at the WTP that have not been tested previously. WTP HLW feed compositions were reviewed to select waste batches that are high in Bi-P and that are reasonably distinct from the Bi-limited waste that has been tested previously. Three such high Bi-P HLW compositions were selected for this work. The focus of the present work was to determine whether the same type of issues as seen in previous work with high-Bi HLW will be seen in HLW with different concentrations of Bi, P and Cr and also whether similar glass formulation development approaches would be successful in mitigating these issues. New glass compositions were developed for each of the three representative Bi-P HLW wastes and characterized with respect to key processing and product quality properties and, in particular, those relating to crystallization and foaming tendency.

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

    SciTech Connect (OSTI)

    Pavel Hrma; Pert Schill; Lubomir Nemec

    2002-01-07T23:59:59.000Z

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

  5. Settling of Spinel in a High-Level Waste Glass Melter

    SciTech Connect (OSTI)

    Hrma, Pavel R.; Schill, Pert; Nemec, Lubomir

    2002-01-18T23:59:59.000Z

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

  6. Operating experience during high-level waste vitrification at the West Valley Demonstration Project

    SciTech Connect (OSTI)

    Valenti, P.J.; Elliott, D.I.

    1999-01-01T23:59:59.000Z

    This report provides a summary of operational experiences, component and system performance, and lessons learned associated with the operation of the Vitrification Facility (VF) at the West Valley Demonstration Project (WVDP). The VF was designed to convert stored high-level radioactive waste (HLW) into a stable waste form (borosilicate glass) suitable for disposal in a federal repository. Following successful completion on nonradioactive test, HLW processing began in July 1995. Completion of Phase 1 of HLW processing was reached on 10 June 1998 and represented the processing of 9.32 million curies of cesium-137 (Cs-137) and strontium-90 (Sr-90) to fill 211 canisters with over 436,000 kilograms of glass. With approximately 85% of the total estimated curie content removed from underground waste storage tanks during Phase 1, subsequent operations will focus on removal of tank heel wastes.

  7. Technical Exchange on Improved Design and Performance of High Level Waste Melters - Final Report

    SciTech Connect (OSTI)

    SK Sundaram; ML Elliott; D Bickford

    1999-11-19T23:59:59.000Z

    SIA Radon is responsible for management of low- and intermediate-level radioactive waste (LILW) produced in Central Russia. In cooperation with Minatom organizations Radon carries out R and D programs on treatment of simulated high level waste (HLW) as well. Radon scientists deal with a study of materials for LILW, HLW, and Nuclear Power Plants (NPP) wastes immobilization, and development and testing of processes and technologies for waste treatment and disposal. Radon is mostly experienced in LILW vitrification. This experience can be carried over to HLW vitrification especially in field of melting systems. The melter chosen as a basic unit for the vitrification plant is a cold crucible. Later on Radon experience in LILW vitrification as well as our results on simulated HLW vitrification are briefly described.

  8. Alternatives for high-level waste forms, containers, and container processing systems

    SciTech Connect (OSTI)

    Crawford, T.W.

    1995-09-22T23:59:59.000Z

    This study evaluates alternatives for high-level waste forms, containers, container processing systems, and onsite interim storage. Glass waste forms considered are cullet, marbles, gems, and monolithic glass. Small and large containers configured with several combinations of overpack confinement and shield casks are evaluated for these waste forms. Onsite interim storage concepts including canister storage building, bore holes, and storage pad were configured with various glass forms and canister alternatives. All favorable options include the monolithic glass production process as the waste form. Of the favorable options the unshielded 4- and 7-canister overpack options have the greatest technical assurance associated with their design concepts due to their process packaging and storage methods. These canisters are 0.68 m and 0.54 m in diameter respectively and 4.57 m tall. Life-cycle costs are not a discriminating factor in most cases, varying typically less than 15 percent.

  9. Notification of Regulated Waste Activity | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere I Geothermal Pwer PlantMunhall,Missouri: EnergyExcellence Seed LLCShores,Activity on Low Risk Sites

  10. Immobilized low-activity waste site borehole 299-E17-21

    SciTech Connect (OSTI)

    Reidel, S.P.; Reynolds, K.D.; Horton, D.G.

    1998-08-01T23:59:59.000Z

    The Tank Waste Remediation System (TWRS) is the group at the Hanford Site responsible for the safe underground storage of liquid waste from previous Hanford Site operations, the storage and disposal of immobilized tank waste, and closure of underground tanks. The current plan is to dispose of immobilized low-activity tank waste (ILAW) in new facilities in the southcentral part of 200-East Area and in four existing vaults along the east side of 200-East Area. Boreholes 299-E17-21, B8501, and B8502 were drilled at the southwest corner of the ILAW site in support of the Performance Assessment activities for the disposal options. This report summarizes the initial geologic findings, field tests conducted on those boreholes, and ongoing studies. One deep (480 feet) borehole and two shallow (50 feet) boreholes were drilled at the southwest corner of the ILAW site. The primary factor dictating the location of the boreholes was their characterization function with respect to developing the geohydrologic model for the site and satisfying associated Data Quality Objectives. The deep borehole was drilled to characterize subsurface conditions beneath the ILAW site, and two shallow boreholes were drilled to support an ongoing environmental tracer study. The tracer study will supply information to the Performance Assessment. All the boreholes provide data on the vadose zone and saturated zone in a previously uncharacterized area.

  11. Treatment of high-level wastes from the IFR fuel cycle

    SciTech Connect (OSTI)

    Johnson, T.R.; Lewis, M.A.; Newman, A.E.; Laidler, J.J.

    1992-01-01T23:59:59.000Z

    The Integral Fast Reactor (IFR) is being developed as a future commercial power source that promises to have important advantages over present reactors, including improved resource conservation and waste management. The spent metal alloy fuels from an IFR will be processed in an electrochemical cell operating at 500{degree}C with a molten chloride salt electrolyte and cadmium metal anode. After the actinides have been recovered from several batches of core and blanket fuels, the salt cadmium in this electrorefiner will be treated to separate fission products from residual transuranic elements. This treatment produces a waste salt that contains the alkali metal, alkaline earth, and halide fission products; some of the rare earths; and less than 100 nCi/g of alpha activity. The treated metal wastes contain the rest of the fission products (except T, Kr, and Xe) small amounts of uranium, and only trace amounts of transuranic elements. The current concept for the salt waste form is an aluminosilicate matrix, and the concept for the metal waste form is a corrosion-resistant metal alloy. The processes and equipment being developed to treat and immobilize the salt and metal wastes are described.

  12. Treatment of high-level wastes from the IFR fuel cycle

    SciTech Connect (OSTI)

    Johnson, T.R.; Lewis, M.A.; Newman, A.E.; Laidler, J.J.

    1992-08-01T23:59:59.000Z

    The Integral Fast Reactor (IFR) is being developed as a future commercial power source that promises to have important advantages over present reactors, including improved resource conservation and waste management. The spent metal alloy fuels from an IFR will be processed in an electrochemical cell operating at 500{degree}C with a molten chloride salt electrolyte and cadmium metal anode. After the actinides have been recovered from several batches of core and blanket fuels, the salt cadmium in this electrorefiner will be treated to separate fission products from residual transuranic elements. This treatment produces a waste salt that contains the alkali metal, alkaline earth, and halide fission products; some of the rare earths; and less than 100 nCi/g of alpha activity. The treated metal wastes contain the rest of the fission products (except T, Kr, and Xe) small amounts of uranium, and only trace amounts of transuranic elements. The current concept for the salt waste form is an aluminosilicate matrix, and the concept for the metal waste form is a corrosion-resistant metal alloy. The processes and equipment being developed to treat and immobilize the salt and metal wastes are described.

  13. Technology development at the Pacific Northwest Laboratory high-level waste management history

    SciTech Connect (OSTI)

    McElroy, J.L. [Geosafe Corp., Richland, WA (United States); Platt, A.M.

    1996-12-31T23:59:59.000Z

    During WWII and the post-WWII years, until the late 1950`s, plutonium production was Hanford`s primary mission. This mission produced an enormous legacy of wastes that have themselves become the new mission at Hanford. Waste management, as practiced at Hanford, during the defense production years was in many ways unique to Hanford, taking advantage of the dry climate, distance from the Columbia river and depth to the water table. Near-surface storage in tanks, ion exchange in seepage trenches and cribs, and near surface burial were the norm. Isolation of the wastes by the high and dry nature of the 200 Area plateau, where reprocessing and waste management took place, was one of the reasons Hanford had been selected for it`s nuclear mission. Thus, location was a significant aspect of the initial waste management program at Hanford. Treatment, other than simple chemical steps such as neutralization and ion exchange, had not been considered necessary to the mission and was therefore not developed. To support the development of commercial nuclear power and to provide improved means of handling nuclear wastes, new waste management programs were initiated in the 1950`s by the Atomic Energy Commission. The programs focused on high level waste. They included `spray calcination/vitrification` at Hanford Laboratories. Hanford Labs later became Pacific Northwest Laboratories (PNL) when Battelle Memorial Institute became the Operating Contractor in 1965. In 1996, it was renamed Pacific Northwest National Laboratory (PNNL). The purpose of this paper is to describe the HLW projects and programs that followed from this early HLW R&D at PNNL.

  14. Large-dimension, high-ZT Thermoelectric Nanocomposites for High-Power High-efficiency Waste Heat Recovery for Electricity Generation

    Broader source: Energy.gov [DOE]

    Large-dimension, high-ZT BiTe and Pb-based nanocomposites produced with a low-cost scalable process were used for development and testing of TE module prototypes, and demonstration of a waste heat recovery system

  15. WASTE INCINERATION wr090203 Activity 090203 SNAP CODE: 090203 SOURCE ACTIVITY TITLE: WASTE INCINERATION Flaring in Oil Refinery NOSE CODE: 109.03.11 NFR CODE:

    E-Print Network [OSTI]

    So Nox; Nmvoc Ch; Co Co; No Nh

    Flares are commonly used during petroleum refining for the safe disposal of waste gases during process upsets (e.g., start-up, shut-down, system blow-down) and emergencies to combust the organic content of waste emission streams without recovering/using the associated energy. 2 CONTRIBUTION TO TOTAL EMISSIONS Although flaring emission estimates are approximate, total hydrocarbon emissions from flaring at Canadian petroleum refineries during 1988 represented about 0.1 % of the refinery sector process and fugitive emissions that also included petroleum marketing emissions (CPPE, 1990). Thus the flaring operation at refineries is estimated to contribute a very small fraction of the total HC emissions in Canada. Emissions from flaring activities may also include: particulate, SOx, NOx, CO and other NMVOC. The CO2 contribution of both miscellaneous vent and flare emission sources represented approximately 9 % of the total petroleum refinery SO2 emission in Canada during 1988. Emissions estimates from flaring in petroleum refineries as reported in the CORINAIR90 inventory are summarised in Table 1. Table 1: Contribution to total emissions of the CORINAIR90 inventory (28 countries) Source-activity SNAP-code Contribution to total emissions [%

  16. Phase I high-level waste pretreatment and feed staging plan

    SciTech Connect (OSTI)

    Manuel, A.F.

    1996-02-05T23:59:59.000Z

    This document provides the preliminary planning basis for the U.S. Department of Energy (DOE) to provide a sufficient quantity of high-level waste feed to the privatization contractor during Phase I. By this analysis of candidate high-level waste feed sources, the initial quantity of high-level waste feed totals more than twice the minimum feed requirements. The flexibility of the current infrastructure within tank farms provides a variety of methods to transfer the feed to the privatization contractor`s site location. The amount and type of pretreatment (sludge washing) necessary for the Phase I processing can be tailored to support the demonstration goals without having a significant impact on glass volume (i.e., either inhibited water or caustic leaching can be used).

  17. Solvent extraction in the treatment of acidic high-level liquid waste : where do we stand?

    SciTech Connect (OSTI)

    Horwitz, E. P.; Schulz, W. W.

    1998-06-18T23:59:59.000Z

    During the last 15 years, a number of solvent extraction/recovery processes have been developed for the removal of the transuranic elements, {sup 90}Sr and {sup 137}Cs from acidic high-level liquid waste. These processes are based on the use of a variety of both acidic and neutral extractants. This chapter will present an overview and analysis of the various extractants and flowsheets developed to treat acidic high-level liquid waste streams. The advantages and disadvantages of each extractant along with comparisons of the individual systems are discussed.

  18. Methods of vitrifying waste with low melting high lithia glass compositions

    DOE Patents [OSTI]

    Jantzen, Carol M. (Aiken, SC); Pickett, John B. (Aiken, SC); Cicero-Herman, Connie A. (Aiken, SC); Marra, James C. (Aiken, SC)

    2001-01-01T23:59:59.000Z

    The invention relates to methods of vitrifying waste and for lowering the melting point of glass forming systems by including lithia formers in the glass forming composition in significant amounts, typically from about 0.16 wt % to about 11 wt %, based on the total glass forming oxides. The lithia is typically included as a replacement for alkali oxide glass formers that would normally be present in a particular glass forming system. Replacement can occur on a mole percent or weight percent basis, and typically results in a composition wherein lithia forms about 10 wt % to about 100 wt % of the alkali oxide glass formers present in the composition. The present invention also relates to the high lithia glass compositions formed by these methods. The invention is useful for stabilization of numerous types of waste materials, including aqueous waste streams, sludge solids, mixtures of aqueous supernate and sludge solids, combinations of spent filter aids from waste water treatment and waste sludges, supernate alone, incinerator ash, incinerator offgas blowdown, or combinations thereof, geological mine tailings and sludges, asbestos, inorganic filter media, cement waste forms in need of remediation, spent or partially spent ion exchange resins or zeolites, contaminated soils, lead paint, etc. The decrease in melting point achieved by the present invention desirably prevents volatilization of hazardous or radioactive species during vitrification.

  19. Reference design and operations for deep borehole disposal of high-level radioactive waste.

    SciTech Connect (OSTI)

    Herrick, Courtney Grant; Brady, Patrick Vane; Pye, Steven; Arnold, Bill Walter; Finger, John Travis; Bauer, Stephen J.

    2011-10-01T23:59:59.000Z

    A reference design and operational procedures for the disposal of high-level radioactive waste in deep boreholes have been developed and documented. The design and operations are feasible with currently available technology and meet existing safety and anticipated regulatory requirements. Objectives of the reference design include providing a baseline for more detailed technical analyses of system performance and serving as a basis for comparing design alternatives. Numerous factors suggest that deep borehole disposal of high-level radioactive waste is inherently safe. Several lines of evidence indicate that groundwater at depths of several kilometers in continental crystalline basement rocks has long residence times and low velocity. High salinity fluids have limited potential for vertical flow because of density stratification and prevent colloidal transport of radionuclides. Geochemically reducing conditions in the deep subsurface limit the solubility and enhance the retardation of key radionuclides. A non-technical advantage that the deep borehole concept may offer over a repository concept is that of facilitating incremental construction and loading at multiple perhaps regional locations. The disposal borehole would be drilled to a depth of 5,000 m using a telescoping design and would be logged and tested prior to waste emplacement. Waste canisters would be constructed of carbon steel, sealed by welds, and connected into canister strings with high-strength connections. Waste canister strings of about 200 m length would be emplaced in the lower 2,000 m of the fully cased borehole and be separated by bridge and cement plugs. Sealing of the upper part of the borehole would be done with a series of compacted bentonite seals, cement plugs, cement seals, cement plus crushed rock backfill, and bridge plugs. Elements of the reference design meet technical requirements defined in the study. Testing and operational safety assurance requirements are also defined. Overall, the results of the reference design development and the cost analysis support the technical feasibility of the deep borehole disposal concept for high-level radioactive waste.

  20. PLUTONIUM SOLUBILITY IN HIGH-LEVEL WASTE ALKALI BOROSILICATE GLASS

    SciTech Connect (OSTI)

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

    2011-01-04T23:59:59.000Z

    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. The incorporation of 1 wt % Pu in the glass did not adversely impact glass viscosity (as assessed using Hf surrogate) or glass durability. Finally, evaluation of DWPF glass pour samples that had Pu concentrations below the 897 g/m{sup 3} limit showed that Pu concentrations in the glass pour stream were close to targeted compositions in the melter feed indicating that Pu neither volatilized from the melt nor stratified in the melter when processed in the DWPF melter.

  1. Assessment of alternatives for management of ORNL retrievable transuranic waste. Nuclear Waste Program: transuranic waste (Activity No. AR 05 15 15 0; ONL-WT04)

    SciTech Connect (OSTI)

    Not Available

    1980-10-01T23:59:59.000Z

    Since 1970, solid waste with TRU or U-233 contamination in excess of 10 ..mu..Ci per kilogram of waste has been stored in a retrievable fashion at ORNL, such as in ss drums, concrete casks, and ss-lined wells. This report describes the results of a study performed to identify and evaluate alternatives for management of this waste and of the additional waste projected to be stored through 1995. The study was limited to consideration of the following basic strategies: Strategy 1: Leave waste in place as is; Strategy 2: Improve waste confinement; and Strategy 3: Retrieve waste and process for shipment to a Federal repository. Seven alternatives were identified and evaluated, one each for Strategies 1 and 2 and five for Strategy 3. Each alternative was evaluated from the standpoint of technical feasibility, cost, radiological risk and impact, regulatory factors and nonradiological environmental impact.

  2. Melt Rate Improvement for High-Level Waste Glass

    SciTech Connect (OSTI)

    Matyas, Josef; Hrma, Pavel R.; Kim, Dong-Sang

    2002-09-09T23:59:59.000Z

    This report summarizes results of research accomplished during the first year of the 3-year project. The data presented in this report have been gathered to support work on the mathematical modeling of waste-glass melters. At this stage, only a qualitative description and interpretation of the observed phenomena has been attempted. Two Savannah Rive feeds were used for the study. These feeds were subjected to thermal gravimetric analysis, differential thermal analysis, differential scanning calorimetry, evolved gas analysis with volume-expansion monitoring, modified reboil test, quantitative X-ray diffraction, scanning electron microscopy with energy dispersive spectroscopy, wet chemical analysis, and M?ssbauer spectroscopy. Glass viscosity was also measured. Finally, it was recommended to use melt-rate furnace test data to measure thermal diffusivity of the feed. Though both feed were reduced to prevent oxygen evolution from the melt, oxygen evolved form one of the melts and COx evolved from both. Hence, foam is likely to form under the cold cap even when the feed is reduced. An important difference between the feeds was in the melt viscosity at the temperature at which the melt interfaces the cold cap. It was suggested that low viscosity destabilizes foam under the cold cap, thus enhancing the rate of melting.

  3. Disposition of actinides released from high-level waste glass

    SciTech Connect (OSTI)

    Ebert, W.L.; Bates, J.K.; Buck, E.C.; Gong, M.; Wolf, S.F.

    1994-05-01T23:59:59.000Z

    A series of static leach tests was conducted using glasses developed for vitrifying tank wastes at the Savannah River Site to monitor the disposition of actinide elements upon corrosion of the glasses. In these tests, glasses produced from SRL 131 and SRL 202 frits were corroded at 90{degrees}C in a tuff groundwater. Tests were conducted using crushed glass at different glass surface area-to-solution volume (S/V) ratios to assess the effect of the S/V on the solution chemistry, the corrosion of the glass, and the disposition of actinide elements. Observations regarding the effects of the S/V on the solution chemistry and the corrosion of the glass matrix have been reported previously. This paper highlights the solution analyses performed to assess how the S/V used in a static leach test affects the disposition of actinide elements between fractions that are suspended or dissolved in the solution, and retained by the altered glass or other materials.

  4. Engineered waste-package-system design specification

    SciTech Connect (OSTI)

    Not Available

    1983-05-01T23:59:59.000Z

    This report documents the waste package performance requirements and geologic and waste form data bases used in developing the conceptual designs for waste packages for salt, tuff, and basalt geologies. The data base reflects the latest geotechnical information on the geologic media of interest. The parameters or characteristics specified primarily cover spent fuel, defense high-level waste, and commercial high-level waste forms. The specification documents the direction taken during the conceptual design activity. A separate design specification will be developed prior to the start of the preliminary design activity.

  5. Annual summary of Immobilized Low-Activity Waste (ILAW) Performance Assessment for 2003 Incorporating the Integrated Disposal Facility Concept

    SciTech Connect (OSTI)

    MANN, F M

    2003-09-01T23:59:59.000Z

    To Erik Olds 09/30/03 - An annual summary of the adequacy of the Hanford Immobilized Low-Activity Tank Waste Performance Assessment (ILAW PA) is necessary in each year in which a full performance assessment is not issued.

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

    SciTech Connect (OSTI)

    Larson, D.E. (ed.)

    1980-09-01T23:59:59.000Z

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

  7. Risk perception on management of nuclear high-level and transuranic waste storage

    SciTech Connect (OSTI)

    Dees, L.A.

    1994-08-15T23:59:59.000Z

    The Department of Energy`s program for disposing of nuclear High-Level Waste (HLW) and transuranic (TRU) waste has been impeded by overwhelming political opposition fueled by public perceptions of actual risk. Analysis of these perceptions shows them to be deeply rooted in images of fear and dread that have been present since the discovery of radioactivity. The development and use of nuclear weapons linked these images to reality and the mishandling of radioactive waste from the nations military weapons facilities has contributed toward creating a state of distrust that cannot be erased quickly or easily. In addition, the analysis indicates that even the highly educated technical community is not well informed on the latest technology involved with nuclear HLW and TRU waste disposal. It is not surprising then, that the general public feels uncomfortable with DOE`s management plans for with nuclear HLW and TRU waste disposal. Postponing the permanent geologic repository and use of Monitored Retrievable Storage (MRS) would provide the time necessary for difficult social and political issues to be resolved. It would also allow time for the public to become better educated if DOE chooses to become proactive.

  8. High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 2

    SciTech Connect (OSTI)

    Cunnane, J.C. [comp.; Bates, J.K.; Bradley, C.R. [Argonne National Lab., IL (United States)] [and others

    1994-03-01T23:59:59.000Z

    The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion.This document is organized into three volumes. Volumes I and II represent a tiered set of information intended for somewhat different audiences. Volume I is intended to provide an overview of waste glass corrosion, and Volume 11 is intended to provide additional experimental details on experimental factors that influence waste glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II. Volume I is intended for managers, decision makers, and modelers, the combined set of Volumes I, II, and III is intended for scientists and engineers working in the field of high-level waste.

  9. Development and results of experimental testing of electromembrane process for liquid active waste purification

    SciTech Connect (OSTI)

    Martinov, B.V.; Smirnov, V.V.; Tugolukov, B.B.; Belyakov, Y.A. [A.A. Bochvar All Russian Scientific Research, Moscow (Russian Federation). Inst. of Inorganic Materials

    1993-12-31T23:59:59.000Z

    This paper discusses the results of studies on electromembrane purification. The concentration of salts in active wastes arising from decontamination is more than 3--5 g/l. For these investigations a solution was chosen that had arisen from the decontamination of metallic items by a two-bath method using permanganate-alkali in the first stage and nitrogen oxalic acid in the second stage. The total salt content of mixed acid and alkaline solutions was 3.0 g/l, with a pH of 8.5 and total beta-activity of 6 {times} 10{sup {minus}8} Ci/l.

  10. Independent Assessment of the Savannah River Site High-Level Waste Salt Disposition Alternatives Evaluation

    SciTech Connect (OSTI)

    J. T. Case (DOE-ID); M. L. Renfro (INEEL)

    1998-12-01T23:59:59.000Z

    This report presents the results of the Independent Project Evaluation (IPE) Team assessment of the Westinghouse Savannah River Company High-Level Waste Salt Disposition Systems Engineering (SE) Team's deliberations, evaluations, and selections. The Westinghouse Savannah River Company concluded in early 1998 that production goals and safety requirements for processing SRS HLW salt to remove Cs-137 could not be met in the existing In-Tank Precipitation Facility as currently configured for precipitation of cesium tetraphenylborate. The SE Team was chartered to evaluate and recommend an alternative(s) for processing the existing HLW salt to remove Cs-137. To replace the In-Tank Precipitation process, the Savannah River Site HLW Salt Disposition SE Team downselected (October 1998) 140 candidate separation technologies to two alternatives: Small-Tank Tetraphenylborate (TPB) Precipitation (primary alternative) and Crystalline Silicotitanate (CST) Nonelutable Ion Exchange (backup alternative). The IPE Team, commissioned by the Department of Energy, concurs that both alternatives are technically feasible and should meet all salt disposition requirements. But the IPE Team judges that the SE Team's qualitative criteria and judgments used in their downselection to a primary and a backup alternative do not clearly discriminate between the two alternatives. To properly choose between Small-Tank TPB and CST Ion Exchange for the primary alternative, the IPE Team suggests the following path forward: Complete all essential R and D activities for both alternatives and formulate an appropriate set of quantitative decision criteria that will be rigorously applied at the end of the R and D activities. Concurrent conceptual design activities should be limited to common elements of the alternatives.

  11. The potential consequences and risks of highway accidents involving gamma-emitting low specific activity (LSA) waste

    SciTech Connect (OSTI)

    Ostmeyer, R.M.; Finley, N.C.; Cashwell, J.W.; McClure, J.D.

    1988-08-01T23:59:59.000Z

    This report documents an analysis of the potential risks and consequences of highway accidents that lead to the release of gamma-emitting low specific activity (LSA) materials from their packagings. Of particular concern is LSA waste with a specific activity level near the existing regulatory limit. The results of the analysis are intended primarily to provide a basis for evaluating limits on the quantity of radionuclides that can be contained in a single LSA waste shipment. 81 refs., 39 figs., 52 tabs.

  12. Structural integrity and potential failure modes of hanford high-level waste tanks

    SciTech Connect (OSTI)

    Han, F.C.

    1996-09-30T23:59:59.000Z

    Structural Integrity of the Hanford High-Level Waste Tanks were evaluated based on the existing Design and Analysis Documents. All tank structures were found adequate for the normal operating and seismic loads. Potential failure modes of the tanks were assessed by engineering interpretation and extrapolation of the existing engineering documents.

  13. Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)

    SciTech Connect (OSTI)

    Burgard, K.C.

    1998-04-09T23:59:59.000Z

    The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

  14. Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)

    SciTech Connect (OSTI)

    Burgard, K.C.

    1998-06-02T23:59:59.000Z

    The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

  15. Engineering development of a lightweight high-pressure scarifier for tank waste retrieval

    SciTech Connect (OSTI)

    Hatchell, B.K.

    1997-09-01T23:59:59.000Z

    The Retrieval Process Development and Enhancements Program (RPD&E) is sponsored by the U.S. Department of Energy Tanks Focus Area to investigate existing and emerging retrieval processes suitable for the retrieval of high-level radioactive waste inside underground storage tanks. This program, represented by industry, national laboratories, and academia, seeks to provide a technical and cost basis to support site-remediation decisions. Part of this program has involved the development of a high-pressure waterjet dislodging system and pneumatic conveyance integrated as a scarifier. Industry has used high-pressure waterjet technology for many years to mine, cut, clean, and scarify materials with a broad range of properties. The scarifier was developed as an alternate means of retrieving waste inside Hanford single-shell tanks, particularly hard, stubborn waste. Testing of the scarifier has verified its ability to retrieve a wide range of tank waste ranging from extremely hard waste that is resistant to other dislodging means to soft sludge and even supernatant fluid. Since the scarifier expends water at a low rate and recovers most of the water as it is used, the scarifier is well suited for retrieval of tanks that leak and cannot be safely sluiced or applications where significant waste dilution is not acceptable. Although the original scarifier was effective, it became evident that a lighter, more compact version that would be compatible with light weight deployment systems under development, such as the Light Duty Utility Arm, was needed. At the end of FY 95, the Light Weight Scarifier (LWS) was designed to incorporate the features of the original scarifier in a smaller, lighter end effector. During FY 96, the detailed design of the LWS was completed and two prototypes were fabricated.

  16. High-Level Waste Corporate Board, Mark Gilbertson

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM Flash2011-12Approved on 24 July 2008 1 Office ofHighHigh-LevelHigh-Level

  17. High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 3

    SciTech Connect (OSTI)

    Cunnane, J.C. [comp.; Bates, J.K.; Bradley, C.R. [Argonne National Lab., IL (United States)] [and others

    1994-03-01T23:59:59.000Z

    The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II.

  18. Simulation of Self-Irradiation of High-Sodium Content Nuclear Waste Glasses

    SciTech Connect (OSTI)

    Pankov, Alexey S.; Ojovan, Michael I. [Immobilisation Science Laboratory, Department of Engineering Materials, University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield, S1 3JD (United Kingdom); Batyukhnova, Olga G. [International Education Training Centre, SUE SIA 'Radon', The 7-th Rostovsky Lane 2/14, Moscow, 119121 (Russian Federation); Lee, William E. [Department of Materials, Imperial College London, South Kensington Campus, Exhibition Road, London, SW7 2AZ (United Kingdom)

    2007-07-01T23:59:59.000Z

    Alkali-borosilicate glasses are widely used in nuclear industry as a matrix for immobilisation of hazardous radioactive wastes. Durability or corrosion resistance of these glasses is one of key parameters in waste storage and disposal safety. It is influenced by many factors such as composition of glass and surrounding media, temperature, time and so on. As these glasses contain radioactive elements most of their properties including corrosion resistance are also impacted by self-irradiation. The effect of external gamma-irradiation on the short-term (up to 27 days) dissolution of waste borosilicate glasses at moderate temperatures (30 deg. to 60 deg. C) was studied. The glasses studied were Magnox Waste glass used for immobilisation of HLW in UK, and K-26 glass used in Russia for ILW immobilisation. Glass samples were irradiated under {gamma}-source (Co-60) up to doses 1 and 11 MGy. Normalised rates of elemental release and activation energy of release were measured for Na, Li, Ca, Mg, B, Si and Mo before and after irradiation. Irradiation up to 1 MGy results in increase of leaching rate of almost all elements from both MW and K-26 with the exception of Na release from MW glass. Further irradiation up to a dose of 11 MGy leads to the decrease of elemental release rates to nearly initial value. Another effect of irradiation is increase of activation energies of elemental release. (authors)

  19. Closure development for high-level nuclear waste containers for the tuff repository; Phase 1, Final report

    SciTech Connect (OSTI)

    Robitz, E.S. Jr.; McAninch, M.D. Jr.; Edmonds, D.P. [Babcock and Wilcox Co., Lynchburg, VA (USA). Nuclear Power Div.]|[Babcock and Wilcox Co., Alliance, OH (USA). Research and Development Div.

    1990-09-01T23:59:59.000Z

    This report summarizes Phase 1 activities for closure development of the high-level nuclear waste package task for the tuff repository. Work was conducted under U.S. Department of Energy (DOE) Contract 9172105, administered through the Lawrence Livermore National Laboratory (LLNL), as part of the Yucca Mountain Project (YMP), funded through the DOE Office of Civilian Radioactive Waste Management (OCRWM). The goal of this phase was to select five closure processes for further evaluation in later phases of the program. A decision tree methodology was utilized to perform an objective evaluation of 15 potential closure processes. Information was gathered via a literature survey, industrial contacts, and discussions with project team members, other experts in the field, and the LLNL waste package task staff. The five processes selected were friction welding, electron beam welding, laser beam welding, gas tungsten arc welding, and plasma arc welding. These are felt to represent the best combination of weldment material properties and process performance in a remote, radioactive environment. Conceptual designs have been generated for these processes to illustrate how they would be implemented in practice. Homopolar resistance welding was included in the Phase 1 analysis, and developments in this process will be monitored via literature in Phases 2 and 3. Work was conducted in accordance with the YMP Quality Assurance Program. 223 refs., 20 figs., 9 tabs.

  20. Potential Application Of Radionuclide Scaling Factors To High Level Waste Characterization

    SciTech Connect (OSTI)

    Reboul, S. H.

    2013-09-30T23:59:59.000Z

    Production sources, radiological properties, relative solubilities in waste, and laboratory analysis techniques for the forty-five radionuclides identified in Hanford?s Waste Treatment and Immobilization Plant (WTP) Feed Acceptance Data Quality Objectives (DQO) document are addressed in this report. Based on Savannah River Site (SRS) experience and waste characteristics, thirteen of the radionuclides are judged to be candidates for potential scaling in High Level Waste (HLW) based on the concentrations of other radionuclides as determined through laboratory measurements. The thirteen radionuclides conducive to potential scaling are: Ni-59, Zr-93, Nb-93m, Cd-113m, Sn-121m, Sn-126, Cs-135, Sm-151, Ra-226, Ra-228, Ac-227, Pa-231, and Th-229. The ability to scale radionuclides is useful from two primary perspectives: 1) it provides a means of checking the radionuclide concentrations that have been determined by laboratory analysis; and 2) it provides a means of estimating radionuclide concentrations in the absence of a laboratory analysis technique or when a complex laboratory analysis technique fails. Along with the rationale for identifying and applying the potential scaling factors, this report also provides examples of using the scaling factors to estimate concentrations of radionuclides in current SRS waste and into the future. Also included in the report are examples of independent laboratory analysis techniques that can be used to check results of key radionuclide analyses. Effective utilization of radionuclide scaling factors requires understanding of the applicable production sources and the chemistry of the waste. As such, the potential scaling approaches identified in this report should be assessed from the perspective of the Hanford waste before reaching a decision regarding WTP applicability.

  1. Development And Initial Testing Of Off-Gas Recycle Liquid From The WTP Low Activity Waste Vitrification Process - 14333

    SciTech Connect (OSTI)

    McCabe, Daniel J.; Wilmarth, William R.; Nash, Charles A.; Taylor-Pashow, Kathryn M.; Adamson, Duane J.; Crawford, Charles L.; Morse, Megan M.

    2014-01-07T23:59:59.000Z

    The Waste Treatment and Immobilization Plant (WTP) process flow was designed to pre-treat feed from the Hanford tank farms, separate it into a High Level Waste (HLW) and Low Activity Waste (LAW) fraction and vitrify each fraction in separate facilities. Vitrification of the waste generates an aqueous condensate stream from the off-gas processes. This stream originates from two off-gas treatment unit operations, the Submerged Bed Scrubber (SBS) and the Wet Electrospray Precipitator (WESP). Currently, the baseline plan for disposition of the stream from the LAW melter is to recycle it to the Pretreatment facility where it gets evaporated and processed into the LAW melter again. If the Pretreatment facility is not available, the baseline disposition pathway is not viable. Additionally, some components in the stream are volatile at melter temperatures, thereby accumulating to high concentrations in the scrubbed stream. It would be highly beneficial to divert this stream to an alternate disposition path to alleviate the close-coupled operation of the LAW vitrification and Pretreatment facilities, and to improve long-term throughput and efficiency of the WTP system. In order to determine an alternate disposition path for the LAW SBS/WESP Recycle stream, a range of options are being studied. A simulant of the LAW Off-Gas Condensate was developed, based on the projected composition of this stream, and comparison with pilot-scale testing. The primary radionuclide that vaporizes and accumulates in the stream is Tc-99, but small amounts of several other radionuclides are also projected to be present in this stream. The processes being investigated for managing this stream includes evaporation and radionuclide removal via precipitation and adsorption. During evaporation, it is of interest to investigate the formation of insoluble solids to avoid scaling and plugging of equipment. Key parameters for radionuclide removal include identifying effective precipitation or ion adsorption chemicals, solid-liquid separation methods, and achievable decontamination factors. Results of the radionuclide removal testing indicate that the radionuclides, including Tc-99, can be removed with inorganic sorbents and precipitating agents. Evaporation test results indicate that the simulant can be evaporated to fairly high concentration prior to formation of appreciable solids, but corrosion has not yet been examined.

  2. Action plan for responses to abnormal conditions in Hanford Site radioactive waste tanks with high organic content. Revision 1

    SciTech Connect (OSTI)

    Fowler, K.D.

    1993-07-01T23:59:59.000Z

    This action plan describes the criteria and the organizational responsibilities required for ensuring that waste storage tanks with high organic contents are maintained in a safe condition at the Hanford Site. In addition, response actions are outlined for (1) prevention or mitigation of excessive temperatures; or (2) a material release from any waste tank with high organic content. Other response actions may be defined by Westinghouse Hanford Company Systems Engineering if a waste tank parameter goes out of specification. Trend analysis indicates the waste tank parameters have seasonal variations, but are otherwise stable.

  3. Greater-than-Class C low-level radioactive waste characterization: Estimated volumes, radionuclide activities, and other characteristics. Revision 1

    SciTech Connect (OSTI)

    Not Available

    1994-09-01T23:59:59.000Z

    The Department of Energy`s (DOE`s) planning for the disposal of greater-than-Class C low-level radioactive waste (GTCC LLW) requires characterization of the waste. This report estimates volumes, radionuclide activities, and waste forms of GTCC LLW to the year 2035. It groups the waste into four categories, representative of the type of generator or holder of the waste: Nuclear Utilities, Sealed Sources, DOE-Held, and Other Generator. GTCC LLW includes activated metals (activation hardware from reactor operation and decommissioning), process wastes (i.e., resins, filters, etc.), sealed sources, and other wastes routinely generated by users of radioactive material. Estimates reflect the possible effect that packaging and concentration averaging may have on the total volume of GTCC LLW. Possible GTCC mixed LLW is also addressed. Nuclear utilities will probably generate the largest future volume of GTCC LLW with 65--83% of the total volume. The other generators will generate 17--23% of the waste volume, while GTCC sealed sources are expected to contribute 1--12%. A legal review of DOE`s obligations indicates that the current DOE-Held wastes described in this report will not require management as GTCC LLW because of the contractual circumstances under which they were accepted for storage. This report concludes that the volume of GTCC LLW should not pose a significant management problem from a scientific or technical standpoint. The projected volume is small enough to indicate that a dedicated GTCC LLW disposal facility may not be justified. Instead, co-disposal with other waste types is being considered as an option.

  4. The kinetics of spinel crystallization from a high-level waste glass

    SciTech Connect (OSTI)

    Reynolds, J.G. [Univ. of Idaho, Moscow, ID (United States). Div. of Soils; Hrma, P. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-12-31T23:59:59.000Z

    The kinetics of spinel crystallization from a molten high-iron simulated high-level nuclear waste glass was studied using isothermal heat treatments. Optical microscopy with image analysis was used to measure volume fraction of spinel as a function of heat treatment time and temperature. The Johnson-Mehl-Avrami equation was fitted to data to determine kinetic coefficients for spinel crystallization. The liquidus temperature and Avrami number are T{sub L} = 1,337K and n = 1.5.

  5. High energy activation data library (HEAD-2009)

    SciTech Connect (OSTI)

    Mashnik, Stepan G [Los Alamos National Laboratory; Korovin, Yury A [NON LANL; Natalenko, Anatoly A [NON LANL; Konobeyev, Alexander Yu [NON LANL; Stankovskiy, A Yu [NON LANL

    2010-01-01T23:59:59.000Z

    A proton activation data library for 682 nuclides from 1 H to 210Po in the energy range from 150 MeV up to 1 GeV was developed. To calculate proton activation data, the MCNPX 2.6.0 and CASCADE/INPE codes were chosen. Different intranuclear cascade, preequilibrium, and equilibrium nuclear reaction models and their combinations were used. The optimum calculation models have been chosen on the basis of statistical correlations for calculated and experimental proton data taken from the EXFOR library of experimental nuclear data. All the data are written in ENDF-6 format. The library is called HEPAD-2008 (High-Energy Proton Activation Data). A revision of IEAF-2005 neutron activation data library has been performed. A set of nuclides for which the cross-section data can be (and were) updated using more modern and improved models is specified, and the corresponding calculations have been made in the present work. The new version of the library is called IEAF-2009. The HEPAD-2008 and IEAF-2009 are merged to the final HEAD-2009 library.

  6. Summary report. Low-level radioactive waste management activities in the states and compacts. Volume 4, No. 2

    SciTech Connect (OSTI)

    NONE

    1996-08-01T23:59:59.000Z

    `Low-Level Radioactive Waste Management Activities in the States and Compacts` is a supplement to `LLW Notes` and is distributed periodically by Afton Associates, Inc. to state, compact and federal officials that receive `LLW Notes`. The Low-Level Radioactive Waste Forum (LLW Forum) is an association of state and compact representatives, appointed by governors and compact commissions, established to facilitate state and compact implementation of the Low- Level Radioactive Waste Policy Act of 1980 and the Low-Level Radioactive Waste Policy Amendments Act of 1985 and to promote the objectives of low-level radioactive waste regional compacts. The LLW Forum provides an opportunity for state and compact officials to share information with one another and to exchange views with officials of federal agencies and other interested parties.

  7. Summary report, low-level radioactive waste management activities in the states and compacts. Vol. 4. No. 1

    SciTech Connect (OSTI)

    NONE

    1996-01-01T23:59:59.000Z

    `Low-Level Radioactive Waste Management Activities in the States and Compacts` is a supplement to `LLW Notes` and is distributed periodically by Afton Associates, Inc. to state, compact and federal officials that receive `LLW Notes`. The Low-Level Radioactive Waste Forum (LLW Forum) is an association of state and compact representatives, appointed by governors and compact commissions, established to facilitate state and compact implementation of the Low- Level Radioactive Waste Policy Act of 1980 and the Low-Level Radioactive Waste Policy Amendments Act of 1985 and to promote the objectives of low-level radioactive waste regional compacts. The LLW Forum provides an opportunity for state and compact officials to share information with one another and to exchange views with officials of federal agencies and other interested parties.

  8. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 7

    SciTech Connect (OSTI)

    Not Available

    1994-04-01T23:59:59.000Z

    This Requirements Identification Document (RID) describes an Occupational Health and Safety Program as defined through the Relevant DOE Orders, regulations, industry codes/standards, industry guidance documents and, as appropriate, good industry practice. The definition of an Occupational Health and Safety Program as specified by this document is intended to address Defense Nuclear Facilities Safety Board Recommendations 90-2 and 91-1, which call for the strengthening of DOE complex activities through the identification and application of relevant standards which supplement or exceed requirements mandated by DOE Orders. This RID applies to the activities, personnel, structures, systems, components, and programs involved in maintaining the facility and executing the mission of the High-Level Waste Storage Tank Farms.

  9. Production Of High Specific Activity Copper-67

    SciTech Connect (OSTI)

    Jamriska, Sr., David J. (Los Alamos, NM); Taylor, Wayne A. (Los Alamos, NM); Ott, Martin A. (Los Alamos, NM); Fowler, Malcolm (Los Alamos, NM); Heaton, Richard C. (Los Alamos, NM)

    2003-10-28T23:59:59.000Z

    A process for the selective production and isolation of high specific activity Cu.sup.67 from proton-irradiated enriched Zn.sup.70 target comprises target fabrication, target irradiation with low energy (<25 MeV) protons, chemical separation of the Cu.sup.67 product from the target material and radioactive impurities of gallium, cobalt, iron, and stable aluminum via electrochemical methods or ion exchange using both anion and cation organic ion exchangers, chemical recovery of the enriched Zn.sup.70 target material, and fabrication of new targets for re-irradiation is disclosed.

  10. Production Of High Specific Activity Copper-67

    SciTech Connect (OSTI)

    Jamriska, Sr., David J. (Los Alamos, NM); Taylor, Wayne A. (Los Alamos, NM); Ott, Martin A. (Los Alamos, NM); Fowler, Malcolm (Los Alamos, NM); Heaton, Richard C. (Los Alamos, NM)

    2002-12-03T23:59:59.000Z

    A process for the selective production and isolation of high specific activity cu.sup.67 from proton-irradiated enriched Zn.sup.70 target comprises target fabrication, target irradiation with low energy (<25 MeV) protons, chemical separation of the Cu.sup.67 product from the target material and radioactive impurities of gallium, cobalt, iron, and stable aluminum via electrochemical methods or ion exchange using both anion and cation organic ion exchangers, chemical recovery of the enriched Zn.sup.70 target material, and fabrication of new targets for re-irradiation is disclosed.

  11. High temperature materials for radioactive waste incineration and vitrification. Revision 1

    SciTech Connect (OSTI)

    Bickford, D F; Ondrejcin, R S; Salley, L

    1986-01-01T23:59:59.000Z

    Incineration or vitrification of radioactive waste subjects equipment to alkaline or acidic fluxing, oxidation, sulfidation, carburization, and thermal shock. It is necessary to select appropriate materials of construction and control operating conditions to avoid rapid equipment failure. Nickel- and cobalt-based alloys with high chromium or aluminum content and aluminum oxide/chromium oxide refractories with high chromium oxide content have provided the best service in pilot-scale melter tests. Inconel 690 and Monofrax K-3 are being used for waste vitrification. Haynes 188 and high alumina refractory are undergoing pilot scale tests for incineration equipment. Laboratory tests indicate that alloys and refractories containing still higher concentrations of chromium or chromium oxide, such as Inconel 671 and Monofrax E, may provide superior resistance to attack in glass melter environments.

  12. High Level Waste Corporate Board Charter | Department of Energy

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742Energy ChinaofSchaefer To:Department of Energy CompletingPresented By:DanielHigh

  13. High-Level Liquid Waste Tank Integrity Workshop - 2008

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM Flash2011-12Approved on 24 July 2008 1 Office ofHigh

  14. Scenarios for the Hanford immobilized Low-Activity waste (ILAW) performance assessment

    SciTech Connect (OSTI)

    MANN, F.M.

    1999-09-09T23:59:59.000Z

    The purpose of the next version of the Hanford Immobilized Low-Activity Tank Waste (ILAW) Performance Assessment (ILAW PA) is to provide an updated estimate of the long-term human health and environmental impact of the disposal of ILAW and to compare these estimates against performance objectives displayed in Tables 1,2, and 3 (Mann 1999a). Such a radiological performance assessment is required by U.S. Department of Energy (DOE) Orders on radioactive waste management (DOE 1988a and DOE 1999a). This document defines the scenarios that will be used for the next update of the PA that is scheduled to be issued in 2001. Since the previous performance assessment (Mann 1998) was issued, considerable additional data on waste form behavior and site-specific soil geotechnical properties have been collected. In addition, the 2001 ILAW PA will benefit from improved computer models and the experience gained from the previous performance assessment. However, the scenarios (that is, the features, events, and processes analyzed in the Performance assessment) for the next PA are very similar to the ones in the 1998 PA.

  15. TWRS Retrieval and Storage Mission and Immobilized Low Activity Waste (ILAW) Disposal Plan

    SciTech Connect (OSTI)

    BURBANK, D.A.

    1999-09-01T23:59:59.000Z

    This project plan has a twofold purpose. First, it provides a waste stream project plan specific to the River Protection Project (RPP) (formerly the Tank Waste Remediation System [TWRS] Project) Immobilized Low-Activity Waste (LAW) Disposal Subproject for the Washington State Department of Ecology (Ecology) that meets the requirements of Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-90-01 (Ecology et al. 1994) and is consistent with the project plan content guidelines found in Section 11.5 of the Tri-Party Agreement action plan (Ecology et al. 1998). Second, it provides an upper tier document that can be used as the basis for future subproject line-item construction management plans. The planning elements for the construction management plans are derived from applicable U.S. Department of Energy (DOE) planning guidance documents (DOE Orders 4700.1 [DOE 1992] and 430.1 [DOE 1995a]). The format and content of this project plan are designed to accommodate the requirements mentioned by the Tri-Party Agreement and the DOE orders. A cross-check matrix is provided in Appendix A to explain where in the plan project planning elements required by Section 11.5 of the Tri-Party Agreement are addressed.

  16. Round-robin testing of a reference glass for low-activity waste forms

    SciTech Connect (OSTI)

    Ebert, W. L.; Wolf, S. F.

    1999-12-06T23:59:59.000Z

    A round robin test program was conducted with a glass that was developed for use as a standard test material for acceptance testing of low-activity waste glasses made with Hanford tank wastes. The glass is referred to as the low-activity test reference material (LRM). The program was conducted to measure the interlaboratory reproducibility of composition analysis and durability test results. Participants were allowed to select the methods used to analyze the glass composition. The durability tests closely followed the Product Consistency Test (PCT) Method A, except that tests were conducted at both 40 and 90 C and that parallel tests with a reference glass were not required. Samples of LRM glass that had been crushed, sieved, and washed to remove fines were provided to participants for tests and analyses. The reproducibility of both the composition and PCT results compare favorably with the results of interlaboratory studies conducted with other glasses. From the perspective of reproducibility of analysis results, this glass is acceptable for use as a composition standard for nonradioactive components of low-activity waste forms present at >0.1 elemental mass % and as a test standard for PCTS at 40 and 90 C. For PCT with LRM glass, the expected test results at the 95% confidence level are as follows: (1) at 40 C: pH = 9.86 {+-} 0.96; [B] = 2.30 {+-} 1.25 mg/L; [Na] = 19.7 {+-} 7.3 mg/L; [Si] = 13.7 {+-} 4.2 mg/L; and (2) at 90 C: pH = 10.92 {+-} 0.43; [B] = 26.7 {+-} 7.2 mg/L; [Na] = 160 {+-} 13 mg/L; [Si] = 82.0 {+-} 12.7 mg/L. These ranges can be used to evaluate the accuracy of PCTS conducted at other laboratories.

  17. Waste Heat Recovery from High Temperature Off-Gases from Electric Arc Furnace

    SciTech Connect (OSTI)

    Nimbalkar, Sachin U [ORNL; Thekdi, Arvind [E3M Inc; Keiser, James R [ORNL; Storey, John Morse [ORNL

    2014-01-01T23:59:59.000Z

    This article presents a study and review of available waste heat in high temperature Electric Arc Furnace (EAF) off gases and heat recovery techniques/methods from these gases. It gives details of the quality and quantity of the sensible and chemical waste heat in typical EAF off gases, energy savings potential by recovering part of this heat, a comprehensive review of currently used waste heat recovery methods and potential for use of advanced designs to achieve a much higher level of heat recovery including scrap preheating, steam production and electric power generation. Based on our preliminary analysis, currently, for all electric arc furnaces used in the US steel industry, the energy savings potential is equivalent to approximately 31 trillion Btu per year or 32.7 peta Joules per year (approximately $182 million US dollars/year). This article describes the EAF off-gas enthalpy model developed at Oak Ridge National Laboratory (ORNL) to calculate available and recoverable heat energy for a given stream of exhaust gases coming out of one or multiple EAF furnaces. This Excel based model calculates sensible and chemical enthalpy of the EAF off-gases during tap to tap time accounting for variation in quantity and quality of off gases. The model can be used to estimate energy saved through scrap preheating and other possible uses such as steam generation and electric power generation using off gas waste heat. This article includes a review of the historical development of existing waste heat recovery methods, their operations, and advantages/limitations of these methods. This paper also describes a program to develop and test advanced concepts for scrap preheating, steam production and electricity generation through use of waste heat recovery from the chemical and sensible heat contained in the EAF off gases with addition of minimum amount of dilution or cooling air upstream of pollution control equipment such as bag houses.

  18. Waste minimization through high-pressure microwave digestion of soils for gross {alpha}/{beta} analyses

    SciTech Connect (OSTI)

    Yaeger, J.S.; Smith, L.L.

    1995-04-01T23:59:59.000Z

    As a result of the U.S. Department of Energy`s (DOE) environmental restoration and waste management activities, laboratories receive numerous analytical requests for gross {alpha}/{beta} analyses. Traditional sample preparation methods for gross {alpha}/{beta} analysis of environmental and mixed waste samples require repetitive leaching, which is time consuming and generates large volumes of secondary wastes. An alternative to leaching is microwave digestion. In the past. microwave technology has had limited application in the radiochemical laboratory because of restrictions on sample size resulting from vessel pressure limitations. However, new microwave vessel designs allow for pressures on the order of 11 MPa (1500 psi). A procedure is described in which microwave digestion is used to prepare environmental soil samples for gross {alpha}/{beta} analysis. Results indicate that the described procedure meets performance requirements for several soil types and is equivalent to traditional digestion techniques. No statistical differences at the 95% confidence interval exist between the measurement on samples prepared from the hot plate and microwave digestion procedures for those soils tested. Moreover, microwave digestion allows samples to be prepared in a fraction of the time with significantly less acid and with lower potential of cross-contamination. In comparison to the traditional hot plate method, the waste volumes required for the microwave procedure are a factor of 10 lower, while the analyst time for sample processing is at least a factor of three lower.

  19. Particle Size Distribution Data From Existing Boreholes at the Immobilized Low-Activity Waste Site

    SciTech Connect (OSTI)

    Valenta, Michelle M.; Martin, Maria B.; Moreno, Jorge R.; Ferri, Rosalie M.; Horton, Duane G.; Reidel, Stephen P.

    2000-09-25T23:59:59.000Z

    This report provides particle size distribution data for samples near the Immobilized Low-Activity Waste (ILAW) Site that were archived in the Hanford Geotechnical Sample Library. Seventy-nine sediment samples were analyzed from four boreholes. Samples were collected from every ten feet in the boreholes. Eightly percent of the samples were classified as slightly gravelly sand. Fifteen percent were classified as gravelly sand, gravelly silty sand, or sandy gravels. These data indicate that the particle size of the sediment is consistent across the ILAW site and is dominated by sand in the upper part of the Hanford formation with more gravel rich units in the lower part.

  20. Radioactive Demonstration Of Mineralized Waste Forms Made From Hanford Low Activity Waste (Tank SX-105 And AN-103) By Fluidized Bed Steam Reformation

    SciTech Connect (OSTI)

    Jantzen, Carol; Herman, Connie; Crawford, Charles; Bannochie, Christopher; Burket, Paul; Daniel, Gene; Cozzi, Alex; Nash, Charles; Miller, Donald; Missimer, David

    2014-01-10T23:59:59.000Z

    One of the immobilization technologies under consideration as a Supplemental Treatment for Hanford’s Low Activity Waste (LAW) is Fluidized Bed Steam Reforming (FBSR). The FBSR technology forms a mineral waste form at moderate processing temperatures thus retaining and atomically bonding the halides, sulfates, and technetium in the mineral phases (nepheline, sodalite, nosean, carnegieite). Additions of kaolin clay are used instead of glass formers and the minerals formed by the FBSR technology offers (1) atomic bonding of the radionuclides and constituents of concern (COC) comparable to glass, (2) short and long term durability comparable to glass, (3) disposal volumes comparable to glass, and (4) higher Na2O and SO{sub 4} waste loadings than glass. The higher FBSR Na{sub 2}O and SO{sub 4} waste loadings contribute to the low disposal volumes but also provide for more rapid processing of the LAW. Recent FBSR processing and testing of Hanford radioactive LAW (Tank SX-105 and AN-103) waste is reported and compared to previous radioactive and non-radioactive LAW processing and testing.

  1. Reproduced with permission of the copyright owner. Further reproduction prohibited without permission. Overcoming tunnel vision: Redirecting the U.S. high-level nuclear waste program

    E-Print Network [OSTI]

    Kammen, Daniel M.

    permission. Overcoming tunnel vision: Redirecting the U.S. high-level nuclear waste program James Flynn

  2. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    SciTech Connect (OSTI)

    Gdowski, G.E.; Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

    1988-08-01T23:59:59.000Z

    Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of containers for disposal of high-level radioactive waste. This waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, they must be retrievable from the disposal site. Shortly after the containers are emplaced in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of the high-level waste. This volume surveys the available data on oxidation and corrosion of the iron- to nickel-based austenitic materials (Types 304L and 316L stainless steels and Alloy 825) and the copper-based alloy materials (CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni)), which are the present candidates for fabrication of the containers. Studies that provided a large amount of data are highlighted, and those areas in which little data exists are identified. Examples of successful applications of these materials are given. On the basis of resistance to oxidation and general corrosion, the austenitic materials are ranked as follows: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is as follows: CDA 715 and CDA 613 (both best), and CDA 102 (worst). 110 refs., 30 figs., 13 tabs.

  3. Statement of Work (SOW) for FY 2001 to FY 2006 for the Hanford Low Activity Tank Waste Performance Assessment Program

    SciTech Connect (OSTI)

    PUIGH, R.J.

    2000-07-25T23:59:59.000Z

    This document describes the tasks included in the Hanford Low-Activity Tank Waste Performance Assessment activity though the close of the project in 2028. Near-term (2001-2006) tasks are described in detail, while tasks further in the future are simply grouped by year. The major tasks are displayed in the table provided. The major goals of the performance assessment activity are to provide the technical basis for the Department of Energy to continue to authorize the construction of disposal facilities, the onsite disposal of immobilized low-activity Hanford tank waste in those facilities, and the closure of the disposal facilities. Other significant goals are to provide the technical basis for the setting of the specifications of the immobilized waste and to support permitting of the disposal facilities.

  4. Methods of calculating the post-closure performance of high-level waste repositories

    SciTech Connect (OSTI)

    Ross, B. (ed.)

    1989-02-01T23:59:59.000Z

    This report is intended as an overview of post-closure performance assessment methods for high-level radioactive waste repositories and is designed to give the reader a broad sense of the state of the art of this technology. As described here, ''the state of the art'' includes only what has been reported in report, journal, and conference proceedings literature through August 1987. There is a very large literature on the performance of high-level waste repositories. In order to make a review of this breadth manageable, its scope must be carefully defined. The essential principle followed is that only methods of calculating the long-term performance of waste repositories are described. The report is organized to reflect, in a generalized way, the logical order to steps that would be taken in a typical performance assessment. Chapter 2 describes ways of identifying scenarios and estimating their probabilities. Chapter 3 presents models used to determine the physical and chemical environment of a repository, including models of heat transfer, radiation, geochemistry, rock mechanics, brine migration, radiation effects on chemistry, and coupled processes. The next two chapters address the performance of specific barriers to release of radioactivity. Chapter 4 treats engineered barriers, including containers, waste forms, backfills around waste packages, shaft and borehole seals, and repository design features. Chapter 5 discusses natural barriers, including ground water systems and stability of salt formations. The final chapters address optics of general applicability to performance assessment models. Methods of sensitivity and uncertainty analysis are described in Chapter 6, and natural analogues of repositories are treated in Chapter 7. 473 refs., 19 figs., 2 tabs.

  5. Granite disposal of U.S. high-level radioactive waste.

    SciTech Connect (OSTI)

    Freeze, Geoffrey A.; Mariner, Paul E.; Lee, Joon H.; Hardin, Ernest L.; Goldstein, Barry; Hansen, Francis D.; Price, Ronald H.; Lord, Anna Snider

    2011-08-01T23:59:59.000Z

    This report evaluates the feasibility of disposing U.S. high-level radioactive waste in granite several hundred meters below the surface of the earth. The U.S. has many granite formations with positive attributes for permanent disposal. Similar crystalline formations have been extensively studied by international programs, two of which, in Sweden and Finland, are the host rocks of submitted or imminent repository license applications. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in granite media. In this report we develop scoping performance analyses, based on the applicable features, events, and processes (FEPs) identified by international investigators, to support generic conclusions regarding post-closure safety. Unlike the safety analyses for disposal in salt, shale/clay, or deep boreholes, the safety analysis for a mined granite repository depends largely on waste package preservation. In crystalline rock, waste packages are preserved by the high mechanical stability of the excavations, the diffusive barrier of the buffer, and favorable chemical conditions. The buffer is preserved by low groundwater fluxes, favorable chemical conditions, backfill, and the rigid confines of the host rock. An added advantage of a mined granite repository is that waste packages would be fairly easy to retrieve, should retrievability be an important objective. The results of the safety analyses performed in this study are consistent with the results of comprehensive safety assessments performed for sites in Sweden, Finland, and Canada. They indicate that a granite repository would satisfy established safety criteria and suggest that a small number of FEPs would largely control the release and transport of radionuclides. In the event the U.S. decides to pursue a potential repository in granite, a detailed evaluation of these FEPs would be needed to inform site selection and safety assessment.

  6. THE STRUCTURAL CHEMISTRY OF MOLYBDENUM IN MODEL HIGH LEVEL NUCLEAR WASTE GLASSES, INVESTIGATED BY MO K-EDGE X-RAY ABSORPTION

    E-Print Network [OSTI]

    Sheffield, University of

    THE STRUCTURAL CHEMISTRY OF MOLYBDENUM IN MODEL HIGH LEVEL NUCLEAR WASTE GLASSES, INVESTIGATED of molybdenum in model UK high level nuclear waste glasses was investigated by X-ray Absorption Spectroscopy (XAS). Molybdenum K-edge XAS data were acquired from several inactive simulant high level nuclear waste

  7. DURABILITY TESTING OF FLUIDIZED BED STEAM REFORMER (FBSR) WASTE FORMS

    SciTech Connect (OSTI)

    Jantzen, C

    2006-01-06T23:59:59.000Z

    Fluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium aqueous radioactive wastes. The addition of clay and a catalyst as co-reactants converts high sodium aqueous low activity wastes (LAW) such as those existing at the Hanford and Idaho DOE sites to a granular ''mineralized'' waste form that may be made into a monolith form if necessary. Simulant Hanford and Idaho high sodium wastes were processed in a pilot scale FBSR at Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low-activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium-bearing waste (SBW). The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The durability of the FBSR waste form products was tested in order to compare the measured durability to previous FBSR waste form testing on Hanford Envelope C waste forms that were made by THOR Treatment Technologies (TTT) and to compare the FBSR durability to vitreous LAW waste forms, specifically the Hanford low activity waste (LAW) glass known as the Low-activity Reference Material (LRM). The durability of the FBSR waste form is comparable to that of the LRM glass for the test responses studied.

  8. An Istrument for Measuring the TRU Concentration in High-Level Liquid Waste

    SciTech Connect (OSTI)

    Brodzinski, Ronald L.; Craig, R. A.; Fink, Samuel D.; Hensley, Walter K.; Holt, Noah O.; Knopf, Michael A.; Lepel, Elwood A.; Mullen, O Dennis; Salaymeh, Saleem R.; Samuel, Todd J.; Smart, John E.; Tinker, Michael R.; Walker, Darrell D.

    2005-02-01T23:59:59.000Z

    An online monitor has been designed, built, and tested, which is capable of measuring the residual transuranic concentrations in processed high-level wastes with a detection limit of 370 Bq/ml (10 nCi/ml) in less than six hours. The monitor measures the neutrons produced by the transuranics, primarily via (?,n) reactions, in the presence of gamma-ray fields up to 1 Sv/h (100 R/h). The optimum design was determined by Monte Carlo modeling and then tempered with practical engineering and cost considerations. Correct operation of the monitor was demonstrated in a hot cell utilizing an actual sample of high-level waste. Results of that demonstration are given, and suggestions for improvements in the next generation system are discussed.

  9. West Valley high-level nuclear waste glass development: a statistically designed mixture study

    SciTech Connect (OSTI)

    Chick, L.A.; Bowen, W.M.; Lokken, R.O.; Wald, J.W.; Bunnell, L.R.; Strachan, D.M.

    1984-10-01T23:59:59.000Z

    The first full-scale conversion of high-level commercial nuclear wastes to glass in the United States will be conducted at West Valley, New York, by West Valley Nuclear Services Company, Inc. (WVNS), for the US Department of Energy. Pacific Northwest Laboratory (PNL) is supporting WVNS in the design of the glass-making process and the chemical formulation of the glass. This report describes the statistically designed study performed by PNL to develop the glass composition recommended for use at West Valley. The recommended glass contains 28 wt% waste, as limited by process requirements. The waste loading and the silica content (45 wt%) are similar to those in previously developed waste glasses; however, the new formulation contains more calcium and less boron. A series of tests verified that the increased calcium results in improved chemical durability and does not adversely affect the other modeled properties. The optimization study assessed the effects of seven oxide components on glass properties. Over 100 melts combining the seven components into a wide variety of statistically chosen compositions were tested. Viscosity, electrical conductivity, thermal expansion, crystallinity, and chemical durability were measured and empirically modeled as a function of the glass composition. The mathematical models were then used to predict the optimum formulation. This glass was tested and adjusted to arrive at the final composition recommended for use at West Valley. 56 references, 49 figures, 18 tables.

  10. DELPHI expert panel evaluation of Hanford high level waste tank failure modes and release quantities

    SciTech Connect (OSTI)

    Dunford, G.L.; Han, F.C.

    1996-09-30T23:59:59.000Z

    The Failure Modes and Release Quantities of the Hanford High Level Waste Tanks due to postulated accident loads were established by a DELPHI Expert Panel consisting of both on-site and off-site experts in the field of Structure and Release. The Report presents the evaluation process, accident loads, tank structural failure conclusion reached by the panel during the two-day meeting.

  11. Safety analysis report vitrified high level waste type B shipping cask

    SciTech Connect (OSTI)

    NONE

    1995-03-01T23:59:59.000Z

    This Safety Analysis Report describes the design, analyses, and principle features of the Vitrified High Level Waste (VHLW) Cask. In preparing this report a detailed evaluation of the design has been performed to ensure that all safety, licensing, and operational goals for the cask and its associated Department of Energy program can be met. The functions of this report are: (1) to fully document that all functional and regulatory requirements of 10CFR71 can be met by the package; and (2) to document the design and analyses of the cask for review by the Nuclear Regulatory Commission. The VHLW Cask is the reusable shipping package designed by GNSI under Department of Energy contract DE-AC04-89AL53-689 for transportation of Vitrified High Level Waste, and to meet the requirements for certification under 10CFR71 for a Type B(U) package. The VHLW cask has been designed as packaging for transport of canisters of Vitrified High Level Waste solidified at Department of Energy facilities.

  12. Comparison of selected foreign plans and practices for spent fuel and high-level waste management

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.; Lakey, L.T.; Johnson, A.B. Jr.; Hazelton, R.F.; Bradley, D.J.

    1990-04-01T23:59:59.000Z

    This report describes the major parameters for management of spent nuclear fuel and high-level radioactive wastes in selected foreign countries as of December 1989 and compares them with those in the United States. The foreign countries included in this study are Belgium, Canada, France, the Federal Republic of Germany, Japan, Sweden, Switzerland, and the United Kingdom. All the countries are planning for disposal of spent fuel and/or high-level wastes in deep geologic repositories. Most countries (except Canada and Sweden) plan to reprocess their spent fuel and vitrify the resultant high-level liquid wastes; in comparison, the US plans direct disposal of spent fuel. The US is planning to use a container for spent fuel as the primary engineered barrier. The US has the most developed repository concept and has one of the earliest scheduled repository startup dates. The repository environment presently being considered in the US is unique, being located in tuff above the water table. The US also has the most prescriptive regulations and performance requirements for the repository system and its components. 135 refs., 8 tabs.

  13. RESULTS OF THE FY09 ENHANCED DOE HIGH LEVEL WASTE MELTER THROUGHPUT STUDIES AT SRNL

    SciTech Connect (OSTI)

    Johnson, F.; Edwards, T.

    2010-06-23T23:59:59.000Z

    High-level waste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). The objective of this task is to develop data, assess property models, and refine or develop the necessary models to support increased WL of HLW at SRS. It is a continuation of the studies initiated in FY07, but is under the specific guidance of a Task Change Request (TCR)/Work Authorization received from DOE headquarters (Project Number RV071301). Using the data generated in FY07, FY08 and historical data, two test matrices (60 glasses total) were developed at the Savannah River National Laboratory (SRNL) in order to generate data in broader compositional regions. These glasses were fabricated and characterized using chemical composition analysis, X-ray Diffraction (XRD), viscosity, liquidus temperature (TL) measurement and durability as defined by the Product Consistency Test (PCT). The results of this study are summarized below: (1) In general, the current durability model predicts the durabilities of higher waste loading glasses quite well. A few of the glasses exhibited poorer durability than predicted. (2) Some of the glasses exhibited anomalous behavior with respect to durability (normalized leachate for boron (NL [B])). The quenched samples of FY09EM21-02, -07 and -21 contained no nepheline or other wasteform affecting crystals, but have unacceptable NL [B] values (> 10 g/L). The ccc sample of FY09EM21-07 has a NL [B] value that is more than one half the value of the quenched sample. These glasses also have lower concentrations of Al{sub 2}O{sub 3} and SiO{sub 2}. (3) Five of the ccc samples (EM-13, -14, -15, -29 and -30) completely crystallized with both magnetite and nepheline, and still had extremely low NL [B] values. These particular glasses have more CaO present than any of the other glasses in the matrix. It appears that while all of the glasses contain nepheline, the NL [B] values decrease as the CaO concentration increases from 2.3 wt% to 4.3 wt%. A different form of nepheline may be created at higher concentrations of CaO that does not significantly reduce glass durability. (4) The T{sub L} model appears to be under-predicting the measured values of higher waste loading glasses. Trends in T{sub L} with composition are not evident in the data from these studies. (5) A small number of glasses in the FY09 matrix have measured viscosities that are much lower than the viscosity range over which the current model was developed. The decrease in viscosity is due to a higher concentration of non-bridging oxygens (NBO). A high iron concentration is the cause of the increase in NBO. Durability, viscosity and T{sub L} data collected during FY07 and FY09 that specifically targeted higher waste loading glasses was compiled and assessed. It appears that additional data may be required to expand the coverage of the T{sub L} and viscosity models for higher waste loading glasses. In general, the compositional regions of the higher waste loading glasses are very different than those used to develop these models. On the other hand, the current durability model seems to be applicable to the new data. At this time, there is no evidence to modify this model; however additional experimental studies should be conducted to determine the cause of the anomalous durability data.

  14. Utilization of high-carbohydrate food wastes as the feedstock for degradable plastics

    SciTech Connect (OSTI)

    Tsai, S.P.; Coleman, R.D.; Tsai, TenLin S.; Bonsignore, P.V.

    1989-01-01T23:59:59.000Z

    Wastestreams from food processing industries have become an economic burden as well as a serious environmental problem. In the United States, billions of pounds of potato processed each year is typically discarded or sold as cattle feed at $3-6/ton. For large food processing plants, removal of more than 1 million gallons of waste/day/plant is required. As a potential solution to this economic and environmental problem, Argonne National Laboratory is developing technology that (1) bioconverts existing food processing wastestream into lactic acid, and (2) utilizes lactic acid for making environmentally safe, degradable plastics. Although the initial substrate for Argonne's process development is potato waste, the process will be applied to many other high-carbohydrate food wastes. Argonne has developed a process to bioconvert greater than 90% of the fermentable starch in solid potato waste to glucose. Lactic acid is produced from glucose via fermentation and subsequently recovered/purified for plastic synthesis. A continuous lactic acid fermentation and recovery process has been designed. Batch fermentation data showed good cell growth and excellent yields (greater than 95%) of lactic acid production from the hydrolyzed potato waste. Three product recovery processes (electrodialysis, liquid-liquid extraction, and esterification) are being evaluated. Plastics containing lactic acid can be designed to have various mechanical properties and degradation rates. Argonne is developing lactic acid plastics that have some novel features. These environmentally-safe, degradable plastics have many attractive applications such as composting bags and agriculture mulch films. Other potential applications of lactic acid polymers include programmable pesticide and fertilizer delivery systems.

  15. Progress in High-Level Waste Tank Cleaning at the Idaho National Environmental and Engineering Laboratory

    SciTech Connect (OSTI)

    Lockie, K. A.; McNaught, W. B.

    2002-02-26T23:59:59.000Z

    The Department of Energy Idaho Operations Office (DOE-ID) is making preparations to close two underground high-level waste (HLW) storage tanks at the Idaho National Engineering and Environmental Laboratory (INEEL) to meet Resource Conservation and Recovery Act (RCRA) regulations and Department of Energy (DOE) orders. Closure of these two tanks is scheduled for 2004 as the first phase in closure of the eleven 300,000 gallon tanks currently in service at the Idaho Nuclear Technology and Engineering Center (INTEC). Design, development, and deployment of a remotely operated tank cleaning system were completed in August 2001. The system incorporates many commercially available components, which have been adapted for application in cleaning high-level waste tanks. The system also uses existing waste transfer technology (steam-jets) to remove tank heel solids from the tank bottoms during the cleaning operations. By using this existing transfer system and commercially available equipment, the cost of developing custom designed cleaning equipment can be avoided. Remotely operated directional spray nozzles, automatic rotating wash balls, video monitoring equipment, decontamination spray-rings, and tank specific access interface devices have been integrated to provide a system that efficiently cleans tank walls and heel solids in an acidic, radioactive environment. This system is also compliant with operational and safety performance requirements at INTEC. Through the deployment of the tank cleaning system, the INEEL High Level Waste Program has demonstrated the capability to clean tanks to meet RCRA clean closure standards and DOE closure performance measures. The tank cleaning system deployed at the INTEC offers unique advantages over other approaches evaluated at the INEEL and throughout the DOE Complex. The system's ability to agitate and homogenize the tank heel sludge will simplify verification-sampling techniques and reduce the total quantity of samples required to demonstrate compliance with the performance standards. This will reduce tank closure budget requirements and improve closure-planning schedules.

  16. Production of high specific activity silicon-32

    DOE Patents [OSTI]

    Phillips, Dennis R. (Los Alamos, NM); Brzezinski, Mark A. (Santa Barbara, CA)

    1994-01-01T23:59:59.000Z

    A process for preparation of silicon-32 is provide and includes contacting an irradiated potassium chloride target, including spallation products from a prior irradiation, with sufficient water, hydrochloric acid or potassium hydroxide to form a solution, filtering the solution, adjusting pH of the solution to from about 5.5 to about 7.5, admixing sufficient molybdate-reagent to the solution to adjust the pH of the solution to about 1.5 and to form a silicon-molybdate complex, contacting the solution including the silicon-molybdate complex with a dextran-based material, washing the dextran-based material to remove residual contaminants such as sodium-22, separating the silicon-molybdate complex from the dextran-based material as another solution, adding sufficient hydrochloric acid and hydrogen peroxide to the solution to prevent reformation of the silicon-molybdate complex and to yield an oxidization state of the molybdate adapted for subsequent separation by an anion exchange material, contacting the solution with an anion exchange material whereby the molybdate is retained by the anion exchange material and the silicon remains in solution, and optionally adding sufficient alkali metal hydroxide to adjust the pH of the solution to about 12 to 13. Additionally, a high specific activity silicon-32 product having a high purity is provided.

  17. Radioactive Waste Management (Minnesota)

    Broader source: Energy.gov [DOE]

    This section regulates the transportation and disposal of high-level radioactive waste in Minnesota, and establishes a Nuclear Waste Council to monitor the federal high-level radioactive waste...

  18. Hanford High-Level Waste Vitrification Program at the Pacific Northwest National Laboratory: technology development - annotated bibliography

    SciTech Connect (OSTI)

    Larson, D.E.

    1996-09-01T23:59:59.000Z

    This report provides a collection of annotated bibliographies for documents prepared under the Hanford High-Level Waste Vitrification (Plant) Program. The bibliographies are for documents from Fiscal Year 1983 through Fiscal Year 1995, and include work conducted at or under the direction of the Pacific Northwest National Laboratory. The bibliographies included focus on the technology developed over the specified time period for vitrifying Hanford pretreated high-level waste. The following subject areas are included: General Documentation; Program Documentation; High-Level Waste Characterization; Glass Formulation and Characterization; Feed Preparation; Radioactive Feed Preparation and Glass Properties Testing; Full-Scale Feed Preparation Testing; Equipment Materials Testing; Melter Performance Assessment and Evaluations; Liquid-Fed Ceramic Melter; Cold Crucible Melter; Stirred Melter; High-Temperature Melter; Melter Off-Gas Treatment; Vitrification Waste Treatment; Process, Product Control and Modeling; Analytical; and Canister Closure, Decontamination, and Handling

  19. TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 2 DF WASTE LINE REMOVAL, BNL

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-07-09T23:59:59.000Z

    5098-SR-02-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 2 DF WASTE LINE REMOVAL, BROOKHAVEN NATIONAL LABORATORY

  20. What are Spent Nuclear Fuel and High-Level Radioactive Waste ?

    SciTech Connect (OSTI)

    DOE

    2002-12-01T23:59:59.000Z

    Spent nuclear fuel and high-level radioactive waste are materials from nuclear power plants and government defense programs. These materials contain highly radioactive elements, such as cesium, strontium, technetium, and neptunium. Some of these elements will remain radioactive for a few years, while others will be radioactive for millions of years. Exposure to such radioactive materials can cause human health problems. Scientists worldwide agree that the safest way to manage these materials is to dispose of them deep underground in what is called a geologic repository.

  1. Conceptual modular description of the high-level waste management system for system studies model development

    SciTech Connect (OSTI)

    McKee, R.W.; Young, J.R.; Konzek, G.J.

    1992-08-01T23:59:59.000Z

    This document presents modular descriptions of possible alternative components of the federal high-level radioactive waste management system and the procedures for combining these modules to obtain descriptions for alternative configurations of that system. The 20 separate system component modules presented here can be combined to obtain a description of any of the 17 alternative system configurations (i.e., scenarios) that were evaluated in the MRS Systems Studies program (DOE 1989a). First-approximation descriptions of other yet-undefined system configurations could also be developed for system study purposes from this database. The descriptions include, in a modular format, both functional descriptions of the processes in the waste management system, plus physical descriptions of the equipment and facilities necessary for performance of those functions.

  2. Thermal-mechanical modeling of deep borehole disposal of high-level radioactive waste.

    SciTech Connect (OSTI)

    Arnold, Bill Walter; Hadgu, Teklu

    2010-12-01T23:59:59.000Z

    Disposal of high-level radioactive waste, including spent nuclear fuel, in deep (3 to 5 km) boreholes is a potential option for safely isolating these wastes from the surface and near-surface environment. Existing drilling technology permits reliable and cost-effective construction of such deep boreholes. Conditions favorable for deep borehole disposal in crystalline basement rocks, including low permeability, high salinity, and geochemically reducing conditions, exist at depth in many locations, particularly in geologically stable continental regions. Isolation of waste depends, in part, on the effectiveness of borehole seals and potential alteration of permeability in the disturbed host rock surrounding the borehole. Coupled thermal-mechanical-hydrologic processes induced by heat from the radioactive waste may impact the disturbed zone near the borehole and borehole wall stability. Numerical simulations of the coupled thermal-mechanical response in the host rock surrounding the borehole were conducted with three software codes or combinations of software codes. Software codes used in the simulations were FEHM, JAS3D, Aria, and Adagio. Simulations were conducted for disposal of spent nuclear fuel assemblies and for the higher heat output of vitrified waste from the reprocessing of fuel. Simulations were also conducted for both isotropic and anisotropic ambient horizontal stress in the host rock. Physical, thermal, and mechanical properties representative of granite host rock at a depth of 4 km were used in the models. Simulation results indicate peak temperature increases at the borehole wall of about 30 C and 180 C for disposal of fuel assemblies and vitrified waste, respectively. Peak temperatures near the borehole occur within about 10 years and decline rapidly within a few hundred years and with distance. The host rock near the borehole is placed under additional compression. Peak mechanical stress is increased by about 15 MPa (above the assumed ambient isotropic stress of 100 MPa) at the borehole wall for the disposal of fuel assemblies and by about 90 MPa for vitrified waste. Simulated peak volumetric strain at the borehole wall is about 420 and 2600 microstrain for the disposal of fuel assemblies and vitrified waste, respectively. Stress and volumetric strain decline rapidly with distance from the borehole and with time. Simulated peak stress at and parallel to the borehole wall for the disposal of vitrified waste with anisotropic ambient horizontal stress is about 440 MPa, which likely exceeds the compressive strength of granite if unconfined by fluid pressure within the borehole. The relatively small simulated displacements and volumetric strain near the borehole suggest that software codes using a nondeforming grid provide an adequate approximation of mechanical deformation in the coupled thermal-mechanical model. Additional modeling is planned to incorporate the effects of hydrologic processes coupled to thermal transport and mechanical deformation in the host rock near the heated borehole.

  3. Probabilistic safety assessment for Hanford high-level waste tank 241-SY-101

    SciTech Connect (OSTI)

    MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W. [Los Alamos National Lab., NM (United States)] [Los Alamos National Lab., NM (United States); Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J. [PLG, Inc., Newport Beach, CA (United States)] [PLG, Inc., Newport Beach, CA (United States)

    1994-05-01T23:59:59.000Z

    Los Alamos National Laboratory (Los Alamos) is performing a comprehensive probabilistic safety assessment (PSA), which will include consideration of external events for the 18 tank farms at the Hanford Site. This effort is sponsored by the Department of Energy (DOE/EM, EM-36). Even though the methodology described herein will be applied to the entire tank farm, this report focuses only on the risk from the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases ({open_quotes}burps{close_quotes}) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed first because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is being conducted in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. At the Hanford Site there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/salt cake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is about 60 million gal., which contains approximately 120 million Ci of radioactivity.

  4. Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 2: Appendices

    SciTech Connect (OSTI)

    Rechard, R.P. [ed.

    1993-12-01T23:59:59.000Z

    This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency`s Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

  5. US DOE-AECL cooperative program for development of high-level radioactive waste container fabrication, closure, and inspection techniques

    SciTech Connect (OSTI)

    Russell, E.W.

    1990-06-01T23:59:59.000Z

    The US Department of Energy (DOE) and Atomic Energy of Canada Limited (AECL) plan to initiate a cooperative research program on development of manufacturing processes for high-level radioactive waste containers. This joint program will benefit both countries in the development of processes for the fabrication, final closure in a hot-cell, and certification of the containers. Program activity objectives can be summarized as follows: to support the selection of suitable container fabrication, final closure, and inspection techniques for the candidate materials and container designs that are under development or are being considered in the US and Canadian repository programs; and to investigate these techniques for alternate materials and/or container designs, to be determined in future optimization studies relating to long-term performance of the waste packages. The program participants will carry out this work in a conditional phased approach, and the scope of work for subsequent years will evolve subject to developments in earlier years. The overall term of this cooperative program is planned to run roughly three years. 5 refs., 2 tabs.

  6. Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.

    SciTech Connect (OSTI)

    Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

    2011-03-01T23:59:59.000Z

    This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are needed for repository modeling are severely lacking. In addition, most of existing reactive transport codes were developed for non-radioactive contaminants, and they need to be adapted to account for radionuclide decay and in-growth. The accessibility to the source codes is generally limited. Because the problems of interest for the Waste IPSC are likely to result in relatively large computational models, a compact memory-usage footprint and a fast/robust solution procedure will be needed. A robust massively parallel processing (MPP) capability will also be required to provide reasonable turnaround times on the analyses that will be performed with the code. A performance assessment (PA) calculation for a waste disposal system generally requires a large number (hundreds to thousands) of model simulations to quantify the effect of model parameter uncertainties on the predicted repository performance. A set of codes for a PA calculation must be sufficiently robust and fast in terms of code execution. A PA system as a whole must be able to provide multiple alternative models for a specific set of physical/chemical processes, so that the users can choose various levels of modeling complexity based on their modeling needs. This requires PA codes, preferably, to be highly modularized. Most of the existing codes have difficulties meeting these requirements. Based on the gap analysis results, we have made the following recommendations for the code selection and code development for the NEAMS waste IPSC: (1) build fully coupled high-fidelity THCMBR codes using the existing SIERRA codes (e.g., ARIA and ADAGIO) and platform, (2) use DAKOTA to build an enhanced performance assessment system (EPAS), and build a modular code architecture and key code modules for performance assessments. The key chemical calculation modules will be built by expanding the existing CANTERA capabilities as well as by extracting useful components from other existing codes.

  7. Shale disposal of U.S. high-level radioactive waste.

    SciTech Connect (OSTI)

    Sassani, David Carl; Stone, Charles Michael; Hansen, Francis D.; Hardin, Ernest L.; Dewers, Thomas A.; Martinez, Mario J.; Rechard, Robert Paul; Sobolik, Steven Ronald; Freeze, Geoffrey A.; Cygan, Randall Timothy; Gaither, Katherine N.; Holland, John Francis; Brady, Patrick Vane

    2010-05-01T23:59:59.000Z

    This report evaluates the feasibility of high-level radioactive waste disposal in shale within the United States. The U.S. has many possible clay/shale/argillite basins with positive attributes for permanent disposal. Similar geologic formations have been extensively studied by international programs with largely positive results, over significant ranges of the most important material characteristics including permeability, rheology, and sorptive potential. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in shale media. We develop scoping performance analyses, based on the applicable features, events, and processes identified by international investigators, to support a generic conclusion regarding post-closure safety. Requisite assumptions for these analyses include waste characteristics, disposal concepts, and important properties of the geologic formation. We then apply lessons learned from Sandia experience on the Waste Isolation Pilot Project and the Yucca Mountain Project to develop a disposal strategy should a shale repository be considered as an alternative disposal pathway in the U.S. Disposal of high-level radioactive waste in suitable shale formations is attractive because the material is essentially impermeable and self-sealing, conditions are chemically reducing, and sorption tends to prevent radionuclide transport. Vertically and laterally extensive shale and clay formations exist in multiple locations in the contiguous 48 states. Thermal-hydrologic-mechanical calculations indicate that temperatures near emplaced waste packages can be maintained below boiling and will decay to within a few degrees of the ambient temperature within a few decades (or longer depending on the waste form). Construction effects, ventilation, and the thermal pulse will lead to clay dehydration and deformation, confined to an excavation disturbed zone within a few meters of the repository, that can be reasonably characterized. Within a few centuries after waste emplacement, overburden pressures will seal fractures, resaturate the dehydrated zones, and provide a repository setting that strongly limits radionuclide movement to diffusive transport. Coupled hydrogeochemical transport calculations indicate maximum extents of radionuclide transport on the order of tens to hundreds of meters, or less, in a million years. Under the conditions modeled, a shale repository could achieve total containment, with no releases to the environment in undisturbed scenarios. The performance analyses described here are based on the assumption that long-term standards for disposal in clay/shale would be identical in the key aspects, to those prescribed for existing repository programs such as Yucca Mountain. This generic repository evaluation for shale is the first developed in the United States. Previous repository considerations have emphasized salt formations and volcanic rock formations. Much of the experience gained from U.S. repository development, such as seal system design, coupled process simulation, and application of performance assessment methodology, is applied here to scoping analyses for a shale repository. A contemporary understanding of clay mineralogy and attendant chemical environments has allowed identification of the appropriate features, events, and processes to be incorporated into the analysis. Advanced multi-physics modeling provides key support for understanding the effects from coupled processes. The results of the assessment show that shale formations provide a technically advanced, scientifically sound disposal option for the U.S.

  8. Integration of complex-wide mixed low-level waste activities for program acceleration and optimization

    SciTech Connect (OSTI)

    McKenney, D.E.

    1998-01-27T23:59:59.000Z

    In July 1996, the US Department of Energy (DOE) chartered a contractor-led effort to develop a suite of technically defensible, integrated alternatives which would allow the Environmental Management program to accomplish its mission objectives in an accelerated fashion and at a reduced cost. These alternatives, or opportunities, could then be evaluated by DOE and stakeholders for possible implementation, given precursor requirements (regulatory changes, etc.) could be met and benefits to the Complex realized. This contractor effort initially focused on six waste types, one of which was Mixed Low-Level Waste (MLLW). Many opportunities were identified by the contractor team for integrating MLLW activities across the DOE Complex. These opportunities were further narrowed to six that had the most promise for implementation and savings to the DOE Complex. The opportunities include six items: (1) the consolidation of individual site analytical services procurement efforts, (2) the consolidation of individual site MLLW treatment services procurement efforts, (3) establishment of ``de minimus`` radioactivity levels, (4) standardization of characterization requirements, (5) increased utilization of existing DOE treatment facilities, and (6) using a combination of DOE and commercial MLLW disposal capacity. The results of the integration effort showed that by managing MLLW activities across the DOE Complex as a cohesive unit rather than as independent site efforts, the DOE could improve the rate of progress toward meeting its objectives and reduce its overall MLLW program costs. Savings potential for MLLW, if the identified opportunities could be implemented, could total $224 million or more. Implementation of the opportunities also could result in the acceleration of the MLLW ``work off schedule`` across the DOE Complex by five years.

  9. The incorporation of technetium into a representative low-activity waste glass

    SciTech Connect (OSTI)

    Ebert, W.L.; Bakel, A.J.; Bowers, D.L.; Buck, E.C.; Emery, J.W.

    1997-08-01T23:59:59.000Z

    A glass that has been tested to understand the corrosion behavior of waste glasses with high soda contents for immobilizing Hanford incidental wastes has been made by melting crushed glass with either TcO{sub 2} or NaTcO{sub 4} at 1,100--1,300 C. Incorporation of technetium in the glass was affected by solubility or kinetic effects. Metallic technetium inclusions formed in all the TcO{sub 2}-doped glasses. Inclusions also formed in glasses with added NaTcO{sub 4} that were melted at 1,100 C, but a glass melted at 1,200 C did not contain detectable inclusions. The presence of Tc-bearing inclusions complicates the interpretation of results from dissolution tests because of the simultaneous release of technetium from more than one phase, the unknown surface areas of each phase, and the possible incorporation of technetium that is released from one phase into another phase. A glass containing about 0.15 mass % Tc dissolved in the glass is being used in dissolution tests to study the release behavior of technetium.

  10. Load requirements for maintaining structural integrity of Hanford single-shell tanks during waste feed delivery and retrieval activities

    SciTech Connect (OSTI)

    JULYK, L.J.

    1999-09-22T23:59:59.000Z

    This document provides structural load requirements and their basis for maintaining the structural integrity of the Hanford Single-Shell Tanks during waste feed delivery and retrieval activities. The requirements are based on a review of previous requirements and their basis documents as well as load histories with particular emphasis on the proposed lead transfer feed tanks for the privatized vitrification plant.

  11. HIGH LEVEL WASTE MECHANCIAL SLUDGE REMOVAL AT THE SAVANNAH RIVER SITE F TANK FARM CLOSURE PROJECT

    SciTech Connect (OSTI)

    Jolly, R; Bruce Martin, B

    2008-01-15T23:59:59.000Z

    The Savannah River Site F-Tank Farm Closure project has successfully performed Mechanical Sludge Removal (MSR) using the Waste on Wheels (WOW) system for the first time within one of its storage tanks. The WOW system is designed to be relatively mobile with the ability for many components to be redeployed to multiple waste tanks. It is primarily comprised of Submersible Mixer Pumps (SMPs), Submersible Transfer Pumps (STPs), and a mobile control room with a control panel and variable speed drives. In addition, the project is currently preparing another waste tank for MSR utilizing lessons learned from this previous operational activity. These tanks, designated as Tank 6 and Tank 5 respectively, are Type I waste tanks located in F-Tank Farm (FTF) with a capacity of 2,840 cubic meters (750,000 gallons) each. The construction of these tanks was completed in 1953, and they were placed into waste storage service in 1959. The tank's primary shell is 23 meters (75 feet) in diameter, and 7.5 meters (24.5 feet) in height. Type I tanks have 34 vertically oriented cooling coils and two horizontal cooling coil circuits along the tank floor. Both Tank 5 and Tank 6 received and stored F-PUREX waste during their operating service time before sludge removal was performed. DOE intends to remove from service and operationally close (fill with grout) Tank 5 and Tank 6 and other HLW tanks that do not meet current containment standards. Mechanical Sludge Removal, the first step in the tank closure process, will be followed by chemical cleaning. After obtaining regulatory approval, the tanks will be isolated and filled with grout for long-term stabilization. Mechanical Sludge Removal operations within Tank 6 removed approximately 75% of the original 95,000 liters (25,000 gallons). This sludge material was transferred in batches to an interim storage tank to prepare for vitrification. This operation consisted of eleven (11) Submersible Mixer Pump(s) mixing campaigns and multiple intraarea transfers utilizing STPs from July 2006 to August 2007. This operation and successful removal of sludge material meets requirement of approximately 19,000 to 28,000 liters (5,000 to 7,500 gallons) remaining prior to the Chemical Cleaning process. Removal of the last 35% of sludge was exponentially more difficult, as less and less sludge was available to mobilize and the lighter sludge particles were likely removed during the early mixing campaigns. The removal of the 72,000 liters (19,000 gallons) of sludge was challenging due to a number factors. One primary factor was the complex internal cooling coil array within Tank 6 that obstructed mixer discharge jets and impacted the Effective Cleaning Radius (ECR) of the Submersible Mixer Pumps. Minimal access locations into the tank through tank openings (risers) presented a challenge because the available options for equipment locations were very limited. Mechanical Sludge Removal activities using SMPs caused the sludge to migrate to areas of the tank that were outside of the SMP ECR. Various SMP operational strategies were used to address the challenge of moving sludge from remote areas of the tank to the transfer pump. This paper describes in detail the Mechanical Sludge Removal activities and mitigative solutions to cooling coil obstructions and other challenges. The performance of the WOW system and SMP operational strategies were evaluated and the resulting lessons learned are described for application to future Mechanical Sludge Removal operations.

  12. Toxicity mitigation and solidification of municipal solid waste incinerator fly ash using alkaline activated coal ash

    SciTech Connect (OSTI)

    Ivan Diaz-Loya, E. [Alternative Cementitious Binders Laboratory (ACBL), Department of Civil Engineering, Louisiana Tech University, Ruston, LA 71272 (United States); Allouche, Erez N., E-mail: allouche@latech.edu [Alternative Cementitious Binders Laboratory (ACBL), Department of Civil Engineering, Louisiana Tech University, Ruston, LA 71272 (United States); Eklund, Sven; Joshi, Anupam R. [Department of Chemistry, Louisiana Tech University, Ruston, LA 71272 (United States); Kupwade-Patil, Kunal [Alternative Cementitious Binders Laboratory (ACBL), Department of Civil Engineering, Louisiana Tech University, Ruston, LA 71272 (United States)

    2012-08-15T23:59:59.000Z

    Highlights: Black-Right-Pointing-Pointer Incinerator fly ash (IFA) is added to an alkali activated coal fly ash (CFA) matrix. Black-Right-Pointing-Pointer Means of stabilizing the incinerator ash for use in construction applications. Black-Right-Pointing-Pointer Concrete made from IFA, CFA and IFA-CFA mixes was chemically characterized. Black-Right-Pointing-Pointer Environmentally friendly solution to IFA disposal by reducing its toxicity levels. - Abstract: Municipal solid waste (MSW) incineration is a common and effective practice to reduce the volume of solid waste in urban areas. However, the byproduct of this process is a fly ash (IFA), which contains large quantities of toxic contaminants. The purpose of this research study was to analyze the chemical, physical and mechanical behaviors resulting from the gradual introduction of IFA to an alkaline activated coal fly ash (CFA) matrix, as a mean of stabilizing the incinerator ash for use in industrial construction applications, where human exposure potential is limited. IFA and CFA were analyzed via X-ray fluorescence (XRF), X-ray diffraction (XRD) and Inductive coupled plasma (ICP) to obtain a full chemical analysis of the samples, its crystallographic characteristics and a detailed count of the eight heavy metals contemplated in US Title 40 of the Code of Federal Regulations (40 CFR). The particle size distribution of IFA and CFA was also recorded. EPA's Toxicity Characteristic Leaching Procedure (TCLP) was followed to monitor the leachability of the contaminants before and after the activation. Also images obtained via Scanning Electron Microscopy (SEM), before and after the activation, are presented. Concrete made from IFA, CFA and IFA-CFA mixes was subjected to a full mechanical characterization; tests include compressive strength, flexural strength, elastic modulus, Poisson's ratio and setting time. The leachable heavy metal contents (except for Se) were below the maximum allowable limits and in many cases even below the reporting limit. The leachable Chromium was reduced from 0.153 down to 0.0045 mg/L, Arsenic from 0.256 down to 0.132 mg/L, Selenium from 1.05 down to 0.29 mg/L, Silver from 0.011 down to .001 mg/L, Barium from 2.06 down to 0.314 mg/L and Mercury from 0.007 down to 0.001 mg/L. Although the leachable Cd exhibited an increase from 0.49 up to 0.805 mg/L and Pd from 0.002 up to 0.029 mg/L, these were well below the maximum limits of 1.00 and 5.00 mg/L, respectively.

  13. RADIOACTIVE HIGH LEVEL WASTE TANK PITTING PREDICTIONS: AN INVESTIGATION INTO CRITICAL SOLUTION CONCENTRATIONS

    SciTech Connect (OSTI)

    Hoffman, E.

    2012-11-08T23:59:59.000Z

    A series of cyclic potentiodynamic polarization tests was performed on samples of ASTM A537 carbon steel in support of a probability-based approach to evaluate the effect of chloride and sulfate on corrosion the steel?s susceptibility to pitting corrosion. Testing solutions were chosen to systemically evaluate the influence of the secondary aggressive species, chloride, and sulfate, in the nitrate based, high-level wastes. The results suggest that evaluating the combined effect of all aggressive species, nitrate, chloride, and sulfate, provides a consistent response for determining corrosion susceptibility. The results of this work emphasize the importance for not only nitrate concentration limits, but also chloride and sulfate concentration limits.

  14. Collaboration, Automation, and Information Management at Hanford High Level Radioactive Waste (HLW) Tank Farms

    SciTech Connect (OSTI)

    Aurah, Mirwaise Y.; Roberts, Mark A.

    2013-12-12T23:59:59.000Z

    Washington River Protection Solutions (WRPS), operator of High Level Radioactive Waste (HLW) Tank Farms at the Hanford Site, is taking an over 20-year leap in technology, replacing systems that were monitored with clipboards and obsolete computer systems, as well as solving major operations and maintenance hurdles in the area of process automation and information management. While WRPS is fully compliant with procedures and regulations, the current systems are not integrated and do not share data efficiently, hampering how information is obtained and managed.

  15. Isothermal crystallization kinetics in simulated high-level nuclear waste glass

    SciTech Connect (OSTI)

    Vienna, J.D.; Hrma, P.; Smith, D.E. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-12-31T23:59:59.000Z

    Crystallization kinetics of a simulated high-level waste (HLW) glass were measured and modelled. Kinetics of acmite growth in the standard HW39-4 glass were measured using the isothermal method. A time-temperature-transformation (TTT) diagram was generated from these data. Classical glass-crystal transformation kinetic models were empirically applied to the crystallization data. These models adequately describe the kinetics of crystallization in complex HLW glasses (i.e., RSquared = 0.908). An approach to measurement, fitting, and use of TTT diagrams for prediction of crystallinity in a HLW glass canister is proposed.

  16. Performance assessment overview for subseabed disposal of high level radioactive waste

    SciTech Connect (OSTI)

    Klett, R.D.

    1997-06-01T23:59:59.000Z

    The Subseabed Disposal Project (SDP) was part of an international program that investigated the feasibility of high-level radioactive waste disposal in the deep ocean sediments. This report briefly describes the seven-step iterative performance assessment procedures used in this study and presents representative results of the last iteration. The results of the performance are compared to interim standards developed for the SDP, to other conceptual repositories, and to related metrics. The attributes, limitations, uncertainties, and remaining tasks in the SDP feasibility phase are discussed.

  17. Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results

    SciTech Connect (OSTI)

    Rechard, R.P. [ed.

    1995-03-01T23:59:59.000Z

    This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM.

  18. Secondary Waste Form Down-Selection Data Package—Fluidized Bed Steam Reforming Waste Form

    SciTech Connect (OSTI)

    Qafoku, Nikolla; Westsik, Joseph H.; Strachan, Denis M.; Valenta, Michelle M.; Pires, Richard P.

    2011-09-12T23:59:59.000Z

    The Hanford Site in southeast Washington State has 56 million gallons of radioactive and chemically hazardous wastes stored in 177 underground tanks (ORP 2010). The U.S. Department of Energy (DOE), Office of River Protection (ORP), through its contractors, is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to convert the radioactive and hazardous wastes into stable glass waste forms for disposal. Within the WTP, the pretreatment facility will receive the retrieved waste from the tank farms and separate it into two treated process streams. These waste streams will be vitrified, and the resulting waste canisters will be sent to offsite (high-level waste [HLW]) and onsite (immobilized low-activity waste [ILAW]) repositories. As part of the pretreatment and ILAW processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility (ETF) on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed of in the Integrated Disposal Facility (IDF). To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is developing data packages to support that down-selection. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilizing and solidifying the liquid secondary wastes. At the Hanford Site, the FBSR process is being evaluated as a supplemental technology for treating and immobilizing Hanford LAW radioactive tank waste and for treating secondary wastes from the WTP pretreatment and LAW vitrification processes.

  19. RCRA Part A Permit Application for Waste Management Activities at the Nevada Test Site, Part B Permit Application Hazardous Waste Storage Unit, Nevada Test Site, and Part B Permit Application - Explosives Ordnance Disposal Unit (EODU)

    SciTech Connect (OSTI)

    NSTec Environmental Programs

    2010-06-17T23:59:59.000Z

    The Area 5 Hazardous Waste Storage Unit (HWSU) was established to support testing, research, and remediation activities at the Nevada Test Site (NTS), a large-quantity generator of hazardous waste. The HWSU, located adjacent to the Area 5 Radioactive Waste Management Site (RWMS), is a prefabricated, rigid steel-framed, roofed shelter used to store hazardous nonradioactive waste generated on the NTS. No offsite generated wastes are managed at the HWSU. Waste managed at the HWSU includes the following categories: Flammables/Combustibles; Acid Corrosives; Alkali Corrosives; Oxidizers/Reactives; Toxics/Poisons; and Other Regulated Materials (ORMs). A list of the regulated waste codes accepted for storage at the HWSU is provided in Section B.2. Hazardous wastes stored at the HWSU are stored in U.S. Department of Transportation (DOT) compliant containers, compatible with the stored waste. Waste transfer (between containers) is not allowed at the HWSU and containers remain closed at all times. Containers are stored on secondary containment pallets and the unit is inspected monthly. Table 1 provides the metric conversion factors used in this application. Table 2 provides a list of existing permits. Table 3 lists operational Resource Conservation and Recovery Act (RCRA) units at the NTS and their respective regulatory status.

  20. Final Report - High Level Waste Vitrification System Improvements, VSL-07R1010-1, Rev 0, dated 04/16/07

    SciTech Connect (OSTI)

    Kruger, Albert A.; Gan, H.; Pegg, I. L.; Gong, W.; Champman, C. C.; Joseph, I.; Matlack, K. S.

    2013-11-13T23:59:59.000Z

    This report describes work conducted to support the development and testing of new glass formulations that extend beyond those that have been previously investigated for the Hanford Waste Treatment and Immobilization Plant (WTP). The principal objective was to investigate maximization of the incorporation of several waste components that are expected to limit waste loading and, consequently, high level waste (HLW) processing rates and canister count. The work was performed with four waste compositions specified by the Office of River Protection (ORP); these wastes contain high concentrations of bismuth, chromium, aluminum, and aluminum plus sodium. The tests were designed to identify glass formulations that maximize waste loading while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM100) tests designed to examine the effects of enhanced glass formulations, increased glass processing temperature, increased crystallinity, and feed solids content on waste processing rate and product quality.

  1. Study on the colloids generated from testing of high-level nuclear waste glasses

    SciTech Connect (OSTI)

    Feng, X.; Buck, E.C.; Mertz, C.; Bates, J.K.; Cunnane, J.C.; Chaiko, D.J.

    1993-03-01T23:59:59.000Z

    The generation of colloids in the interaction of high-level nuclear waste glasses with groundwater at 90{degrees}C has been investigated. The stability of the colloidal suspensions has been characterized with respect to salt concentration, pH time, particle size, and zeta potential. The compositions and the morphology of the colloids have also been determined with transmission electron microscopy (TEM). From ourtest results combined with earlier ones, we conclude that the waste glass may contribute to the colloid formation by increasing ion concentration in groundwater, which causes nucleation of colloids; by releasing radionuclides that adsorb onto existing groundwater colloids; and by spalling colloidal-size fragments from the surface layer of the reacted glass. The colloids are silicon-rich particles, such as smectites and uranium silicates. When the salt concentration in the solution is high the colloidal suspensions agglomerate. However, the agglomerated particles can be resuspended if the salt concentration is lowered by dilution with groundwater. The colloids agglomerate quickly after the leachate is cooled to room temperature. Most of the colloids settle out of the solution within a few days at ambient temperature. The isoelectric point is at a pH of approximately 1.0. Between pH 1 and 10.5, the colloids are negatively charged, which suggests that they will deposit readily on, positively charged surfaces. The average particle size islargest at the isoelectric point and is smallest around pH 6.

  2. Study on the colloids generated from testing of high-level nuclear waste glasses

    SciTech Connect (OSTI)

    Feng, X.; Buck, E.C.; Mertz, C.; Bates, J.K.; Cunnane, J.C.; Chaiko, D.J.

    1993-01-01T23:59:59.000Z

    The generation of colloids in the interaction of high-level nuclear waste glasses with groundwater at 90[degrees]C has been investigated. The stability of the colloidal suspensions has been characterized with respect to salt concentration, pH time, particle size, and zeta potential. The compositions and the morphology of the colloids have also been determined with transmission electron microscopy (TEM). From ourtest results combined with earlier ones, we conclude that the waste glass may contribute to the colloid formation by increasing ion concentration in groundwater, which causes nucleation of colloids; by releasing radionuclides that adsorb onto existing groundwater colloids; and by spalling colloidal-size fragments from the surface layer of the reacted glass. The colloids are silicon-rich particles, such as smectites and uranium silicates. When the salt concentration in the solution is high the colloidal suspensions agglomerate. However, the agglomerated particles can be resuspended if the salt concentration is lowered by dilution with groundwater. The colloids agglomerate quickly after the leachate is cooled to room temperature. Most of the colloids settle out of the solution within a few days at ambient temperature. The isoelectric point is at a pH of approximately 1.0. Between pH 1 and 10.5, the colloids are negatively charged, which suggests that they will deposit readily on, positively charged surfaces. The average particle size islargest at the isoelectric point and is smallest around pH 6.

  3. LBNL: High Performance Active Perimeter Building Systems - 2015...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Eleanor Lee, LBNL View the Presentation LBNL: High Performance Active Perimeter Building Systems - 2015 Peer Review More Documents & Publications FLEXLAB LBNL: NYC Office...

  4. Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers

    SciTech Connect (OSTI)

    Vinson, D.W.; Nutt, W.M.; Bullen, D.B. [Iowa State Univ. of Science and Technology, Ames, IA (United States)

    1995-06-01T23:59:59.000Z

    Oxidation and atmospheric corrosion data suggest that addition of Cr provides the greatest improvement in oxidation resistance. Cr-bearing cast irons are resistant to chloride environments and solutions containing strongly oxidizing constituents. Weathering steels, including high content and at least 0.04% Cu, appear to provide adequate resistance to oxidation under temperate conditions. However, data from long-term, high-temperature oxidation studies on weathering steels were not available. From the literature, it appears that the low alloy steels, plain carbon steels, cast steels, and cast irons con-ode at similar rates in an aqueous environment. Alloys containing more than 12% Cr or 36% Ni corrode at a lower rate than plain carbon steels, but pitting may be worse. Short term tests indicate that an alloy of 9Cr-1Mo may result in increased corrosion resistance, however long term data are not available. Austenitic cast irons show the best corrosion resistance. A ranking of total corrosion performance of the materials from most corrosion resistant to least corrosion resistant is: Austenitic Cast Iron; 12% Cr = 36% Ni = 9Cr-1Mo; Carbon Steel = Low Alloy Steels; and Cast Iron. Since the materials to be employed in the Advanced Conceptual Design (ACD) waste package are considered to be corrosion allowance materials, the austenitic cast irons, high Cr steels, high Ni steels and the high Cr-Mo steels should not be considered as candidates for the outer containment barrier. Based upon the oxidation and corrosion data available for carbon steels, low alloy steels, and cast irons, a suitable list of candidate materials for a corrosion allowance outer barrier for an ACD waste package could include, A516, 2.25%Cr -- 1%Mo Steel, and A27.

  5. Proceedings of the 1993 international conference on nuclear waste management and environmental remediation. Volume 2: High level radioactive waste and spent fuel management

    SciTech Connect (OSTI)

    Ahlstroem, P.E.; Chapman, C.C.; Kohout, R.; Marek, J. [eds.

    1993-12-31T23:59:59.000Z

    This conference was held in 1993 in Prague, Czech Republic to provide a forum for exchange of state-of-the-art information on radioactive waste management. Volume 2 contains 109 papers divided into the following sections: recent developments in environmental remediation technologies; decommissioning of nuclear power reactors; environmental restoration site characterization and monitoring; decontamination and decommissioning of other nuclear facilities; prediction of contaminant migration and related doses; treatment of wastes from decontamination and decommissioning operations; management of complex environmental cleanup projects; experiences in actual cleanup actions; decontamination and decommissioning demolition technologies; remediation of obsolete sites from uranium mining and milling; ecological impacts from radioactive environmental contamination; national environmental management regulations--issues and assessments; significant issues and strategies in environmental management; acceptance criteria for very low-level radioactive wastes; processes for public involvement in environmental activities and decisions; recent experiences in public participation activities; established and emerging environmental management organizations; and economic considerations in environmental management. Individual papers have been processed separately for inclusion in the appropriate data bases.

  6. The Waste Isolation Pilot Plant Deep Geological Repository: A Domestic and Global Blueprint for Safe Disposal of High-Level Radioactive Waste - 12081

    SciTech Connect (OSTI)

    Eriksson, Leif G. [Nuclear Waste Dispositions, Winter Park, Florida 32789 (United States); Dials, George E. [B and W Conversion Services, LLC, Lexington, Kentucky 40513 (United States)

    2012-07-01T23:59:59.000Z

    At the end of 2011, the world's first used/spent nuclear fuel and other long-lived high-level radioactive waste (HLW) repository is projected to open in 2020, followed by two more in 2025. The related pre-opening periods will be at least 40 years, as it also would be if USA's candidate HLW-repository is resurrected by 2013. If abandoned, a new HLW-repository site would be needed. On 26 March 1999, USA began disposing long-lived radioactive waste in a deep geological repository in salt at the Waste Isolation Pilot Plant (WIPP) site. The related pre-opening period was less than 30 years. WIPP has since been re-certified twice. It thus stands to reason the WIPP repository is the global proof of principle for safe deep geological disposal of long-lived radioactive waste. It also stands to reason that the lessons learned since 1971 at the WIPP site provide a unique, continually-updated, blueprint for how the pre-opening period for a new HLW repository could be shortened both in the USA and abroad. (authors)

  7. HIGH LEVEL WASTE (HLW) VITRIFICATION EXPERIENCE IN THE US: APPLICATION OF GLASS PRODUCT/PROCESS CONTROL TO OTHERHLW AND HAZARDOUS WASTES

    SciTech Connect (OSTI)

    Jantzen, C; James Marra, J

    2007-09-17T23:59:59.000Z

    Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. At the Savannah River Site (SRS) actual HLW tank waste has successfully been processed to stringent product and process constraints without any rework into a stable borosilicate glass waste since 1996. A unique 'feed forward' statistical process control (SPC) has been used rather than statistical quality control (SQC). In SPC, the feed composition to the melter is controlled prior to vitrification. In SQC, the glass product is sampled after it is vitrified. Individual glass property models form the basis for the 'feed forward' SPC. The property models transform constraints on the melt and glass properties into constraints on the feed composition. The property models are mechanistic and depend on glass bonding/structure, thermodynamics, quasicrystalline melt species, and/or electron transfers. The mechanistic models have been validated over composition regions well outside of the regions for which they were developed because they are mechanistic. Mechanistic models allow accurate extension to radioactive and hazardous waste melts well outside the composition boundaries for which they were developed.

  8. Role of Congress in the High Level Radioactive Waste Odyssey: The Wisdom and Will of the Congress - 13096

    SciTech Connect (OSTI)

    Vieth, Donald L. [DOE/NVOO Project Manager for Yucca Mountain, 1982 thru 1987, 1154 Cheltenham Place, Maineville, OH 45039 (United States)] [DOE/NVOO Project Manager for Yucca Mountain, 1982 thru 1987, 1154 Cheltenham Place, Maineville, OH 45039 (United States); Voegele, Michael D. [Nye County Nuclear Waste Repository Project Office, 7404 Oak Grove Ave, Las Vegas, NV 89117 (United States)] [Nye County Nuclear Waste Repository Project Office, 7404 Oak Grove Ave, Las Vegas, NV 89117 (United States)

    2013-07-01T23:59:59.000Z

    Congress has had a dual role with regard to high level radioactive waste, being involved in both its creation and its disposal. A significant amount of time has passed between the creation of the nation's first high level radioactive waste and the present day. The pace of addressing its remediation has been highly irregular. Congress has had to consider the technical, regulatory, and political issues and all have had specific difficulties. It is a true odyssey framed by an imperative and accountability, by a sense of urgency, by an ability or inability to finish the job and by consequences. Congress had set a politically acceptable course by 1982. However, President Obama intervened in the process after he took office in January 2009. Through the efforts of his Administration, by the end of 2012, the US government has no program to dispose of high level radioactive waste and no reasonable prospect of a repository for high level radioactive waste. It is not obvious how the US government program will be reestablished or who will assume responsibility for leadership. The ultimate criteria for judging the consequences are 1) the outcome of the ongoing NRC's Nuclear Waste Confidence Rulemaking and 2) the concomitant permissibility of nuclear energy supplying electricity from operating reactors in the US. (authors)

  9. An Instrument for Measuring the TRU Concentration in High-Level Liquid Waste

    SciTech Connect (OSTI)

    Brodzinski, Ronald L.; Craig, R A.; Fink, Samuel D.; Hensley, Walter K.; Holt, Noah OA; Knopf, Michael A.; Lepel, Elwood A.; Mullen, O Dennis; Salaymeh, Saleem R.; Samuel, Todd J.; Smart, John E.; Tinker, Mike R.; Walker, D

    2005-02-01T23:59:59.000Z

    An online monitor has been designed, built, and tested that is capable of measuring the residual transuranic concentrations in processed high-level wastes with a detection limit of 370 Bq/ml (10 nCi/ml) in less than six hours. The monitor measures the ({alpha},n) neutrons in the presence of gamma-ray fields up to 1 Sv/h (100 R/h). The optimum design was determined by Monte Carlo modeling and then tempered with practical engineering and cost considerations. A multiplicity counter is used in data acquisition to reject the large fraction of coincident and highly variable cosmic-ray-engendered background events and results in a S/N ratio {approx}1.

  10. Using electrochemical separation to reduce the volume of high-level nuclear waste

    SciTech Connect (OSTI)

    Slater, S.A.; Gay, E.C.

    1998-07-01T23:59:59.000Z

    Argonne National Laboratory (ANL) has developed an electrochemical separation technique called electrorefining that will treat a variety of metallic spent nuclear fuel and reduce the volume of high-level nuclear waste that requires disposal. As part of that effort, ANL has developed a high throughput electrorefiner (HTER) that has a transport rate approximately three times faster than electrorefiners previously developed at ANL. This higher rate is due to the higher electrode surface area, a shorter transport path, and more efficient mixing, which leads to smaller boundary layers about the electrodes. This higher throughput makes electrorefining an attractive option in treating Department of Energy spent nuclear fuels. Experiments have been done to characterize the HTER, and a simulant metallic fuel has been successfully treated. The HTER design and experimental results is discussed.

  11. activity-long living wastes: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the appeal of some scientific disciplines. Sparavigna, A C 2012-01-01 7 Municipal solid waste characteristics and management in Allahabad, India Renewable Energy Websites...

  12. activity-long life waste: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    The heat of an eternally inherent low heating value on the other. Current status of Solid Waste Management The MSW Rules Columbia University 25 Recommendations for developing...

  13. Design and operating features of the high-level waste vitrification system for the West Valley demonstration project

    SciTech Connect (OSTI)

    Siemens, D.H.; Beary, M.M.; Barnes, S.M.; Berger, D.N.; Brouns, R.A.; Chapman, C.C.; Jones, R.M.; Peters, R.D.; Peterson, M.E.

    1986-03-01T23:59:59.000Z

    A liquid-fed joule-heated ceramic melter system is the reference process for immobilization of the high-level liquid waste in the US and several foreign countries. This system has been under development for over ten years at Pacific Northwest Laboratory and other national laboratories operated for the US Department of Energy. Pacific Northwest Laboratory contributed to this research through its Nuclear Waste Treatment Program and used applicable data to design and test melters and related systems using remote handling of simulated radioactive wastes. This report describes the equipment designed in support of the high-level waste vitrification program at West Valley, New York. Pacific Northwest Laboratory worked closely with West Valley Nuclear Services Company to design a liquid-fed ceramic melter, a liquid waste preparation and feed tank and pump, an off-gas treatment scrubber, and an enclosed turntable for positioning the waste canisters. Details of these designs are presented including the rationale for the design features and the alternatives considered.

  14. Technology of high-level nuclear waste disposal. Advances in the science and engineering of the management of high-level nuclear wastes. Volume 1

    SciTech Connect (OSTI)

    Hofmann, P.L.; Breslin, J.J. (eds.)

    1981-01-01T23:59:59.000Z

    The papers in this volume cover the following subjects: waste isolation and the natural geohydrologic system; repository perturbations of the natural system; radionuclide migration through the natural system; and repository design technology. Individual papers are abstracted.

  15. Nondestructive assay and nondestructive examination of remote-handled transuranic waste at the ORNL waste handling and packaging plant

    SciTech Connect (OSTI)

    Schultz, F.J.; Caldwell, J.T. (Oak Ridge National Lab., TN (USA); Pajarito Scientific Corp. (USA))

    1989-01-01T23:59:59.000Z

    The purpose of this investigation is to examine the use of an electron linear accelerator (LINAC) in the performance of nondestructive assay (NDA) and nondestructive examination (NDE) measurements of remote-handled transuranic wastes. The system will be used to perform waste characterization and certification activities at the Oak Ridge National Laboratory's proposed Waste Handling and Packaging Plant. The NDA and NDE technologies which were developed for contact-handled wastes are inadequate to perform such measurements on high gamma and neutron dose-rate wastes. A single LINAC will provide the interrogating fluxes required for both NDA and NDE measurements of the wastes. 11 refs., 6 figs.

  16. Femtosecond Chemically Activated Reactions: Concept of Nonstatistical Activation at High Thermal Energies

    E-Print Network [OSTI]

    Kim, Sang Kyu

    Femtosecond chemical activation of reactions at very high thermal energies, much above the bond energyFemtosecond Chemically Activated Reactions: Concept of Nonstatistical Activation at High Thermal Energies Sang Kyu Kim, Ju Guo, J. Spencer Baskin, and Ahmed H. Zewail* Arthur Amos Noyes Chemical Physics

  17. TWRS retrieval and disposal mission, immobilized high-level waste storage plan

    SciTech Connect (OSTI)

    Calmus, R.B.

    1998-01-07T23:59:59.000Z

    This project plan has a two fold purpose. First, it provides a plan specific to the Hanford Tank Waste Remediation System (TWRS) Immobilized High-Level Waste (EMW) Storage Subproject for the Washington State Department of Ecology (Ecology) that meets the requirements of Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) milestone M-90-01 (Ecology et al. 1996) and is consistent with the project plan content guidelines found in Section 11.5 of the Tri-Party Agreement action plan. Second, it provides an upper tier document that can be used as the basis for future subproject line item construction management plans. The planning elements for the construction management plans are derived from applicable U.S. Department of Energy (DOE) planning guidance documents (DOE Orders 4700.1 (DOE 1992a) and 430.1 (DOE 1995)). The format and content of this project plan are designed to accommodate the plan`s dual purpose. A cross-check matrix is provided in Appendix A to explain where in the plan project planning elements required by Section 11.5 of the Tri-Party Agreement are addressed.

  18. SUMO, System performance assessment for a high-level nuclear waste repository: Mathematical models

    SciTech Connect (OSTI)

    Eslinger, P.W.; Miley, T.B.; Engel, D.W.; Chamberlain, P.J. II

    1992-09-01T23:59:59.000Z

    Following completion of the preliminary risk assessment of the potential Yucca Mountain Site by Pacific Northwest Laboratory (PNL) in 1988, the Office of Civilian Radioactive Waste Management (OCRWM) of the US Department of Energy (DOE) requested the Performance Assessment Scientific Support (PASS) Program at PNL to develop an integrated system model and computer code that provides performance and risk assessment analysis capabilities for a potential high-level nuclear waste repository. The system model that has been developed addresses the cumulative radionuclide release criteria established by the US Environmental Protection Agency (EPA) and estimates population risks in terms of dose to humans. The system model embodied in the SUMO (System Unsaturated Model) code will also allow benchmarking of other models being developed for the Yucca Mountain Project. The system model has three natural divisions: (1) source term, (2) far-field transport, and (3) dose to humans. This document gives a detailed description of the mathematics of each of these three divisions. Each of the governing equations employed is based on modeling assumptions that are widely accepted within the scientific community.

  19. PERFORMANCE OF A BURIED RADIOACTIVE HIGH LEVEL WASTE GLASS AFTER 24 YEARS

    SciTech Connect (OSTI)

    Jantzen, C; Daniel Kaplan, D; Ned Bibler, N; David Peeler, D; John Plodinec, J

    2008-05-05T23:59:59.000Z

    A radioactive high level waste glass was made in 1980 with Savannah River Site (SRS) Tank 15 waste. This glass was buried in the SRS burial ground for 24 years but lysimeter data was only available for the first 8 years. The glass was exhumed and analyzed in 2004. The glass was predicted to be very durable and laboratory tests confirmed the durability response. The laboratory results indicated that the glass was very durable as did analysis of the lysimeter data. Scanning electron microscopy of the glass burial surface showed no significant glass alteration consistent with the results of the laboratory and field tests. No detectable Pu, Am, Cm, Np, or Ru leached from the glass into the surrounding sediment. Leaching of {beta}/{delta} from {sup 90}Sr and {sup 137}Cs in the glass was diffusion controlled. Less than 0.5% of the Cs and Sr in the glass leached into the surrounding sediment, with >99% of the leached radionuclides remaining within 8 centimeters of the glass pellet.

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

    SciTech Connect (OSTI)

    Fox, K. M.

    2014-02-27T23:59:59.000Z

    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 observed in any of the pour stream glass samples. Spinel was observed at the bottom of DWPF Melter 1 as a result of K-3 refractory corrosion. Issues have occurred with accumulation of spinel in the pour spout during periods of operation at higher waste loadings. Given that both DWPF melters were or have been in operation for greater than 8 years, the service life of the melters has far exceeded design expectations. It is possible that the DWPF liquidus temperature approach is conservative, in that it may be possible to successfully operate the melter with a small degree of allowable crystallization in the glass. This could be a viable approach to increasing waste loading in the glass assuming that the crystals are suspended in the melt and swept out through the riser and pour spout. Additional study is needed, and development work for WTP might be leveraged to support a different operating limit for the DWPF. Several recommendations are made regarding considerations that need to be included as part of the WTP crystal tolerant strategy based on the DWPF development work and operational data reviewed here. These include: Identify and consider the impacts of potential heat sinks in the WTP melter and glass pouring system; Consider the contributions of refractory corrosion products, which may serve to nucleate additional crystals leading to further accumulation; Consider volatilization of components from the melt (e.g., boron, alkali, halides, etc.) and determine their impacts on glass crystallization behavior; Evaluate the impacts of glass REDuction/OXidation (REDOX) conditions and the distribution of temperature within the WTP melt pool and melter pour chamber on crystal accumulation rate; Consider the impact of precipitated crystals on glass viscosity; Consider the impact of an accumulated crystalline layer on thermal convection currents and bubbler effectiveness within the melt pool; Evaluate the impact of spinel accumulation on Joule heating of the WTP melt pool; and Include noble metals in glass melt experiments because of their potential to act as nucleation site

  1. Iodine Solubility in Low-Activity Waste Borosilicate Glass at 1000 °C

    SciTech Connect (OSTI)

    Riley, Brian J.; Schweiger, Michael J.; Kim, Dong-Sang; Lukens, Wayne W.; Williams, Benjamin D.; Iovin, Cristian; Rodriguez, Carmen P.; Overman, Nicole R.; Bowden, Mark E.; Dixon, Derek R.; Crum, Jarrod V.; Mccloy, John S.; Kruger, Albert A.

    2014-04-30T23:59:59.000Z

    The purpose of this study was to determine the solubility of iodine in a low-activity waste borosilicate glass when heated inside an evacuated and sealed fused quartz ampoule. The iodine was added to glass frit as KI in quantities of 100–24000 ppm iodine (by mass), each mixture was added to an ampoule, the ampoule was heated at 1000 °C for 2 h and then air quenched. In samples with ?12000 ppm iodine, low viscosity salt phases were observed on the surface of the melts during cooling that solidified into a white coating upon cooling. These salts were identified as mixtures of KI, NaI, and Na2SO4 with X-ray diffraction (XRD). The iodine concentrations in glass specimens were analyzed with inductively-coupled plasma mass spectrometry and the overall iodine solubility was determined to be 10000 ppm by mass. Several crystalline inclusions of iodine sodalite, Na8(AlSiO4)6I2, were observed in the 24000 ppm specimen and were verified with micro-XRD and wavelength dispersive spectroscopy.

  2. Redox-Dependent Solubility of Technetium in Low Activity Waste Glass

    SciTech Connect (OSTI)

    Soderquist, Chuck Z.; Schweiger, Michael J.; Kim, Dong-Sang; Lukens, Wayne W.; Mccloy, John S.

    2014-03-01T23:59:59.000Z

    The solubility of technetium was measured in a Hanford low activity waste glass simulant. The simulant glass was melted, quenched and pulverized to make a stock of powdered glass. A series of glass samples were prepared using the powdered glass and varying amounts of solid potassium pertechnetate. Samples were melted at 1000°C in sealed fused quartz ampoules. After cooling, the bulk glass and the salt phase above the glass (when present) were sampled for physical and chemical characterization. Technetium was found in the bulk glass up to 2000 ppm (using the glass as prepared) and 3000 ppm (using slightly reducing conditions). The chemical form of technetium obtained by x-ray absorption near edge spectroscopy can be mainly assigned to isolated Tc(IV), with a minority of Tc(VII) in some glasses and TcO2 in two glasses. The concentration and speciation of technetium depends on glass redox and amount of technetium added. Solid crystals of pertechnetate salts were found in the salt cake layer that formed at the top of some glasses during the melt.

  3. Implementation of advanced matrix corrections for active interrogation of waste drums using the CTEN instrument

    SciTech Connect (OSTI)

    Melton, S.; Estep, R.; Hollas, C.

    1998-12-31T23:59:59.000Z

    The combined thermal/epithermal neutron instrument (CTEN) was designed at Los Alamos to improve measurement accuracy and mitigate self shielding effects inherent in the differential dieaway technique (DDT). A major goal in this research effort has been the development of a calibration technique that incorporates recently developed matrix and self-shielding corrections using data generated from additional detectors and new acquisition techniques. A comprehensive data set containing both active and passive measurements was generated using 26 different matrices and comprising a total of 1,400 measurements. In all, 31 flux-and-matrix-dependent parameters, 24 positional parameters, two dieaway times, and a correlated ratio were determined from each of the over 1,400 measurements. A reduced list of matrix indicators, prioritized using the alternating conditional expectation (ACE) algorithm, was used to train a neural network using a generalized regression technique (GRNN) to determine matrix- and position-corrected calibration factors. This paper describes the experimental, analytical, and empirical techniques used to determine the corrected calibration factor for an unknown waste drum. Results from a range of cases are compared with those obtained using a mobile DDT instrument and traditional DDT algorithms.

  4. Pilot-scale anaerobic co-digestion of municipal biomass waste and waste activated sludge in China: Effect of organic loading rate

    SciTech Connect (OSTI)

    Liu Xiao, E-mail: liuxiao07@mails.tsinghua.edu.cn [School of Environment, Tsinghua University, Beijing 100084 (China); Wang Wei; Shi Yunchun; Zheng Lei [School of Environment, Tsinghua University, Beijing 100084 (China); Gao Xingbao [Chinese Research Academy of Environmental Sciences, Beijing 100012 (China); Qiao Wei [State Key Laboratory of Heavy Oil Processing, China University of Petroleum, Beijing 102249 (China); Zhou Yingjun [Department of Urban and Environmental Engineering, Graduate School of Engineering, Kyoto University, Katsura, Nisikyo-ku, Kyoto 615-8540 (Japan)

    2012-11-15T23:59:59.000Z

    Highlights: Black-Right-Pointing-Pointer Co-digestion of municipal biomass waste (MBW) and waste activated sludge (WAS) was examined on a pilot-scale reactor. Black-Right-Pointing-Pointer System performance and stability under OLR of 1.2, 2.4, 3.6, 4.8, 6.0 and 8.0 kg VS (m{sup 3} d){sup -1} were analyzed. Black-Right-Pointing-Pointer A maximum methane production rate of 2.94 m{sup 3} (m{sup 3} d){sup -1} was achieved at OLR of 8.0 kg VS (m{sup 3} d){sup -1} and HRT of 15d. Black-Right-Pointing-Pointer With the increasing OLRs, pH values, VS removal rate and methane concentration decreased and VFA increased. Black-Right-Pointing-Pointer The changing of biogas production rate can be a practical approach to monitor and control anaerobic digestion system. - Abstract: The effects of organic loading rate on the performance and stability of anaerobic co-digestion of municipal biomass waste (MBW) and waste activated sludge (WAS) were investigated on a pilot-scale reactor. The results showed that stable operation was achieved with organic loading rates (OLR) of 1.2-8.0 kg volatile solid (VS) (m{sup 3} d){sup -1}, with VS reduction rates of 61.7-69.9%, and volumetric biogas production of 0.89-5.28 m{sup 3} (m{sup 3} d){sup -1}. A maximum methane production rate of 2.94 m{sup 3} (m{sup 3} d){sup -1} was achieved at OLR of 8.0 kg VS (m{sup 3} d){sup -1} and hydraulic retention time of 15 days. With increasing OLRs, the anaerobic reactor showed a decrease in VS removal rate, average pH value and methane concentration, and a increase of volatile fatty acid concentration. By monitoring the biogas production rate (BPR), the anaerobic digestion system has a higher acidification risk under an OLR of 8.0 kg VS (m{sup 3} d){sup -1}. This result remarks the possibility of relating bioreactor performance with BPR in order to better understand and monitor anaerobic digestion process.

  5. Waste reduction at the Savannah River Site

    SciTech Connect (OSTI)

    Stevens, W.E.; Lee, R.A.; Reynolds, R.W.

    1990-01-01T23:59:59.000Z

    The Savannah River Site (SRS) is a key installation for the production and research of nuclear materials for national defense and peace time applications and has been operating a full nuclear fuel cycle since the early 1950s. Wastes generated include high level radioactive, transuranic, low level radioactive, hazardous, mixed, sanitary, and aqueous wastes. Much progress has been made during the last several years to reduce these wastes including management systems, characterization, and technology programs. The reduction of wastes generated and the proper handling of the wastes have always been a part of the Site's operation. This paper summarizes the current status and future plans with respect to waste reduction to waste reduction and reviews some specific examples of successful activities.

  6. Waste reduction at the Savannah River Site

    SciTech Connect (OSTI)

    Stevens, W.E.; Lee, R.A.; Reynolds, R.W.

    1990-12-31T23:59:59.000Z

    The Savannah River Site (SRS) is a key installation for the production and research of nuclear materials for national defense and peace time applications and has been operating a full nuclear fuel cycle since the early 1950s. Wastes generated include high level radioactive, transuranic, low level radioactive, hazardous, mixed, sanitary, and aqueous wastes. Much progress has been made during the last several years to reduce these wastes including management systems, characterization, and technology programs. The reduction of wastes generated and the proper handling of the wastes have always been a part of the Site`s operation. This paper summarizes the current status and future plans with respect to waste reduction to waste reduction and reviews some specific examples of successful activities.

  7. Catalysis over activated high silica zeolites

    SciTech Connect (OSTI)

    Chang, C. D.; Miale, N.

    1985-07-23T23:59:59.000Z

    A process is provided for conducting organic compound conversion over a catalyst composition comprising a crystalline zeolite having a high initial silica-to-alumina mole ratio, said zeolite being prepared by calcining the zeolite, contacting said calcined zeolite with solid aluminum fluoride, and coverting said aluminum fluoride contacted material to hydrogen form.

  8. Removing Phosphate from Hanford High-Phosphate Tank Wastes: FY 2010 Results

    SciTech Connect (OSTI)

    Lumetta, Gregg J.; Braley, Jenifer C.; Edwards, Matthew K.; Qafoku, Odeta; Felmy, Andrew R.; Carter, Jennifer C.; MacFarlan, Paul J.

    2010-09-22T23:59:59.000Z

    The U.S. Department of Energy (DOE) is responsible for environmental remediation at the Hanford Site in Washington State, a former nuclear weapons production site. Retrieving, processing, immobilizing, and disposing of the 2.2 × 105 m3 of radioactive wastes stored in the Hanford underground storage tanks dominates the overall environmental remediation effort at Hanford. The cornerstone of the tank waste remediation effort is the Hanford Tank Waste Treatment and Immobilization Plant (WTP). As currently designed, the capability of the WTP to treat and immobilize the Hanford tank wastes in the expected lifetime of the plant is questionable. For this reason, DOE has been pursuing supplemental treatment options for selected wastes. If implemented, these supplemental treatments will route certain waste components to processing and disposition pathways outside of WTP and thus will accelerate the overall Hanford tank waste remediation mission.

  9. The effect of high-level waste glass composition on spinel liquidus temperature

    SciTech Connect (OSTI)

    Hrma, Pavel R.; Riley, Brian J.; Crum, Jarrod V.; Matyas, Josef

    2014-01-15T23:59:59.000Z

    Spinel crystals precipitate in high-level waste glasses containing Fe, Cr, Ni , Mn, Zn, and Ru. The liquidus temperature (TL) of spinel as the primary crystallization phase is a function of glass composition and the spinel solubility (c0) is a function of both glass composition and temperature (T). Previously reported models of TL as a function of composition are based on TL measured directly, which requires laborious experimental procedures. Viewing the curve of c0 versus T as the liquidus line allows a significant broadening of the composition region for model fitting. This paper estimates TL as a function of composition based on c0 data obtained with the X-ray diffraction technique.

  10. Crucible Study of Spinel Settling in Molten High-Level Waste Glass

    SciTech Connect (OSTI)

    Klouzek, Jaroslav; Alton, Jesse; Hrma, Pavel R.; Plaisted, Trevor J.

    2000-04-12T23:59:59.000Z

    To produce the conditions for settling of spinel crystals in a quiescent high-level waste glass melt, we used a double crucible assembly that eliminated Marangoni convection and limited bubble generation in a portion of melt volume. We observed the movement of the settling front as a function of time at temperatures 900, 950 and 1000?C. The shape of the settling front was approximately parabolic with a flap tip indicating that the settling crystals drove a convective cell within the melt. The rate of settling was close to that predicted by the Stokes' law when the growth rate of spinel crystals was taken into account. The calculated settling velocity was modified by a semi-empirical settling function providing an agreement with experimental results within 5%. In addition, spinel settling was simulated by the mathematical model that predicted the concentration distribution of spinel in glass melt and the accumulation of particles at the bottom of the crucible.

  11. The effect of high-level waste glass composition on spinel liquidus temperature

    SciTech Connect (OSTI)

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Riley, Brian J. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hrma, Pavel [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

    2012-11-15T23:59:59.000Z

    Spinel crystals precipitate in high-level waste glasses containing Fe, Cr, Ni, Mn, Zn, and Ru. The liquidus temperature (T{sub L}d) of spinel as the primary crystallization phase is a function of glass composition, and the spinel solubility (c{sub o}) is a function of both glass composition and temperature (T). Previously reported models of T{sub L} as a function of composition are based on T{sub L} measured directly, which requires laborious experimental procedures. Viewing the curve of c{sub o} versus T as the liquidus line allows a significant broadening of the composition region for model fitting. This paper estimates T{sub L} as a function of composition based on c{sub o} data obtained with the X-ray diffraction technique.

  12. Phase chemistry and radionuclide retention of high level radioactive waste tank sludges

    SciTech Connect (OSTI)

    KRUMHANSL,JAMES L.; BRADY,PATRICK V.; ZHANG,PENGCHU; ARTHUR,SARA E.; HUTCHERSON,SHEILA K.; LIU,J.; QIAN,M.; ANDERSON,HOWARD L.

    2000-05-19T23:59:59.000Z

    The US Department of Energy (DOE) has millions of gallons of high level nuclear waste stored in underground tanks at Hanford, Washington and Savannah River, South Carolina. These tanks will eventually be emptied and decommissioned. This will leave a residue of sludge adhering to the interior tank surfaces that may contaminate groundwaters with radionuclides and RCRA metals. Experimentation on such sludges is both dangerous and prohibitively expensive so there is a great advantage to developing artificial sludges. The US DOE Environmental Management Science Program (EMSP) has funded a program to investigate the feasibility of developing such materials. The following text reports on the success of this program, and suggests that much of the radioisotope inventory left in a tank will not move out into the surrounding environment. Ultimately, such studies may play a significant role in developing safe and cost effective tank closure strategies.

  13. C-106 High-Level Waste Solids: Washing/Leaching and Solubility Versus Temperature Studies

    SciTech Connect (OSTI)

    GJ Lumetta; DJ Bates; PK Berry; JP Bramson; LP Darnell; OT Farmer III; LR Greenwood; FV Hoopes; RC Lettau; GF Piepel; CZ Soderquist; MJ Steele; RT Steele; MW Urie; JJ Wagner

    2000-01-26T23:59:59.000Z

    This report describes the results of a test conducted by Battelle to assess the effects of inhibited water washing and caustic leaching on the composition of the Hanford tank C-106 high-level waste (HLW) solids. The objective of this work was to determine the composition of the C-106 solids remaining after washing with 0.01M NaOH or leaching with 3M NaOH. Another objective of this test was to determine the solubility of various C-106 components as a function of temperature. The work was conducted according to test plan BNFL-TP-29953-8,Rev. 0, Determination of the Solubility of HLW Sludge Solids. The test went according to plan, with only minor deviations from the test plan. The deviations from the test plan are discussed in the experimental section.

  14. Non-Invasive Diagnostics for Measuring Physical Properties and Processes in High Level Wastes

    SciTech Connect (OSTI)

    Robert Powell; David Pfund

    2005-07-17T23:59:59.000Z

    This research demonstrated the usefulness of tomographic techniques for determining the physical properties of slurry suspensions. Of particular interest was the measurement of the viscosity of suspensions in complex liquids and modeling these. We undertook a long rage program that used two techniques, magnetic resonance imaging and ultrasonic pulsed Doppler velocimetry. Our laboratory originally developed both of these for the measurement of viscosity of complex liquids and suspensions. We have shown that the relationship between shear viscosity and shear rate can be determined over a wide range of shear rates from a single measurement. We have also demonstrated these techniques for many non-Newtonian fluids which demonstrate highly shear thinning behavior. This technique was extended to determine the yield stress with systems of interacting particles. To model complex slurries that may be found in wastes applications, we have also used complex slurries that are found in industrial applications

  15. Radioactive Bench-scale Steam Reformer Demonstration of a Monolithic Steam Reformed Mineralized Waste Form for Hanford Waste Treatment Plant Secondary Waste - 12306

    SciTech Connect (OSTI)

    Evans, Brent; Olson, Arlin; Mason, J. Bradley; Ryan, Kevin [THOR Treatment Technologies, LLC - 106 Newberry St. SW, Aiken, SC 29801 (United States); Jantzen, Carol; Crawford, Charles [Savannah River Nuclear Solutions (SRNL), LLC, Aiken, SC 29808 (United States)

    2012-07-01T23:59:59.000Z

    Hanford currently has 212,000 m{sup 3} (56 million gallons) of highly radioactive mixed waste stored in the Hanford tank farm. This waste will be processed to produce both high-level and low-level activity fractions, both of which are to be vitrified. Supplemental treatment options have been under evaluation for treating portions of the low-activity waste, as well as the liquid secondary waste from the low-activity waste vitrification process. One technology under consideration has been the THOR{sup R} fluidized bed steam reforming process offered by THOR Treatment Technologies, LLC (TTT). As a follow-on effort to TTT's 2008 pilot plant FBSR non-radioactive demonstration for treating low-activity waste and waste treatment plant secondary waste, TTT, in conjunction with Savannah River National Laboratory, has completed a bench scale evaluation of this same technology on a chemically adjusted radioactive surrogate of Hanford's waste treatment plant secondary waste stream. This test generated a granular product that was subsequently formed into monoliths, using a geo-polymer as the binding agent, that were subjected to compressibility testing, the Product Consistency Test and other leachability tests, and chemical composition analyses. This testing has demonstrated that the mineralized waste form, produced by co-processing waste with kaolin clay using the TTT process, is as durable as low-activity waste glass. Testing has shown the resulting monolith waste form is durable, leach resistant, and chemically stable, and has the added benefit of capturing and retaining the majority of Tc-99, I-129, and other target species at high levels. (authors)

  16. EM-21 HIGHER WASTE LOADING GLASSES FOR ENHANCED DOE HIGH-LEVEL WASTE MELTER THROUGHPUT STUDIES - 10194

    SciTech Connect (OSTI)

    Raszewski, F.; Peeler, D.; Edwards, T.

    2009-11-18T23:59:59.000Z

    Supplemental validation data has been generated that will be used to determine the applicability of the current Defense Waste Processing Facility (DWPF) liquidus temperature (T{sub L}) model to expanded DWPF glass regions of interest based on higher waste loadings. For those study glasses which had very close compositional overlap with the model development and/or model validation ranges (except TiO{sub 2} and MgO concentrations), there was very little difference in the predicted and measured TL values, even though the TiO{sub 2} contents were above the 2 wt% upper limit. The results indicate that the current T{sub L} model is applicable in these compositional regions. As the compositional overlap between the model validation ranges diverged from the target glass compositions, the T{sub L} data suggest that the model under-predicted the measured values. These discrepancies imply that there are individual oxides or their combinations that were outside of the model development and/or validation range over which the model was previously assessed. These oxides include B{sub 2}O{sub 3}, SiO{sub 2}, MnO, TiO{sub 2} and/or their combinations. More data is required to fill in these anticipated DWPF compositional regions so that the model coefficients could be refit to account for these differences.

  17. PROGRAMMATIC ASSESSMENT OF RADIOACTIVE WASTE MANAGEMENT NUCLEAR FUEL AND WASTE PROGRAMS. Operational Planning and Development (Activity No. AR OS 10 05 K; ONL-WN06)

    SciTech Connect (OSTI)

    None

    1980-06-30T23:59:59.000Z

    Gilbert/Commonwealth (G/C) has performed an assessment of the waste management operations at Oak Ridge National Laboratory (ORNL). The objective of this study was to review radioactive waste management as practiced at ORNL and to recommend improvements or alternatives for further study. The study involved: 1) an on-site survey of ORNL radioactive waste management operations; 2) a review of radioactive waste source data, records, and regulatory requirements; 3) an assessment of existing and planned treatment, storage, and control facilities; and 4) identification of alternatives for improving waste management operations. Information for this study was obtained from both personal interviews and written reports. The G/C suggestions for improving ORNL waste management operations are summarized. Regulatory requirements governing ORNL waste management operations are discussed. Descriptions and discussions of the radioactive liquid, solid, and gaseous waste systems are presented. The waste operations control complex is discussed.

  18. Waste Form Development for the Solidification of PDCF/MOX Liquid Waste Streams

    SciTech Connect (OSTI)

    COZZI, ALEX

    2004-02-18T23:59:59.000Z

    At the Savannah River Site, part of the Department of Energy's nuclear materials complex located in South Carolina, cementation has been selected as the solidification method for high-alpha and low-activity waste streams generated in the planned plutonium disposition facilities. A Waste Solidification Building (WSB) that will be used to treat and solidify three radioactive liquid waste streams generated by the Pit Disassembly and Conversion Facility) and the Mixed Oxide Fuel Fabrication Facility is in the preliminary design stage. The WSB is expected to treat a transuranic (TRU) waste stream composed primarily of americium and two low-level waste (LLW) streams. The acidic wastes will be concentrated in the WSB evaporator and neutralized in a cement head tank prior to solidification. A series of TRU mixes were prepared to produce waste forms exhibiting a range of processing and cured properties. The LLW mixes were prepared using the premix from the preferred TRU waste form. All of the waste forms tested passed the Toxicity Characteristic Leaching Procedure. After processing in the WSB, current plans are to dispose of the solidified TRU waste at the Waste Isolation Pilot Plant in New Mexico and the solidified LLW waste at an approved low-level waste disposal facility.

  19. Remote Handled TRU Waste Status and Activities and Challenges at the Hanford Site

    SciTech Connect (OSTI)

    MCKENNEY, D.E.

    2000-02-01T23:59:59.000Z

    A significant portion of the Department of Energy's forecast volume of remote-handled (RH) transuranic (TRU) waste will originate from the Hanford Site. The forecasted Hanford RH-TRU waste volume of over 2000 cubic meters may constitute over one-third of the forecast inventory of RH-TRU destined for disposal at the Waste Isolation Pilot Plant (WIPP). To date, the Hanford TRU waste program has focused on the retrieval, treatment and certification of the contact-handled transuranic (CH-TRU) wastes. This near-term focus on CH-TRU is consistent with the National TRU Program plans and capabilities. The first shipment of CH-TRU waste from Hanford to the WIPP is scheduled early in Calendar Year 2000. Shipments of RH-TRU from Hanford to the WIPP are scheduled to begin in Fiscal Year 2006 per the National TRU Waste Management Plan. This schedule has been incorporated into milestones within the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement). These Tri-Party milestones (designated the ''M-91'' series of milestones) relate to development of project management plans, completion of design efforts, construction and contracting schedules, and initiation of process operations. The milestone allows for modification of an existing facility, construction of a new facility, and/or commercial contracting to provide the capabilities for processing and certification of RH-TRU wastes for disposal at the WIPP. The development of a Project Management Plan (PMP) for TRU waste is the first significant step in the development of a program for disposal of Hanford's RH-TRU waste. This PMP will address the path forward for disposition of waste streams that cannot be prepared for disposal in the Hanford Waste Receiving and Processing facility (a contact-handled, small container facility) or other Site facilities. The PMP development effort has been initiated, and the PMP will be provided to the regulators for their approval by June 30, 2000. This plan will detail the path forward for the Hanford RH-TRU program.

  20. Use of depleted uranium metal as cask shielding in high-level waste storage, transport, and disposal systems

    SciTech Connect (OSTI)

    Yoshimura, H.R.; Ludwigsen, J.S.; McAllaster, M.E. [and others

    1996-09-01T23:59:59.000Z

    The US DOE has amassed over 555,000 metric tons of depleted uranium from its uranium enrichment operations. Rather than dispose of this depleted uranium as waste, this study explores a beneficial use of depleted uranium as metal shielding in casks designed to contain canisters of vitrified high-level waste. Two high-level waste storage, transport, and disposal shielded cask systems are analyzed. The first system employs a shielded storage and disposal cask having a separate reusable transportation overpack. The second system employs a shielded combined storage, transport, and disposal cask. Conceptual cask designs that hold 1, 3, 4 and 7 high-level waste canisters are described for both systems. In all cases, cask design feasibility was established and analyses indicate that these casks meet applicable thermal, structural, shielding, and contact-handled requirements. Depleted uranium metal casting, fabrication, environmental, and radiation compatibility considerations are discussed and found to pose no serious implementation problems. About one-fourth of the depleted uranium inventory would be used to produce the casks required to store and dispose of the nearly 15,400 high-level waste canisters that would be produced. This study estimates the total-system cost for the preferred 7-canister storage and disposal configuration having a separate transportation overpack would be $6.3 billion. When credits are taken for depleted uranium disposal cost, a cost that would be avoided if depleted uranium were used as cask shielding material rather than disposed of as waste, total system net costs are between $3.8 billion and $5.5 billion.

  1. A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests

    SciTech Connect (OSTI)

    Thien, Mike G. [Washington River Protection Solutions, LLC, Richland, WA (United States); Barnes, Steve M. [URS, Richland, WA (United States)

    2013-01-17T23:59:59.000Z

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described.

  2. A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests - 13342

    SciTech Connect (OSTI)

    Thien, Mike G. [Washington River Protection Solutions, LLC, P.O Box 850, Richland WA, 99352 (United States)] [Washington River Protection Solutions, LLC, P.O Box 850, Richland WA, 99352 (United States); Barnes, Steve M. [Waste Treatment Plant, 2435 Stevens Center Place, Richland WA 99354 (United States)] [Waste Treatment Plant, 2435 Stevens Center Place, Richland WA 99354 (United States)

    2013-07-01T23:59:59.000Z

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described. (authors)

  3. Active control system for high speed windmills

    DOE Patents [OSTI]

    Avery, D.E.

    1988-01-12T23:59:59.000Z

    A pump stroke is matched to the operating speed of a high speed windmill. The windmill drives a hydraulic pump for a control. Changes in speed of a wind driven shaft open supply and exhaust valves to opposite ends of a hydraulic actuator to lengthen and shorten an oscillating arm thereby lengthening and shortening the stroke of an output pump. Diminishing wind to a stall speed causes the valves to operate the hydraulic cylinder to shorten the oscillating arm to zero. A pressure accumulator in the hydraulic system provides the force necessary to supply the hydraulic fluid under pressure to drive the actuator into and out of the zero position in response to the windmill shaft speed approaching and exceeding windmill stall speed. 4 figs.

  4. Active control system for high speed windmills

    DOE Patents [OSTI]

    Avery, Don E. (45-437 Akimala St., Honolulu, HI 96744)

    1988-01-01T23:59:59.000Z

    A pump stroke is matched to the operating speed of a high speed windmill. The windmill drives a hydraulic pump for a control. Changes in speed of a wind driven shaft open supply and exhaust valves to opposite ends of a hydraulic actuator to lengthen and shorten an oscillating arm thereby lengthening and shortening the stroke of an output pump. Diminishing wind to a stall speed causes the valves to operate the hydraulic cylinder to shorten the oscillating arm to zero. A pressure accumulator in the hydraulic system provides the force necessary to supply the hydraulic fluid under pressure to drive the actuator into and out of the zero position in response to the windmill shaft speed approaching and exceeding windmill stall speed.

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

    SciTech Connect (OSTI)

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

    2008-09-23T23:59:59.000Z

    The objective of this task was to develop glass formulations for (Department of Energy) DOE waste streams with high aluminum concentrations to avoid nepheline formation while maintaining or meeting waste loading and/or waste throughput expectations as well as satisfying critical process and product performance related constraints. Liquidus temperatures and crystallization behavior were carefully characterized to support model development for higher waste loading glasses. The experimental work, characterization, and data interpretation necessary to meet these objectives were performed among three partnering laboratories: the V.G. Khlopin Radium Institute (KRI), Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL). Projected glass compositional regions that bound anticipated Defense Waste Processing Facility (DWPF) and Hanford high level waste (HLW) glass regions of interest were developed and used to generate glass compositions of interest for meeting the objectives of this study. A thorough statistical analysis was employed to allow for a wide range of waste glass compositions to be examined while minimizing the number of glasses that had to be fabricated and characterized in the laboratory. The glass compositions were divided into two sets, with 45 in the test matrix investigated by the U.S. laboratories and 30 in the test matrix investigated by KRI. Fabrication and characterization of the US and KRI-series glasses were generally handled separately. This report focuses mainly on the US-series glasses. Glasses were fabricated and characterized by SRNL and PNNL. Crystalline phases were identified by X-ray diffraction (XRD) in the quenched and canister centerline cooled (CCC) glasses and were generally iron oxides and spinels, which are not expected to impact durability of the glass. Nepheline was detected in five of the glasses after the CCC heat treatment. Chemical composition measurements for each of the glasses were conducted following an analytical plan. A review of the individual oxides for each glass revealed that there were no errors in batching significant enough to impact the outcome of the study. A comparison of the measured compositions of the replicates indicated an acceptable degree of repeatability as the percent differences for most of the oxides were less than 5% and percent differences for all of the oxides were less than 10 wt%. Chemical durability was measured using the Product Consistency Test (PCT). All but two of the study glasses had normalized leachate for boron (NL [B]) values that were well below that of the Environmental Assessment (EA) reference glass. The two highest NL [B] values were for the CCC versions of glasses US-18 and US-27 (10.498 g/L and 15.962 g/L, respectively). Nepheline crystallization was identified by qualitative XRD in five of the US-series glasses. Each of these five glasses (US-18, US-26, US-27, US-37 and US-43) showed a significant increase in NL [B] values after the CCC heat treatment. This reduction in durability can be attributed to the formation of nepheline during the slow cooling cycle and the removal of glass formers from the residual glass network. The liquidus temperature (T{sub L}) of each glass in the study was determined by both optical microscopy and XRD methods. The correlation coefficient of the measured XRD TL data versus the measured optical TL data was very good (R{sup 2} = 0.9469). Aside from a few outliers, the two datasets aligned very well across the entire temperature range (829 C to 1312 C for optical data and 813 C to 1310 C for XRD crystal fraction data). The data also correlated well with the predictions of a PNNL T{sub L} model. The correlation between the measured and calculated data had a higher degree of merit for the XRD crystal fraction data than for the optical data (higher R{sup 2} value of 0.9089 versus 0.8970 for the optical data). The SEM-EDS analysis of select samples revealed the presence of undissolved RuO{sub 2} in all glasses due to the low solubility of RuO{sub 2} in borosilicate glass. These

  6. Technology of high-level nuclear waste disposal. Advances in the science and engineering of the management of high-level nuclear wastes. Volume 2

    SciTech Connect (OSTI)

    Hofmann, P.L. (ed.)

    1982-01-01T23:59:59.000Z

    The twenty papers in this volume are divided into three parts: site exploration and characterization; repository development and design; and waste package development and design. These papers represent the status of technology that existed in 1981 and 1982. Individual papers were processed for inclusion in the Energy Data Base.

  7. Savannah River Site High-Level Waste Tank Closure Final Environmental Impact Statement

    SciTech Connect (OSTI)

    N /A

    2002-05-31T23:59:59.000Z

    The U.S. Atomic Energy Commission, a U.S. Department of Energy (DOE) predecessor agency, established the Savannah River Site (SRS) near Aiken, South Carolina, in the early 1950s. The primary mission of SRS was to produce nuclear materials for national defense. With the end of the Cold War and the reduction in the size of the United States stockpile of nuclear weapons, the SRS mission has changed. While national defense is still an important facet of the mission, SRS no longer produces nuclear materials and the mission is focused on material stabilization, environmental restoration, waste management, and decontamination and decommissioning of facilities that are no longer needed. As a result of its nuclear materials production mission, SRS generated large quantities of high-level radioactive waste (HLW). The HLW resulted from dissolving spent reactor fuel and nuclear targets to recover the valuable radioactive isotopes. DOE had stored the HLW in 51 large underground storage tanks located in the F- and H-Area Tank Farms at SRS. DOE has emptied and closed two of those tanks. DOE is treating the HLW, using a process called vitrification. The highly radioactive portion of the waste is mixed with a glass like material and stored in stainless steel canisters at SRS, pending shipment to a geologic repository for disposal. This process is currently underway at SRS in the Defense Waste Processing Facility (DWPF). The HLW tanks at SRS are of four different types, which provide varying degrees of protection to the environment due to different degrees of containment. The tanks are operated under the authority of the Atomic Energy Act of 1954 (AEA) and DOE Orders issued under the AEA. The tanks are permitted by the South Carolina Department of Health and Environmental Control (SCDHEC) under South Carolina wastewater regulations, which require permitted facilities to be closed after they are removed from service. DOE has entered into an agreement with the U.S. Environmental Protection Agency (EPA) and SCDHEC to close the HLW tanks after they have been removed from service. Closure of the HLW tanks would comply with DOE's responsibilities under the AEA and the South Carolina closure requirements and be carried out under a schedule agreed to by DOE, EPA, and SCDHEC. There are several ways to close the HLW tanks. DOE has prepared this Environmental Impact Statement (EIS) to ensure that the public and DOE's decision makers have a thorough understanding of the potential environmental impacts of alternative means of closing the tanks. This Summary: (1) describes the HLW tanks and the closure process, (2) describes the National Environmental Policy Act (NEPA) process that DOE is using to aid in decision making, (3) summarizes the alternatives for closing the HLW tanks and identifies DOE.s preferred alternative, and (4) identifies the major conclusions regarding environmental impacts, areas of controversy, and issues that remain to be resolved as DOE proceeds with the HLW tank closure process.

  8. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 2

    SciTech Connect (OSTI)

    Not Available

    1994-04-01T23:59:59.000Z

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Document (S/RID) is contained in multiple volumes. This document (Volume 2) presents the standards and requirements for the following sections: Quality Assurance, Training and Qualification, Emergency Planning and Preparedness, and Construction.

  9. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID)

    SciTech Connect (OSTI)

    Not Available

    1994-04-01T23:59:59.000Z

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 3) presents the standards and requirements for the following sections: Safeguards and Security, Engineering Design, and Maintenance.

  10. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 5

    SciTech Connect (OSTI)

    Not Available

    1994-04-01T23:59:59.000Z

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 5) outlines the standards and requirements for the Fire Protection and Packaging and Transportation sections.

  11. Immobilized High Level Waste (HLW) Interim Storage Alternative Generation and analysis and Decision Report 2nd Generation Implementing Architecture

    SciTech Connect (OSTI)

    CALMUS, R.B.

    2000-09-14T23:59:59.000Z

    Two alternative approaches were previously identified to provide second-generation interim storage of Immobilized High-Level Waste (IHLW). One approach was retrofit modification of the Fuel and Materials Examination Facility (FMEF) to accommodate IHLW. The results of the evaluation of the FMEF as the second-generation IHLW interim storage facility and subsequent decision process are provided in this document.

  12. High Performance Zero-Bleed CLSM/Grout Mixes for High-Level Waste Tank Closures Strategic Research and Development - FY99 Report

    SciTech Connect (OSTI)

    Langton, C.A.

    2000-08-11T23:59:59.000Z

    The overall objective of this program, SRD-99-08, was to design and test suitable materials, which can be used to close high-level waste tanks at SRS. Fill materials can be designed to perform several functions including chemical stabilization and/or physical encapsulation of incidental waste so that the potential for transport of contaminants into the environment is reduced. Also they are needed to physically stabilize the void volume in the tanks to prevent/minimize future subsidence and inadvertent intrusion. The intent of this work was to develop a zero-bleed soil CLSM (ZBS-CLSM) and a zero-bleed concrete mix (ZBC) which meet the unique placement and stabilization/encapsulation requirements for high-level waste tank closures. These mixes in addition to the zero-bleed CLSM mixes formulated for closure of Tanks 17-F and 20-F provide design engineers with a suite of options for specifying materials for future tank closures.

  13. Adsorption of Ruthenium, Rhodium and Palladium from Simulated High-Level Liquid Waste by Highly Functional Xerogel - 13286

    SciTech Connect (OSTI)

    Onishi, Takashi [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan)] [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan); Koyama, Shin-ichi [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan)] [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan); Mimura, Hitoshi [Dept. of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University Aramaki-Aza-Aoba 6-6-01-2,Aoba-ku, Sendai-shi, Miyagi-ken, 980-8579 (Japan)] [Dept. of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University Aramaki-Aza-Aoba 6-6-01-2,Aoba-ku, Sendai-shi, Miyagi-ken, 980-8579 (Japan)

    2013-07-01T23:59:59.000Z

    Fission products are generated by fission reactions in nuclear fuel. Platinum group (Pt-G) elements, such as palladium (Pd), rhodium (Rh) and ruthenium (Ru), are also produced. Generally, Pt-G elements play important roles in chemical and electrical industries. Highly functional xerogels have been developed for recovery of these useful Pt-G elements from high - level radioactive liquid waste (HLLW). An adsorption experiment from simulated HLLW was done by the column method to study the selective adsorption of Pt-G elements, and it was found that not only Pd, Rh and Ru, but also nickel, zirconium and tellurium were adsorbed. All other elements were not adsorbed. Adsorbed Pd was recovered by washing the xerogel-packed column with thiourea solution and thiourea - nitric acid mixed solution in an elution experiment. Thiourea can be a poison for automotive exhaust emission system catalysts, so it is necessary to consider its removal. Thermal decomposition and an acid digestion treatment were conducted to remove sulfur in the recovered Pd fraction. The relative content of sulfur to Pd was decreased from 858 to 0.02 after the treatment. These results will contribute to design of the Pt-G element separation system. (authors)

  14. The French national program for spent fuel and high-level waste management

    SciTech Connect (OSTI)

    Giraud, J.P.; Demontalembert, J.A. [COGEMA, Velizy-Villacoublay (France)

    1993-12-31T23:59:59.000Z

    From its very beginning, the French national program for spent fuel and HLW management is aimed at the recycling of energetic materials and the safe disposal of nuclear waste. Spent fuel reprocessing is the cornerstone of this program, since it directly opens the way to energetic material recycling, waste minimization and safe conditioning. It is complemented by the HLW management program which is defined by the HLW disposal regulation and the Waste Act issued in 1991.

  15. Low-temperature lithium diffusion in simulated high-level boroaluminosilicate nuclear waste glasses

    SciTech Connect (OSTI)

    Neeway, James J.; Kerisit, Sebastien N.; Gin, Stephane; Wang, Zhaoying; Zhu, Zihua; Ryan, Joseph V.

    2014-12-01T23:59:59.000Z

    Ion exchange is recognized as an integral, if underrepresented, mechanism influencing glass corrosion. However, due to the formation of various alteration layers in the presence of water, it is difficult to conclusively deconvolute the mechanisms of ion exchange from other processes occurring simultaneously during corrosion. In this work, an operationally inert non-aqueous solution was used as an alkali source material to isolate ion exchange and study the solid-state diffusion of lithium. Specifically, the experiments involved contacting glass coupons relevant to the immobilization of high-level nuclear waste, SON68 and CJ-6, which contained Li in natural isotope abundance, with a non-aqueous solution of 6LiCl dissolved in dimethyl sulfoxide at 90 °C for various time periods. The depth profiles of major elements in the glass coupons were measured using time-of-flight secondary ion mass spectrometry (ToF-SIMS). Lithium interdiffusion coefficients, DLi, were then calculated based on the measured depth profiles. The results indicate that the penetration of 6Li is rapid in both glasses with the simplified CJ-6 glass (D6Li ? 4.0-8.0 × 10-21 m2/s) exhibiting faster exchange than the more complex SON68 glass (DLi ? 2.0-4.0 × 10-21 m2/s). Additionally, sodium ions present in the glass were observed to participate in ion exchange reactions; however, different diffusion coefficients were necessary to fit the diffusion profiles of the two alkali ions. Implications of the diffusion coefficients obtained in the absence of alteration layers to the long-term performance of nuclear waste glasses in a geological repository system are also discussed.

  16. Expected environments in high-level nuclear waste and spent fuel repositories in salt

    SciTech Connect (OSTI)

    Claiborne, H.C.; Rickertsen, L.D., Graham, R.F.

    1980-08-01T23:59:59.000Z

    The purpose of this report is to describe the expected environments associated with high-level waste (HLW) and spent fuel (SF) repositories in salt formations. These environments include the thermal, fluid, pressure, brine chemistry, and radiation fields predicted for the repository conceptual designs. In this study, it is assumed that the repository will be a room and pillar mine in a rock-salt formation, with the disposal horizon located approx. 2000 ft (610 m) below the surface of the earth. Canistered waste packages containing HLW in a solid matrix or SF elements are emplaced in vertical holes in the floor of the rooms. The emplacement holes are backfilled with crushed salt or other material and sealed at some later time. Sensitivity studies are presented to show the effect of changing the areal heat load, the canister heat load, the barrier material and thickness, ventilation of the storage room, and adding a second row to the emplacement configuration. The calculated thermal environment is used as input for brine migration calculations. The vapor and gas pressure will gradually attain the lithostatic pressure in a sealed repository. In the unlikely event that an emplacement hole will become sealed in relatively early years, the vapor space pressure was calculated for three scenarios (i.e., no hole closure - no backfill, no hole closure - backfill, and hole closure - no backfill). It was assumed that the gas in the system consisted of air and water vapor in equilibrium with brine. A computer code (REPRESS) was developed assuming that these changes occur slowly (equilibrium conditions). The brine chemical environment is outlined in terms of brine chemistry, corrosion, and compositions. The nuclear radiation environment emphasized in this report is the stored energy that can be released as a result of radiation damage or crystal dislocations within crystal lattices.

  17. alkaline high-level waste: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the atmosphere Geosciences Websites Summary: -solid waste for CO2 mitigation and reduction of greenhouse effect gases into the atmosphere. ? 2008 ElsevierCarbonation of...

  18. Thirty-year solid waste generation forecast for facilities at SRS

    SciTech Connect (OSTI)

    Not Available

    1994-07-01T23:59:59.000Z

    The information supplied by this 30-year solid waste forecast has been compiled as a source document to the Waste Management Environmental Impact Statement (WMEIS). The WMEIS will help to select a sitewide strategic approach to managing present and future Savannah River Site (SRS) waste generated from ongoing operations, environmental restoration (ER) activities, transition from nuclear production to other missions, and decontamination and decommissioning (D&D) programs. The EIS will support project-level decisions on the operation of specific treatment, storage, and disposal facilities within the near term (10 years or less). In addition, the EIS will provide a baseline for analysis of future waste management activities and a basis for the evaluation of the specific waste management alternatives. This 30-year solid waste forecast will be used as the initial basis for the EIS decision-making process. The Site generates and manages many types and categories of waste. With a few exceptions, waste types are divided into two broad groups-high-level waste and solid waste. High-level waste consists primarily of liquid radioactive waste, which is addressed in a separate forecast and is not discussed further in this document. The waste types discussed in this solid waste forecast are sanitary waste, hazardous waste, low-level mixed waste, low-level radioactive waste, and transuranic waste. As activities at SRS change from primarily production to primarily decontamination and decommissioning and environmental restoration, the volume of each waste s being managed will change significantly. This report acknowledges the changes in Site Missions when developing the 30-year solid waste forecast.

  19. Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 1, Methodology and results

    SciTech Connect (OSTI)

    Rechard, R.P. [ed.

    1993-12-01T23:59:59.000Z

    This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste. Although numerous caveats must be placed on the results, the general findings were as follows: Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

  20. Minimizing High Spatial Frequency Residual in Active Space Telescope Mirrors

    E-Print Network [OSTI]

    . Miller June 2008 SSL # 4-08 #12;#12;Minimizing High Spatial Frequency Residual in Active Space Telescope Mirrors Thomas Gray, David W. Miller June 2008 SSL # 4-08 This work is based on the unaltered text

  1. High Level Overview of DOE Biomass Logistics II Project Activities

    Broader source: Energy.gov [DOE]

    Breakout Session 1B—Integration of Supply Chains I: Breaking Down Barriers High Level Overview of DOE Biomass Logistics II Project Activities Kevin Comer, Associate Principal, Antares Group Inc.

  2. Nuclear Waste Assessment System for Technical Evaluation (NUWASTE)

    E-Print Network [OSTI]

    .S. Department of Energy (DOE) activities related to the management of spent nuclear fuel (SNF) and high-cycle initiatives on the generation of spent nuclear fuel and high-level radioactive waste. The effort is led by Dr of the U.S. Nuclear Waste Technical Review Board's analysis of the impact of alternative fuel

  3. A data base and a standard material for use in acceptance testing of low-activity waste products

    SciTech Connect (OSTI)

    Wolf, S.F.; Ebert, W.L.; Luo, J.S.; Strachan, D.M.

    1998-04-01T23:59:59.000Z

    The authors have conducted replicate dissolution tests following the product consistency test (PCT) procedure to measure the mean and standard deviation of the solution concentrations of B, Na, and Si at various combinations of temperature, duration, and glass/water mass ratio. Tests were conducted with a glass formulated to be compositionally similar to low-activity waste products anticipated for Hanford to evaluate the adequacy of test methods that have been designated in privatization contracts for use in product acceptance. An important finding from this set of tests is that the solution concentrations generated in tests at 20 C will likely be too low to measure the dissolution rates of waste products reliably. Based on these results, the authors recommend that the acceptance test be conducted at 40 C. Tests at 40 C generated higher solution concentrations, were more easily conducted, and the measured rates were easily related to those at 20 C. Replicate measurements of other glass properties were made to evaluate the possible use of LRM-1 as a standard material. These include its composition, homogeneity, density, compressive strength, the Na leachability index with the ANSI/ANS 16.1 leach test, and if the glass is characteristically hazardous with the toxicity characteristic leach procedure. The values of these properties were within the acceptable limits identified for Hanford low-activity waste products. The reproducibility of replicate tests and analyses indicates that the glass would be a suitable standard material.

  4. Advanced oxidation treatment of high strength bilge and aqueous petroleum waste

    SciTech Connect (OSTI)

    Hulsey, R.A.; Kobylinski, E.A. [Black and Veatch, Kansas City, MO (United States); Leach, B. [EEC, Inc., Virginia Beach, VA (United States); Pearce, L. [TRITECH, Greensboro, NC (United States)

    1996-11-01T23:59:59.000Z

    The Craney Island Fuel Depot is the largest US Navy fuel terminal in the continental US. Services provided at this facility include fuel storage (current capacity is 1.5 million barrels), fuel reclamation (recovery of oil from oily wastewater), and physical/chemical treatment for the removal of residual oil from bilge water and from aqueous petroleum waste. Current wastewater treatment consists of storage/equalization, oil/water separation, dissolved air flotation, sand filtration, and carbon adsorption. The Navy initiated this study to comply with the State requirement that its existing physical/chemical oily wastewater treatment plant be upgraded to remove soluble organics and produce an effluent which would meet acute toxicity limits. The pilot tests conducted during the study included several variations of chemical and biological wastewater treatment processes. While biological treatment alone was capable of meeting the proposed BOD limit of 26 mg/L, the study showed that the effluent of the biological process contained a high concentration of refractory (nonbiodegradable) organics and could not consistently meet the proposed limits for COD and TOC when treating high-strength wastewater. Additional tests were conducted with advanced oxidation processes (AOPs). AOPs were evaluated for use as independent treatment processes as well as polishing processes following biological treatment. The AOP processes used for this study included combinations of ozone (O{sub 3}) ultraviolet radiation (UV), and hydrogen peroxide (H{sub 2}O{sub 2}).

  5. ALPHN: A computer program for calculating ([alpha], n) neutron production in canisters of high-level waste

    SciTech Connect (OSTI)

    Salmon, R.; Hermann, O.W.

    1992-10-01T23:59:59.000Z

    The rate of neutron production from ([alpha], n) reactions in canisters of immobilized high-level waste containing borosilicate glass or glass-ceramic compositions is significant and must be considered when estimating neutron shielding requirements. The personal computer program ALPHA calculates the ([alpha], n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the ([alpha], n) neutron production of each actinide in neutrons per second and the total for the canister. The ([alpha], n) neutron production rates are source terms only; that is, they are production rates within the glass and do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister. In a typical application, these cases might represent the same canister of vitrified high-level waste at eight different decay times. Run time for a typical problem containing 20 chemical species, 24 actinides, and 8 decay times was 35 s on an IBM AT personal computer. Results of an example based on an expected canister composition at the Defense Waste Processing Facility are shown.

  6. ALPHN: A computer program for calculating ({alpha}, n) neutron production in canisters of high-level waste

    SciTech Connect (OSTI)

    Salmon, R.; Hermann, O.W.

    1992-10-01T23:59:59.000Z

    The rate of neutron production from ({alpha}, n) reactions in canisters of immobilized high-level waste containing borosilicate glass or glass-ceramic compositions is significant and must be considered when estimating neutron shielding requirements. The personal computer program ALPHA calculates the ({alpha}, n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the ({alpha}, n) neutron production of each actinide in neutrons per second and the total for the canister. The ({alpha}, n) neutron production rates are source terms only; that is, they are production rates within the glass and do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister. In a typical application, these cases might represent the same canister of vitrified high-level waste at eight different decay times. Run time for a typical problem containing 20 chemical species, 24 actinides, and 8 decay times was 35 s on an IBM AT personal computer. Results of an example based on an expected canister composition at the Defense Waste Processing Facility are shown.

  7. EXPERIMENTS ON CAKE DEVELOPMENT IN CROSSFLOW FILTRATION FOR HIGH LEVEL WASTE

    SciTech Connect (OSTI)

    Duignan, M.; Nash, C.

    2011-04-14T23:59:59.000Z

    Crossflow filtration is a key process step in many operating and planned waste treatment facilities to separate undissolved solids from supernate slurries. This separation technology generally has the advantage of self cleaning through the action of wall shear stress, which is created by the flow of waste slurry through the filter tubes. However, the ability of filter wall self cleaning depends on the slurry being filtered. Many of the alkaline radioactive wastes are extremely challenging to filtration, e.g., those containing compounds of aluminum and iron, which have particles whose size and morphology reduces permeability. Low filter flux can be a bottleneck in waste processing facilities such as the Salt Waste Processing Facility at the Savannah River Site and the Waste Treatment Plant at the Hanford Site. Any improvement to the filtration rate would lead directly to increased throughput of the entire process. To date, increased rates are generally realized by either increasing the crossflow filter axial flowrate, which is limited by pump capacity, or by increasing filter surface area, which is limited by space and increases the required pump load. In the interest of accelerating waste treatment processing, DOE has funded studies to better understand filtration with the goal of improving filter fluxes in existing crossflow equipment. The Savannah River National Laboratory (SRNL) was included in those studies, with a focus on startup techniques and filter cake development. This paper discusses those filter studies. SRNL set up both dead-end and crossflow filter tests to better understand filter performance based on filter media structure, flow conditions, and filter cleaning. Using non-radioactive simulated wastes, which were both chemically and physically similar to the actual radioactive wastes, the authors performed several tests to demonstrate increases in filter performance. With the proper use of filter flow conditions filter flow rates can be increased over rates currently realized today. This paper describes the selection of a challenging simulated waste and crossflow filter tests to demonstrate how performance can be improved over current operation.

  8. Highly Conductive and Porous Activated Reduced Graphene Oxide Films for High-Power Supercapacitors

    E-Print Network [OSTI]

    Highly Conductive and Porous Activated Reduced Graphene Oxide Films for High-Power Supercapacitors for a free-standing carbon film reported to date. A two-electrode supercapacitor using these carbon films. KEYWORDS: Graphene, flexible film, chemical activation, supercapacitors Free-standing thin film materials

  9. US - Former Soviet Union environmental restoration and waste management activities, March 1994

    SciTech Connect (OSTI)

    Not Available

    1994-03-01T23:59:59.000Z

    The Peaceful Uses of Atomic Energy Agreement was signed between DOE and the Ministry of Atomic Energy for the Russian Federation and provides a mechanism for cooperation in research, development, and safe utilization of nuclear energy. Under the umbrella of this agreement, DOE and the former Ministry of Atomic Power and Industry signed a Memorandum of Cooperation in the areas of environmental restoration and waste management in September 1990. This document discusses the environmental situation, science and technology process, technical projects (separations, contaminant transport, waste treatment, environmental restoration), scientist exchanges, enhanced data transfer, the US-Russia industry partnership (conference, centers), and future actions.

  10. Thermal and Radiolytic Gas Generation in Hanford High-Level Waste

    SciTech Connect (OSTI)

    Bryan, Samuel A.; Pederson, Larry R.; King, C. M.

    2000-01-31T23:59:59.000Z

    The Hanford Site has 177 underground storage tanks containing radioactive wastes that are complex mixes of radioactive and chemical products. Some of these wastes are known to generate and retain large quantities of flammable gases consisting of hydrogen, nitrous oxide, nitrogen, and ammonia. Because these gases are flammable and have the potential for rapid release, the gas generation rate for each tank must be determined to establish the flammability hazard (Johnson et al. 1997). An understanding of gas generation is important to operation of the waste tanks for several reasons. First, knowledge of the overall rate of generation is needed to verify that any given tank has sufficient ventilation to ensure that flammable gases are maintained at a safe level within the dome space. Understanding the mechanisms for production of the various gases is important so that future waste operations do not create conditions that promote the production of hydrogen, ammonia, and nitrous oxide. Studying the generation of gases also provides important data for the composition of the gas mixture, which in turn is needed to assess the flammability characteristics. Finally, information about generation of gases, including the influence of various chemical constituents, temperature, and dose, would aid in assessing the future behavior of the waste during interim storage, implementation of controls, and final waste treatment. This paper summarizes the current knowledge of gas generation pathways and discusses models used in predicting gas generation rates from actual Hanford radioactive wastes. A comparison is made between measured gas generation rates and rates by the predictive models.

  11. Evaluation of alternative treatments for spent fuel rod consolidation wastes and other miscellaneous commercial transuranic wastes

    SciTech Connect (OSTI)

    Ross, W.A.; Schneider, K.J.; Oma, K.H.; Smith, R.I.; Bunnell, L.R.

    1986-05-01T23:59:59.000Z

    Eight alternative treatments (and four subalternatives) are considered for both existing commercial transuranic wastes and future wastes from spent fuel consolidation. Waste treatment is assumed to occur at a hypothetical central treatment facility (a Monitored Retrieval Storage facility was used as a reference). Disposal in a geologic repository is also assumed. The cost, process characteristics, and waste form characteristics are evaluated for each waste treatment alternative. The evaluation indicates that selection of a high-volume-reduction alternative can save almost $1 billion in life-cycle costs for the management of transuranic and high-activity wastes from 70,000 MTU of spent fuel compared to the reference MRS process. The supercompaction, arc pyrolysis and melting, and maximum volume reduction alternatives are recommended for further consideration; the latter two are recommended for further testing and demonstration.

  12. Radioactive and chemotoxic wastes: Only radioactive wastes?

    SciTech Connect (OSTI)

    Eletti, G.F.; Tocci, M. [ENEA DISP, Rome (Italy)

    1993-12-31T23:59:59.000Z

    Radioactive waste arising from Italian Nuclear Power Plants and Research Centers, classified as 1st and 2nd Category wastes, are managed only as radioactive wastes following the Technical Guide No. 26 issued by the Italian Regulatory Body: ENEA DISP on 1987. A very important Regulatory Regime revision for Italian Nuclear Activities started at the end of 1991. This paper considers the need to develop a new strategy dedicated to mixed waste in line with current international trends.

  13. Kinetic model for quartz and spinel dissolution during melting of high-level-waste glass batch

    SciTech Connect (OSTI)

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

    2013-07-24T23:59:59.000Z

    The dissolution of quartz particles and the growth and dissolution of crystalline phases during the conversion of batch to glass potentially affects both the glass melting process and product quality. Crystals of spinel exiting the cold cap to molten glass below can be troublesome during the vitrification of iron-containing high-level wastes. To estimate the distribution of quartz and spinel fractions within the cold cap, we used kinetic models that relate fractions of these phases to temperature and heating rate. Fitting the model equations to data showed that the heating rate, apart from affecting quartz and spinel behavior directly, also affects them indirectly via concurrent processes, such as the formation and motion of bubbles. Because of these indirect effects, it was necessary to allow one kinetic parameter (the pre-exponential factor) to vary with the heating rate. The resulting kinetic equations are sufficiently simple for the detailed modeling of batch-to-glass conversion as it occurs in glass melters. The estimated fractions and sizes of quartz and spinel particles as they leave the cold cap, determined in this study, will provide the source terms needed for modeling the behavior of these solid particles within the flow of molten glass in the melter.

  14. US Department of Energy Storage of Spent Fuel and High Level Waste

    SciTech Connect (OSTI)

    Sandra M Birk

    2010-10-01T23:59:59.000Z

    ABSTRACT This paper provides an overview of the Department of Energy's (DOE) spent nuclear fuel (SNF) and high level waste (HLW) storage management. Like commercial reactor fuel, DOE's SNF and HLW were destined for the Yucca Mountain repository. In March 2010, the DOE filed a motion with the Nuclear Regulatory Commission (NRC) to withdraw the license application for the repository at Yucca Mountain. A new repository is now decades away. The default for the commercial and DOE research reactor fuel and HLW is on-site storage for the foreseeable future. Though the motion to withdraw the license application and delay opening of a repository signals extended storage, DOE's immediate plans for management of its SNF and HLW remain the same as before Yucca Mountain was designated as the repository, though it has expanded its research and development efforts to ensure safe extended storage. This paper outlines some of the proposed research that DOE is conducting and will use to enhance its storage systems and facilities.

  15. Sedimentation behavior of noble metal particles in simulated high-level waste borosilicate glasses

    SciTech Connect (OSTI)

    Nakajima, M.; Ohyama, K.; Morikawa, Y.; Miyauchi, A.; Yamashita, T. [Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319-1109 (Japan); Komamine, S.; Ochi, E. [Japan Nuclear Fuel Limited, Bussan-Bldg. Bekkan, 1-1-5 Nishi-Shinbashi Minato-ku, Tokyo 105-0003 (Japan)

    2013-07-01T23:59:59.000Z

    Solubility of noble metal elements (NME) in the melted borosilicate glass is much smaller than its normal concentration of the high level liquid waste. Thus most of NME show small particles in the melted glass and tend to sediment in the bottom region of the vitrification melter due to their higher density than that of glass. Experiments of the sedimentation of NME particles in the melted glass were carried out under static condition. Three conditions of initial NME concentration (1.1, 3.0, 6.1 wt % with an equivalent for each oxide) in the simulated glass were set and held at 1100 C. degrees up to 2880 hours. The specimen with 1.1 wt % initial NME concentration indicated zone settling, and the settling rate of the interface is constant: 2.4 mm/h. This sedimentation behavior is the type of rapid settling. Following the rapid settling, the settling rate goes gradually slower; this is the type of compressive settling. The specimens with 3.0 wt % and 6.1 wt % initial NME concentration showed compression settling from the beginning. From the settling curve of the interface, the maximum concentration of NME in sediment was estimated to be around 23- 26 wt %. Growth of NME particles was observed by holding at 1100 C. degrees for up to 2880 hours. The viscosity becomes higher as NME concentration increases and the dependence on shear rate becomes simultaneously stronger. The effect of the particle growth to viscosity appears to be not significant.

  16. Design Improvements and Analysis of Innovative High-Level Waste Pipeline Unplugging Technologies - 12171

    SciTech Connect (OSTI)

    Pribanic, Tomas; Awwad, Amer; Crespo, Jairo; McDaniel, Dwayne; Varona, Jose; Gokaltun, Seckin; Roelant, David [Florida International University, Miami, Florida (United States)

    2012-07-01T23:59:59.000Z

    Transferring high-level waste (HLW) between storage tanks or to treatment facilities is a common practice performed at the Department of Energy (DoE) sites. Changes in the chemical and/or physical properties of the HLW slurry during the transfer process may lead to the formation of blockages inside the pipelines resulting in schedule delays and increased costs. To improve DoE's capabilities in the event of a pipeline plugging incident, FIU has continued to develop two novel unplugging technologies: an asynchronous pulsing system and a peristaltic crawler. The asynchronous pulsing system uses a hydraulic pulse generator to create pressure disturbances at two opposite inlet locations of the pipeline to dislodge blockages by attacking the plug from both sides remotely. The peristaltic crawler is a pneumatic/hydraulic operated crawler that propels itself by a sequence of pressurization/depressurization of cavities (inner tubes). The crawler includes a frontal attachment that has a hydraulically powered unplugging tool. In this paper, details of the asynchronous pulsing system's ability to unplug a pipeline on a small-scale test-bed and results from the experimental testing of the second generation peristaltic crawler are provided. The paper concludes with future improvements for the third generation crawler and a recommended path forward for the asynchronous pulsing testing. (authors)

  17. A COMPLETE HISTORY OF THE HIGH-LEVEL WASTE PLANT AT THE WEST VALLEY DEMONSTRATION PROJECT

    SciTech Connect (OSTI)

    Petkus, Lawrence L.; Paul, James; Valenti, Paul J.; Houston, Helene; May, Joseph

    2003-02-27T23:59:59.000Z

    The West Valley Demonstration Project (WVDP) vitrification melter was shut down in September 2002 after being used to vitrify High Level Waste (HLW) and process system residuals for six years. Processing of the HLW occurred from June 1996 through November 2001, followed by a program to flush the remaining HLW through to the melter. Glass removal and shutdown followed. The facility and process equipment is currently in a standby mode awaiting deactivation. During HLW processing operations, nearly 24 million curies of radioactive material were vitrified into 275 canisters of HLW glass. At least 99.7% of the curies in the HLW tanks at the WVDP were vitrified using the melter. Each canister of HLW holds approximately 2000 kilograms of glass with an average contact dose rate of over 2600 rem per hour. After vitrification processing ended, two more cans were filled using the Evacuated Canister Process to empty the melter at shutdown. This history briefly summarizes the initial stages of process development and earlier WVDP experience in the design and operation of the vitrification systems, followed by a more detailed discussion of equipment availability and failure rates during six years of operation. Lessons learned operating a system that continued to function beyond design expectations also are highlighted.

  18. Liquidus Temperature of High-Level Waste Borosilicate Glasses with Spinel Primary Phase

    SciTech Connect (OSTI)

    Hrma, Pavel R. (BATTELLE (PACIFIC NW LAB)); Vienna, John D. (BATTELLE (PACIFIC NW LAB)); Crum, Jarrod V. (BATTELLE (PACIFIC NW LAB)); Piepel, Gregory F. (BATTELLE (PACIFIC NW LAB)); Mika, Martin (ASSOC WESTERN UNIVERSITY); Robert W. Smith; David W. Shoesmith

    2000-01-01T23:59:59.000Z

    Liquidus temperatures (TL) were measured for high-level waste (HLW) borosilicate glasses covering a Savannah River composition region. The primary crystallization phase for most glasses was spinel, a solid solution of trevorite (NiFe2O4) with other oxides (FeO, MnO, and Cr2O3). The TL values ranged from 859 to 1310?C. Component additions increased the TL (per mass%) as Cr2O3 261?C, NiO 85?C, TiO2 42?C, MgO 33?C, Al2O3 18?C, and Fe2O3 18?C and decreased the TL (per mass%) as Na2O -29?C, Li2O -28?C, K2O -20?C, and B2O3 -8?C. Other oxides (CaO, MnO, SiO2, and U3O8) had little effect. The effect of RuO2 is not clear.

  19. Strategy for addressing composition uncertainties in a Hanford high-level waste vitrification plant

    SciTech Connect (OSTI)

    Bryan, M.F.; Piepel, G.F.

    1996-03-01T23:59:59.000Z

    Various requirements will be imposed on the feed material and glass produced by the high-level waste (HLW) vitrification plant at the Hanford Site. A statistical process/product control system will be used to control the melter feed composition and to check and document product quality. Two general types of uncertainty are important in HLW vitrification process/product control: model uncertainty and composition uncertainty. Model uncertainty is discussed by Hrma, Piepel, et al. (1994). Composition uncertainty includes the uncertainties inherent in estimates of feed composition and other process measurements. Because feed composition is a multivariate quantity, multivariate estimates of composition uncertainty (i.e., covariance matrices) are required. Three components of composition uncertainty will play a role in estimating and checking batch and glass attributes: batch-to-batch variability, within-batch uncertainty, and analytical uncertainty. This document reviews the techniques to be used in estimating and updating composition uncertainties and in combining these composition uncertainties with model uncertainty to yield estimates of (univariate) uncertainties associated with estimates of batch and glass properties.

  20. The Effect of Composition on Spinel Equilibrium and Crystal Size in High-Level Waste Glass

    SciTech Connect (OSTI)

    Wilson, B. K.; Hrma, Pavel R.; Alton, Jesse; Plaisted, Trevor J.; Vienna, John D.

    2002-12-15T23:59:59.000Z

    The equilibrium concentration (Co) of spinel was measured in 16 high-level waste (HLW) glasses as a function of temperature (T). Glasses were formulated by increasing or decreasing concentrations of Al2O3, Cr2O3, Fe2O3, Li2O, MgO, Na2O, or NiO, one-at-a-time, from a baseline composition. Data were fitted using the quasi-ideal-solution relationship between Co and T. The coefficients of this relationship were expressed as functions of glass composition using first-order approximation. All glass components had an effect on liquidus temperature (TL), but only NiO and Fe2O3 had a significant impact on spinel concentration below TL. The temperature at which Co had a given value was also expressed as a function of glass composition. These results can be used to optimize a HLW glass formulation to meet a constraint of either no spinel or a limited spinel fraction in the melter. In addition, the measurement of the size of spinel crystals and subsequent calculation of crystal number density (n) showed that Cr2O4 and Al2O3 increase n.

  1. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM 2009

    SciTech Connect (OSTI)

    West, B.; Waltz, R.

    2010-06-21T23:59:59.000Z

    Aqueous radioactive wastes from Savannah River Site (SRS) separations and vitrification processes are contained in large underground carbon steel tanks. Inspections made during 2009 to evaluate these vessels and other waste handling facilities along with evaluations based on data from previous inspections are the subject of this report. The 2009 inspection program revealed that the structural integrity and waste confinement capability of the Savannah River Site waste tanks were maintained. All inspections scheduled per LWO-LWE-2008-00423, HLW Tank Farm Inspection Plan for 2009, were completed. All Ultrasonic measurements (UT) performed in 2009 met the requirements of C-ESG-00006, In-Service Inspection Program for High Level Waste Tanks, Rev. 1, and WSRC-TR-2002-00061, Rev.4. UT inspections were performed on Tank 29 and the findings are documented in SRNL-STI-2009-00559, Tank Inspection NDE Results for Fiscal Year 2009, Waste Tank 29. Post chemical cleaning UT measurements were made in Tank 6 and the results are documented in SRNL-STI-2009-00560, Tank Inspection NDE Results Tank 6, Including Summary of Waste Removal Support Activities in Tanks 5 and 6. A total of 6669 photographs were made and 1276 visual and video inspections were performed during 2009. Twenty-Two new leaksites were identified in 2009. The locations of these leaksites are documented in C-ESR-G-00003, SRS High Level Waste Tank Leaksite Information, Rev.4. Fifteen leaksites at Tank 5 were documented during tank wall/annulus cleaning activities. Five leaksites at Tank 6 were documented during tank wall/annulus cleaning activities. Two new leaksites were identified at Tank 19 during waste removal activities. Previously documented leaksites were reactivated at Tanks 5 and 12 during waste removal activities. Also, a very small amount of additional leakage from a previously identified leaksite at Tank 14 was observed.

  2. The Italian Activities in the Field of Nuclear Waste Management - 12439

    SciTech Connect (OSTI)

    Giorgiantoni, Giorgio; Marzo, Giuseppe A.; Sepielli, Massimo [ENEA, C. R. Casaccia, Roma (Italy)

    2012-07-01T23:59:59.000Z

    The Italian situation in the field of nuclear waste management is characterized by a relative small quantity of wastes, as a consequence of the giving up of energy production by nuclear generation in 1986. Notwithstanding this situation, Italy is a unique case study since the country needs to undertake the final decommissioning of four shut-down NPPs (size 100-200 MWe), each one different from the others. Therefore all the regulatory, technical, and financial actions are needed in the same way as if there was actual nuclear generation. Furthermore, the various non-power generating applications of nuclear energy still require management, a legal framework, a regulatory body, an industrial structure, and technical know-how. Notwithstanding the absence of energy production from nuclear sources, the country has the burden of radioactive waste management from the previous nuclear operations, which obliges it to implement at first a robust legislative framework, then to explore all the complex procedures to achieve the localization of the national interim storage facility, not excluding the chance to have a European regional facility for geologic disposal, under the clauses of the Council Directive of 19 July 2011 'Establishing a Community Framework for the Responsible and Safe Management of Radioactive Waste'. Then, as far as industrial, medical and R and D aspects, the improvement of the legislative picture, the creation of a regulatory body, is a good start for the future, to achieve the best efficiency of the Italian system. (authors)

  3. Geochemical assessment of nuclear waste isolation. Report of activities during fiscal year 1982

    SciTech Connect (OSTI)

    Not Available

    1983-07-01T23:59:59.000Z

    The status of the following investigations is reported: canister/overpack-backfill chemical interactions and mechanisms; backfill and near-field host rock chemical interactions mechanisms; far-field host rock geochemical interactions; verification and improvement of predictive algorithms for radionuclide migration; and geologic systems as analogues for long-term radioactive waste isolation.

  4. Performance objectives for the Hanford immobilized low-activity waste (ILAW) performance assessment

    SciTech Connect (OSTI)

    MANN, F.M.

    1999-09-09T23:59:59.000Z

    Before low-level waste may be disposed of, a performance assessment must be written and then approved by the DOE (DOE 1988a DOE 1999a). The performance assessment is to determine whether ''reasonable assurance'' exists that the performance objectives of the disposal facility will be met. The DOE requirements for waste disposal (DOE 1988a DOE 1999a) require the protection of public health and safety; and the protection of the environment. Although quantitative limits are sometimes stated (for example, the all-pathways exposure limit is 25 mrem/year), usually the requirements are stated in a general nature. Quantitative limits were established by: investigating all potentially applicable regulations as well as interpretations of the review panels which DOE has established to review performance assessments, interacting with program management to establish the additional requirements of the program, and interacting with the public (i.e., the Hanford Advisory Board members; as well as affected Tribal Governments) to understand the values of residents in the Pacific Northwest. Because of space considerations, not all radionuclides and dangerous chemicals are listed in this document. The radionuclides listed here are those which were explicitly treated in the ILAW PA (Mann 1998). The dangerous chemicals listed here are those most often detected in Hanford tank waste as documented in the Regulatory Data Quality Objectives Supporting Tank Waste Remediation System Privatization Project (Wiemers 1998).

  5. The Mochovce final treatment center for liquid radioactive waste introduced to active trial operation

    SciTech Connect (OSTI)

    Krajc, T.; Stubna, M.; Kravarik, K.; Zatkulak, M. [VUJE Trnava, Inc. (Slovakia); Slezak, M.; Remias, V. [Javys - Jadrova a vyradovacia spolocnost, a.s. - Nuclear and Decommissioning Company, plc., Tomasikova 22, 821 02 Bratislava (Slovakia)

    2007-07-01T23:59:59.000Z

    The Final Treatment Centre (FTC) for Mochovce Nuclear Power Plant (NPP) have been designed for treatment and final conditioning of radioactive liquid and wet waste produced by named NPP equipped with Russian VVER-440 type of reactors. Treated wastes comprise radioactive concentrates, spent resin and sludge. VUJE Inc. as an experienced company in field of treatment of radioactive waste in Slovakia has been chosen as main contractor for technological part of FTC. During the realisation of project the future operator of Centre required the contractor to solve the treatment of wastes produced in the process of NPP A-1 decommissioning. On the basis of this requirement the project was modified in order to enable manipulations with waste products from A-1 NPP transported to Centre in steel drums. The initial project was prepared in 2003. The design and manufacture of main components were performed in 2004 and 2005. FTC civil works started in August 2004. Initial nonradioactive testing of the system parts were carried out from April to September 2006, then the tests of systems started with model concentrates and non-radioactive resins. After the processes evaluation the radioactive test performed from February 2007. A one-year trial operation of facility is planned for completion during 2007 and 2008. The company JAVYS, Inc. is responsible for radioactive waste and spent fuel treatment in the Slovak republic and will operate the FTC during trial operation and after its completion. This Company has also significant experience with operation of Jaslovske Bohunice Treatment Centre. The overall capacity of the FTC is 820 m{sup 3}/year of concentrates and 40 m{sup 3}/year of spent resin and sludge. Bituminization and cementation were provided as main technologies for treatment of these wastes. Treatment of concentrate is performed by bituminization on Thin Film Evaporator with rotating wiping blades. Spent resin and sludge are decanted, dried and mixed with bitumen in blade homogeniser. The bitumen product is discharged into 200 dm{sup 3} steel drums. Drums with bitumen product or drums originated from A-1 NPP are loaded into Fibre Reinforced Concrete containers (FRC) and grouted with cement. Cement grout is prepared from the mixture of cement, additive and radioactive over-concentrate. By formulating the cement grout with evaporator concentrates the maximum radioactivity is fixed in cement matrix and volume of final waste product is minimized. A batch mixer with rotating blades is used to produce the cement grout. The grouted FRC containers are stored in the expedition hall and after 28 days of curing are transported to final disposal. After the start of routine operation, the FTC provides treatment for all liquid and wet LLW produced from the operation of the Mochovce NPP. The final product of the FTC is a FRC loaded with bitumen product in drums and filled with radioactive cement product. This container meets all limits for final disposal in the National Radioactive Waste Repository at Mochovce. This paper introducing the main parts of FTC and describes the technological procedures including the basic technological parameters for both used technologies, their working capacity and the overall waste flow. The evaluation of experience gained in the phases of Centre construction and commissioning and partially trial operation as well is a part of this paper (Evaluation of completion works process and time schedule, the process of individual system parts testing, testing of systems using model media, radioactive testing and trial operation). (authors)

  6. EIS-0081: Long-Term Management of Liquid High-Level Radioactive Waste Stored at Western New York Nuclear Service Center, West Valley, New York

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy Office of Terminal Waste Disposal and Remedial Action prepared this environmental impact statement to analyze the environmental and socioeconomic impacts resulting from the Department’s proposed action to construct and operate facilities necessary to solidify the liquid high-level wastes currently stored in underground tanks at West Valley, New York.

  7. Assessment of chemical vulnerabilities in the Hanford high-level waste tanks

    SciTech Connect (OSTI)

    Meacham, J.E. [and others

    1996-02-15T23:59:59.000Z

    The purpose of this report is to summarize results of relevant data (tank farm and laboratory) and analysis related to potential chemical vulnerabilities of the Hanford Site waste tanks. Potential chemical safety vulnerabilities examined include spontaneous runaway reactions, condensed phase waste combustibility, and tank headspace flammability. The major conclusions of the report are the following: Spontaneous runaway reactions are not credible; condensed phase combustion is not likely; and periodic releases of flammable gas can be mitigated by interim stabilization.

  8. Acceleration of Los Alamos National Laboratory transuranic waste disposition

    SciTech Connect (OSTI)

    O'Leary, G.A.; Palmer, B.A.; Starke, T.P.; Phelps, A.K. [Los Alamos National Security, L.L.C., Los Alamos National Laboratory, Los Alamos, NM (United States)

    2007-07-01T23:59:59.000Z

    One of Los Alamos National Laboratory's (LANL's) most significant risks is the site's inventory of transuranic waste retrievably stored above and below-ground in Technical Area (TA) 54 Area G, particularly the dispersible high-activity waste stored above-ground in deteriorating facilities. The high activity waste represents approximately 50% (by activity) of the total 292,000 PE-Ci inventory remaining to be disposed. The transuranic waste inventory includes contact-handled and remote-handled waste packaged in drums, boxes, and oversized containers which are retrievably stored both above and below-ground. Although currently managed as transuranic waste, some of the inventory is low-level waste that can be disposed onsite or at approved offsite facilities. Dis-positioning the transuranic waste inventory requires retrieval of the containers from above and below- ground storage, examination and repackaging or remediation as necessary, characterization, certification and loading for shipment to the Waste Isolation Pilot Plant in Carlsbad, New Mexico, all in accordance with well-defined requirements and controls. Although operations are established to process and characterize the lower-activity contact-handled transuranic waste containers, LANL does not currently have the capability to repack high activity contact-handled transuranic waste containers (> 56 PE-Ci) or to process oversized containers with activity levels over 0.52 PE-Ci. Operational issues and compliance requirements have resulted in less than optimal processing capabilities for lower activity contact-handled transuranic waste containers, limiting preparation and reducing dependability of shipments to the Waste Isolation Pilot Plant. Since becoming the Los Alamos National Laboratory contractor in June 2006, Los Alamos National Security (LANS) L.L.C. has developed a comprehensive, integrated plan to effectively and efficiently disposition the transuranic waste inventory, working in concert with the Department of Energy Los Alamos Site Office, Carlsbad Field Office and the Department of Energy Headquarters. Rather than simply processing containers as retrieved, the plan places priority on efficient curie disposition, a direct correlation to reducing risk. Key elements of the approach include balancing inventory and operational risks, tailoring methods to meet requirements, optimizing existing facilities, equipment and staff, and incorporating best practices from other Department of Energy sites. With sufficient funding this will enable LANL to ship the above-ground high activity contact-handled transuranic waste offsite by the end of Fiscal Year (FY) 2007 and to disposition the remaining above- and below-ground contact-handled and remote-handled transuranic waste inventory by December 2010. Nearly 70% of the contact-handled transuranic waste containers, including the high activity waste, require processing and repackaging before characterization and certification for shipment to the Waste Isolation Pilot Plant. LANL is employing a balanced risk approach that accomplishes significant long-term risk reduction by accepting short-term increased facility operations risk under well-developed and justified interim controls. Reviews of facility conditions and additional analyses show that the Waste Characterization, Reduction and Repackaging Facility and the Radioassay and Nondestructive Testing Facility are the most appropriate facilities to safely remediate, repackage, and ship lower activity and the remaining high activity drums. Updated safety documentation supporting limited Hazard Category 2 operations in these facilities has been developed. Once approved, limited-term operations to process the high activity drums can begin in early 2007, building upon the experience base established performing Hazard Category 3 operations processing lower activity waste in these facilities. LANL is also implementing a series of actions to improve and sustain operations for processing contact-handled transuranic waste inventory. Building 412 Decontamination and Volume Reduction Fa

  9. ACCELERATION OF LOS ALAMOS NATIONAL LABORATORY TRANSURANIC WASTE DISPOSITION

    SciTech Connect (OSTI)

    O'LEARY, GERALD A. [Los Alamos National Laboratory

    2007-01-04T23:59:59.000Z

    One of Los Alamos National Laboratory's (LANL's) most significant risks is the site's inventory of transuranic waste retrievably stored above and below-ground in Technical Area (TA) 54 Area G, particularly the dispersible high-activity waste stored above-ground in deteriorating facilities. The high activity waste represents approximately 50% (by activity) of the total 292,000 PE-Ci inventory remaining to be disposed. The transuramic waste inventory includes contact-handled and remote-handled waste packaged in drums, boxes, and oversized containers which are retrievably stored both above and below-ground. Although currently managed as transuranic waste, some of the inventory is low-level waste that can be disposed onsite or at approved offsite facilities. Dispositioning the transuranic waste inventory requires retrieval of the containers from above and below-ground storage, examination and repackaging or remediation as necessary, characterization, certification and loading for shipment to the Waste Isolation Pilot Plant in Carlsbad New Mexico, all in accordance with well-defined requirements and controls. Although operations are established to process and characterize the lower-activity contact-handled transuranic waste containers, LAN L does not currently have the capability to repack high activity contact-handled transuranic waste containers (> 56 PE-Ci) or to process oversized containers with activity levels over 0.52 PE-Ci. Operational issues and compliance requirements have resulted in less than optimal processing capabilities for lower activity contact-handled transuranic waste containers, limiting preparation and reducing dependability of shipments to the Waste Isolation Pilot Plant. Since becoming the Los Alamos National Laboratory contract in June 2006, Los Alamos National Security (LANS) L.L.C. has developed a comprehensive, integrated plan to effectively and efficiently disposition the transuranic waste inventory, working in concert with the Department of Energy Los Alamos Site Office, Carlsbad Field Office and the Department of Energy Headquaeters. Rather than simply processing containers as retrieved, the plan places priority on efficient curie disposition, a direct correlation to reducing risk. Key elements of the approch include balancing inventory and operational risks, tailoring methods to meet requirements, optimizing existing facilities, equipment and staff, and incorporating best practices from other Department of Energy sites. With sufficient funding this will enable LANL to ship the above-ground high activity contact-handled transuranic waste offsite by the end of Fiscal Year (FY) 2007 and to disposition the remaining above- and below-ground contact-handled and remote-handled transuranic waste inventory by December 2010. Nearly 70% of the contact-handled transuranic waste containers, including the high activity waste, require processing and repackaging before characterization and certification for shipment to the Waste Isolation Pilot Plant. LANL is employing a balanced risk approach that accomplishes significant long-term risk reduction by accepting short-term increased facility operations risk under well-developed and justified interim controls. Reviews of facility conditions and additional analyses show that the Waste Characterization, Reduction and Repackaging Facility and the Radioassay and Nondestructive Testing Facility are the most appropriate facilities to safetly remediate, repackage, and ship lower activity and the remaining high activity drums. Updated safety documentation supporting limited Hazard Category 2 operations in these facilities has been developed. Once approved, limited-term operations to process the high activity drums can begin in early 2007, building upon the experience base established performing Hazard Category 3 operations processing lower activity waste in these facilities. LANL is also implementing a series of actions to improve and sustain operations for processing contact-handled transuranic waste inventory. Building 412 Decontamination and Volume Facility and Dom

  10. Radioactive Waste Management Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1999-07-09T23:59:59.000Z

    This Manual further describes the requirements and establishes specific responsibilities for implementing DOE O 435.1, Radioactive Waste Management, for the management of DOE high-level waste, transuranic waste, low-level waste, and the radioactive component of mixed waste. Change 1 dated 6/19/01 removes the requirement that Headquarters is to be notified and the Office of Environment, Safety and Health consulted for exemptions for use of non-DOE treatment facilities. Certified 1-9-07.

  11. Effect of feed melting, temperature history, and minor component addition on spinel crystallization in high-level waste glass

    SciTech Connect (OSTI)

    Izak, Pavel (ASSOC WESTERN UNIVERSITY) [ASSOC WESTERN UNIVERSITY; Hrma, Pavel R.(BATTELLE (PACIFIC NW LAB)) [BATTELLE (PACIFIC NW LAB); Arey, Bruce W.(BATTELLE (PACIFIC NW LAB)) [BATTELLE (PACIFIC NW LAB); Plaisted, Trevor J.(ASSOC WESTERN UNIVERSITY) [ASSOC WESTERN UNIVERSITY

    2001-01-01T23:59:59.000Z

    Spinel crystallization affects the anticipated cost and risk of high-level waste (HLW) vitrification. Spinel, (Fe,Ni) (Fe,Cr)2O4, is the primary crystalline phase that precipitates from melts containing Fe and Ni in sufficient concentrations. This study was undertaken to help design and verify mathematical models for a HLW glass melter in which spinel crystals precipitate and partially settle.

  12. CHEMICAL ANALYSIS OF SIMULATED HIGH LEVEL WASTE GLASSES TO SUPPORT SULFATE SOLUBILITY MODELING

    SciTech Connect (OSTI)

    Fox, K.; Marra, J.

    2014-08-14T23:59:59.000Z

    The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms both within the DOE complex and to some extent at U.K. sites. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerated cleanup missions. Much of the previous work on improving sulfate retention in waste glasses has been done on an empirical basis, making it difficult to apply the findings to future waste compositions despite the large number of glass systems studied. A more fundamental, rather than empirical, model of sulfate solubility in glass, under development at Sheffield Hallam University (SHU), could provide a solution to the issues of sulfate solubility. The model uses the normalized cation field strength index as a function of glass composition to predict sulfate capacity, and has shown early success for some glass systems. The objective of the current scope is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DOE waste vitrification efforts, allowing for enhanced waste loadings and waste throughput. A series of targeted glass compositions was selected to resolve data gaps in the current model. SHU fabricated these glasses and sent samples to the Savannah River National Laboratory (SRNL) for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for simulated waste glasses fabricated SHU in support of sulfate solubility model development. A review of the measured compositions revealed that there are issues with the B{sub 2}O{sub 3} and Fe{sub 2}O{sub 3} concentrations missing their targeted values by a significant amount for several of the study glasses. SHU is reviewing the fabrication of these glasses and the chemicals used in batching them to identify the source of these issues. The measured sulfate concentrations were all below their targeted values. This is expected, as the targeted concentrations likely exceeded the solubility limit for sulfate in these glass compositions. Some volatilization of sulfate may also have occurred during fabrication of the glasses. Measurements of the other oxides in the study glasses were reasonably close to their targeted values

  13. Semi-continuous anaerobic co-digestion of thickened waste activated sludge and fat, oil and grease

    SciTech Connect (OSTI)

    Wan Caixia; Zhou Quancheng; Fu Guiming [Department of Food, Agricultural, and Biological Engineering, Ohio State University/Ohio Agricultural Research and Development Center, 1680 Madison Ave., Wooster, OH 44691-4096 (United States); Li Yebo, E-mail: li.851@osu.edu [Department of Food, Agricultural, and Biological Engineering, Ohio State University/Ohio Agricultural Research and Development Center, 1680 Madison Ave., Wooster, OH 44691-4096 (United States)

    2011-08-15T23:59:59.000Z

    Highlights: > Co-digestion of thickened waste activated sludge (TWAS) with fat, oil and grease (FOG). > Co-digestion of TWAS and FOG at 64% VS increased biogas production by 137%. > FOG addition ratio at 74% of total VS caused inhibition of the anaerobic digestion process. > Micronutrients addition did not significantly improve the biogas production and digestion stabilization. - Abstract: Co-digestion of thickened waste activated sludge (TWAS) and fat, oil and grease (FOG) was conducted semi-continuously under mesophilic conditions. The results showed that daily methane yield at the steady state was 598 L/kg VS{sub added} when TWAS and FOG (64% of total VS) were co-digested, which was 137% higher than that obtained from digestion of TWAS alone. The biogas composition was stabilized at a CH{sub 4} and CO{sub 2} content of 66.8% and 29.5%, respectively. Micronutrients added to co-digestion did not improve the biogas production and digestion stabilization. With a higher addition of FOG (74% of total VS), the digester initially failed but was slowly self-recovered; however, the methane yield was only about 50% of a healthy reactor with the same organic loading rate.

  14. A Strategy to Conduct an Analysis of the Long-Term Performance of Low-Activity Waste Glass in a Shallow Subsurface Disposal System at Hanford

    SciTech Connect (OSTI)

    BP McGrail, WL Ebert, DH Bacon, DM Strachan

    1998-02-18T23:59:59.000Z

    Privatized services are being procured to vitrify low-activity tank wastes for eventual disposal in a shallow subsurface facility at the Hanford Site. Over 500,000 metric tons of low-activity waste glass will be generated, which is among the largest volumes of waste within the U.S. Department of Energy (DOE) complex and is one of the largest inventories of long-lived radionuclides planned for disposal in a low-level waste facility. Before immobilized waste can be disposed, DOE must approve a "performance assessment," which is a document that describes the impacts of the disposal facility on public health and environmental resources. Because the release rate of radionuclides from the glass waste form is a key factor determining these impacts, a sound scientific basis for determining their long-term release rates must be developed if this disposal action is to be accepted by regulatory agencies, stakeholders, and the public. In part, the scientific basis is determined from a sound testing strategy. The foundation of the proposed testing strategy is a well accepted mechanistic model that is being used to calculate the glass corrosion behavior over the geologic time scales required for performance assessment. This model requires that six parameters be determined, and the testing program is defined by an appropriate set of laboratory experiments to determine these parameters, and is combined with a set of field experiments to validate the model as a whole. Three general classes of laboratory tests are proposed in this strategy: 1) characterization, 2) accelerated, and 3) service condition. Characterization tests isolate and provide specific information about processes or parameters in theoretical models. Accelerated tests investigate corrosion behavior that will be important over the regulated service life of a disposal system within a laboratory time frame of a few years or less. Service condition tests verify that the techniques used in accelerated tests do not change the alteration mechanisms. The recommended characterization tests are single-pass flow-through tests using a batch reactor design, Accelerated and service conditions tests include product consistency and pressurized unsaturated flow (PUF) tests. Nonradioactive glasses will be used for the majority of the laboratory testing (-80%), with the remainder performed with glasses containing a selected set of key radionuclides. Additionally, a series of PUF experiments with a natural analog of basaltic glass is recommended to confirm that the alteration products observed under accelerated conditions in the PUF tests are similar to those found associated with the natural analog. This will provide additional confidence in using the PUF test results to infer long-term corrosion behavior. Field tests are proposed as a unique way to validate the glass corrosion and contaminant transport models being used in the performance assessment. To better control the test conditions, the field tests are to be performed in lysimeters (corrugated steel containers buried flush with the ground surface). Lysimeters provide a way to combine a glass, Hanford soil, and perhaps other engineered materials in a well-controlled test, but on a scale that is not practicable in the laboratory. The recommended field tests include some experiments where a steady flow rate of water is artificially applied. These tests use a glass designed to have a high corrosion rate so that it is easier to monitor contaminant release and transport. Existing lysimeters at the Hanford Site can be used for these experiments or new lysimeters that have been equipped with the latest in monitoring equipment and located near the proposed disposal site.

  15. DATA SHARING REPORT CHARACTERIZATION OF POPULATION 7: PERSONAL PROTECTIVE EQUIPMENT, DRY ACTIVE WASTE, AND MISCELLANEOUS DEBRIS, SURVEILLANCE AND MAINTENANCE PROJECT OAK RIDGE NATIONAL LABORATORY OAK RIDGE, TENNESSEE

    SciTech Connect (OSTI)

    Harpenau, Evan M

    2013-10-10T23:59:59.000Z

    The U.S. Department of Energy (DOE) Oak Ridge Office of Environmental Management (EM-OR) requested that Oak Ridge Associated Universities (ORAU), working under the Oak Ridge Institute for Science and Education (ORISE) contract, provide technical and independent waste management planning support under the American Recovery and Reinvestment Act (ARRA). Specifically, DOE EM-OR requested that ORAU plan and implement a sampling and analysis campaign targeting certain URS|CH2M Oak Ridge, LLC (UCOR) surveillance and maintenance (S&M) process inventory waste. Eight populations of historical and reoccurring S&M waste at the Oak Ridge National Laboratory (ORNL) have been identified in the Waste Handling Plan for Surveillance and Maintenance Activities at the Oak Ridge National Laboratory, DOE/OR/01-2565&D2 (WHP) (DOE 2012) for evaluation and processing to determine a final pathway for disposal. Population 7 (POP 7) consists of 56 containers of aged, low-level and potentially mixed S&M waste that has been staged in various locations around ORNL. Several of these POP 7 containers primarily contain personal protective equipment (PPE) and dry active waste (DAW), but may contain other miscellaneous debris. This data sharing report addresses the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) specified waste in a 13-container subpopulation (including eight steel boxes, three 55-gal drums, one sealand, and one intermodal) that lacked sufficient characterization data for possible disposal at the Environmental Management Waste Management Facility (EMWMF) using the approved Waste Lot (WL) 108.1 profile.

  16. Systems analysis approach to the disposal of high-level waste in deep ocean sediments

    SciTech Connect (OSTI)

    de Marsily, G.; Hill, M. D.; Murray, C. N.; Talbert, D. M.; Van Dorp, F.; Webb, G. A.M.

    1980-01-01T23:59:59.000Z

    Among the different options being studied for disposal of high-level solidified waste, increasing attention is being paid to that of emplacement of glasses incorporating the radioactivity in deep oceanic sediments. This option has the advantage that the areas of the oceans under investigation appear to be relatively unproductive biologically, are relatively free from cataclysmic events, and are areas in which the natural processes are slow. Thus the environment is stable and predictable so that a number of barriers to the release and dispersion of radioactivity can be defined. Task Groups set up in the framework of the International Seabed Working Group have been studying many aspects of this option since 1976. In order that the various parts of the problem can be assessed within an integrated framework, the methods of systems analysis have been applied. In this paper the Systems Analysis Task Group members report the development of an overall system model. This will be used in an iterative process in which a preliminary analysis, together with a sensitivity analysis, identifies the parameters and data of most importance. The work of the other task groups will then be focussed on these parameters and data requirements so that improved results can be fed back into an improved overall systems model. The major requirements for the development of a preliminary overall systems model are that the problem should be separated into identified elements and that the interfaces between the elements should be clearly defined. The model evolved is deterministic and defines the problem elements needed to estimate doses to man.

  17. Hierarchical nanostructured conducting polymer hydrogel with high electrochemical activity

    E-Print Network [OSTI]

    Cui, Yi

    and have been used in many applications such as bioelectronics and energy storage devices. They are often demonstrated great potential for a broad range of applications from energy storage devices such as biofuelHierarchical nanostructured conducting polymer hydrogel with high electrochemical activity Lijia

  18. Evaluation of Flygt Propeller Xixers for Double Shell Tank (DST) High Level Waste Auxiliary Solids Mobilization

    SciTech Connect (OSTI)

    PACQUET, E.A.

    2000-07-20T23:59:59.000Z

    The River Protection Project (RPP) is planning to retrieve radioactive waste from the single-shell tanks (SST) and double-shell tanks (DST) underground at the Hanford Site. This waste will then be transferred to a waste treatment plant to be immobilized (vitrified) in a stable glass form. Over the years, the waste solids in many of the tanks have settled to form a layer of sludge at the bottom. The thickness of the sludge layer varies from tank to tank, from no sludge or a few inches of sludge to about 15 ft of sludge. The purpose of this technology and engineering case study is to evaluate the Flygt{trademark} submersible propeller mixer as a potential technology for auxiliary mobilization of DST HLW solids. Considering the usage and development to date by other sites in the development of this technology, this study also has the objective of expanding the knowledge base of the Flygt{trademark} mixer concept with the broader perspective of Hanford Site tank waste retrieval. More specifically, the objectives of this study delineated from the work plan are described.

  19. Conditioning matrices from high level waste resulting from pyrochemical processing in fluorine salt

    SciTech Connect (OSTI)

    Grandjean, Agnes; Advocat, Thierry; Bousquet, Nicolas [SCDV - Service de Confinement des Dechets et Vitrification - Laboratoire d'Etudes de Base sur les Verres, CEA Valrho, Centre de Marcoule, 30207 Bagnols sur Ceze (France); Jegou, Christophe [SECM - Service d'Etude du Confinement et Materiaux - Laboratoire des Materiaux et Procedes Actifs - CEA Valrho, Centre de Marcoule, 30207 Bagnols sur Ceze (France)

    2007-07-01T23:59:59.000Z

    Separating the actinides from the fission products through reductive extraction by aluminium in a LiF/AlF{sub 3} medium is a process investigated for pyrometallurgical reprocessing of spent fuel. The process involves separation by reductive salt-metal extraction. After dissolving the fuel or the transmutation target in a salt bath, the noble metal fission products are first extracted by contacting them with a slightly reducing metal. After extracting the metal fission products, then the actinides are selectively separated from the remaining fission products. In this hypothesis, all the unrecoverable fission products would be conditioned as fluorides. Therefore, this process will generate first a metallic waste containing the 'reducible' fission products (Pd, Mo, Ru, Rh, Tc, etc.) and a fluorine waste containing alkali-metal, alkaline-earth and rare earth fission products. Immobilization of these wastes in classical borosilicate glasses is not feasible due to the very low solubility of noble metals, and of fluoride in these hosts. Alternative candidates have therefore been developed including silicate glass/ceramic system for fluoride fission products and metallic ones for noble metal fission products. These waste-forms were evaluated for their confinement properties like homogeneity, waste loading, volatility during the elaboration process, chemical durability, etc. using appropriate techniques. (authors)

  20. Understanding radioactive waste

    SciTech Connect (OSTI)

    Murray, R.L.

    1981-12-01T23:59:59.000Z

    This document contains information on all aspects of radioactive wastes. Facts are presented about radioactive wastes simply, clearly and in an unbiased manner which makes the information readily accessible to the interested public. The contents are as follows: questions and concerns about wastes; atoms and chemistry; radioactivity; kinds of radiation; biological effects of radiation; radiation standards and protection; fission and fission products; the Manhattan Project; defense and development; uses of isotopes and radiation; classification of wastes; spent fuels from nuclear reactors; storage of spent fuel; reprocessing, recycling, and resources; uranium mill tailings; low-level wastes; transportation; methods of handling high-level nuclear wastes; project salt vault; multiple barrier approach; research on waste isolation; legal requiremnts; the national waste management program; societal aspects of radioactive wastes; perspectives; glossary; appendix A (scientific American articles); appendix B (reference material on wastes). (ATT)

  1. Implementing waste minimization at an active plutonium processing facility: Successes and progress at technical area (TA) -55 of the Los Alamos National Laboratory

    SciTech Connect (OSTI)

    Balkey, J.J.; Robinson, M.A.; Boak, J.

    1997-12-01T23:59:59.000Z

    The Los Alamos National Laboratory has ongoing national security missions that necessitate increased plutonium processing. The bulk of this activity occurs at Technical Area -55 (TA-55), the nations only operable plutonium facility. TA-55 has developed and demonstrated a number of technologies that significantly minimize waste generation in plutonium processing (supercritical CO{sub 2}, Mg(OH){sub 2} precipitation, supercritical H{sub 2}O oxidation, WAND), disposition of excess fissile materials (hydride-dehydride, electrolytic decontamination), disposition of historical waste inventories (salt distillation), and Decontamination & Decommissioning (D&D) of closed nuclear facilities (electrolytic decontamination). Furthermore, TA-55 is in the process of developing additional waste minimization technologies (molten salt oxidation, nitric acid recycle, americium extraction) that will significantly reduce ongoing waste generation rates and allow volume reduction of existing waste streams. Cost savings from reduction in waste volumes to be managed and disposed far exceed development and deployment costs in every case. Waste minimization is also important because it reduces occupational exposure to ionizing radiation, risks of transportation accidents, and transfer of burdens from current nuclear operations to future generations.

  2. RCRA Part A and Part B Permit Application for Waste Management Activities at the Nevada Test Site: Proposed Mixed Waste Disposal Unit (MWSU)

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2010-07-19T23:59:59.000Z

    The proposed Mixed Waste Storage Unit (MWSU) will be located within the Area 5 Radioactive Waste Management Complex (RWMC). Existing facilities at the RWMC will be used to store low-level mixed waste (LLMW). Storage is required to accommodate offsite-generated LLMW shipped to the Nevada Test Site (NTS) for disposal in the new Mixed Waste Disposal Unit (MWDU) currently in the design/build stage. LLMW generated at the NTS (onsite) is currently stored on the Transuranic (TRU) Pad (TP) in Area 5 under a Mutual Consent Agreement (MCA) with the Nevada Division of Environmental Protection, Bureau of Federal Facilities (NDEP/BFF). When the proposed MWSU is permitted, the U.S. Department of Energy (DOE) will ask that NDEP revoke the MCA and onsite-generated LLMW will fall under the MWSU permit terms and conditions. The unit will also store polychlorinated biphenyl (PCB) waste and friable and non-friable asbestos waste that meets the acceptance criteria in the Waste Analysis Plan (Exhibit 2) for disposal in the MWDU. In addition to Resource Conservation and Recovery Act (RCRA) requirements, the proposed MWSU will also be subject to Department of Energy (DOE) orders and other applicable state and federal regulations. Table 1 provides the metric conversion factors used in this application. Table 2 provides a list of existing permits. Table 3 lists operational RCRA units at the NTS and their respective regulatory status.

  3. Waste Growth Challenges Local Democracy. The Politics of Waste between Europe and the Mediterranean: a Focus on Italy

    E-Print Network [OSTI]

    Mengozzi, Alessandro

    2010-01-01T23:59:59.000Z

    2008). ISWA. International Solid Waste Association. http://and Health Impact of Solid Waste Management Activities. InPerformance of Alternative Solid Waste Management Options: A

  4. An international initiative on long-term behavior of high-level nuclear waste glass

    SciTech Connect (OSTI)

    Gin, Stephane [CEA Marcoule DTCD SECM LCLT, Bagnols/Ceze (France); Abdelouas, Abdessalam [SUBATECH, Nantes (France); Criscenti, Louise J. [Sandia National Laboratories, Albuquerque, NM (United States); Ebert, W. L. [Argonne National Laboratory (ANL), Argonne, IL (United States); Ferrand, Karine [SCK·CEN, Mol (Belgium); Geisler, Thorsten [Rheinische Friedrich-Wilhelms-Univ., Bonn (Germany); Harrison, Mike T. [National Nuclear Laboratory, Sellafield, Cumbria (United Kingdom); Inagaki, Yaohiro [Kyushu Univ. (Japan). Dept. Appl. Quantum Physics and Nuclear Engineering; Mitsui, Seiichiro [Japan Atomic Energy Agency, Ibaraki (Japan); Mueller, Karl T. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States). Environmental and Molecular Science Lab.; Marra, James C. [Savannah River National Laboratory, Aiken, SC (United States); Pantano, Carlo G. [Penn State Univ., State College, PA (United States); Pierce, Eric M. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Ryan, Joseph V. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Schofield, James M. [AMEC, Harwell Oxford (United Kingdom); Steefel, Carl I. [Lawrence Berkeley National Laboratory (LBNL), Berkeley, CA (United States). Earth Sciences Div.; Vienna, John D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

    2013-06-01T23:59:59.000Z

    Nations producing borosilicate glass as an immobilization material for radioactive wastes resulting from spent nuclear fuel reprocessing have reinforced scientific collaboration to obtain consensus on mechanisms controlling the long-term dissolution rate of glass. This goal is deemed to be crucial for the development of reliable performance assessment models for geological disposal. The collaborating laboratories all conduct fundamental and/or applied research with modern materials science techniques. The paper briefly reviews the radioactive waste vitrification programmes of the six participant nations and summarizes the state-of-the-art of glass corrosion science, emphasizing common scientific needs and justifications for on-going initiatives.

  5. PERFORMANCE IMPROVEMENT OF CROSS-FLOW FILTRATION FOR HIGH LEVEL WASTE TREATMENT

    SciTech Connect (OSTI)

    Duignan, M.; Nash, C.; Poirier, M.

    2011-01-12T23:59:59.000Z

    In the interest of accelerating waste treatment processing, the DOE has funded studies to better understand filtration with the goal of improving filter fluxes in existing cross-flow equipment. The Savannah River National Laboratory (SRNL) was included in those studies, with a focus on start-up techniques, filter cake development, the application of filter aids (cake forming solid precoats), and body feeds (flux enhancing polymers). This paper discusses the progress of those filter studies. Cross-flow filtration is a key process step in many operating and planned waste treatment facilities to separate undissolved solids from supernate slurries. This separation technology generally has the advantage of self-cleaning through the action of wall shear stress created by the flow of waste slurry through the filter tubes. However, the ability of filter wall self-cleaning depends on the slurry being filtered. Many of the alkaline radioactive wastes are extremely challenging to filtration, e.g., those containing compounds of aluminum and iron, which have particles whose size and morphology reduce permeability. Unfortunately, low filter flux can be a bottleneck in waste processing facilities such as the Savannah River Modular Caustic Side Solvent Extraction Unit and the Hanford Waste Treatment Plant. Any improvement to the filtration rate would lead directly to increased throughput of the entire process. To date increased rates are generally realized by either increasing the cross-flow filter axial flowrate, limited by pump capacity, or by increasing filter surface area, limited by space and increasing the required pump load. SRNL set up both dead-end and cross-flow filter tests to better understand filter performance based on filter media structure, flow conditions, filter cleaning, and several different types of filter aids and body feeds. Using non-radioactive simulated wastes, both chemically and physically similar to the actual radioactive wastes, the authors performed several tests to demonstrate increases in filter performance. With the proper use of filter flow conditions and filter enhancers, filter flow rates can be increased over rates currently realized today.

  6. Radioactive Waste Radioactive Waste

    E-Print Network [OSTI]

    Slatton, Clint

    form · Separate liquid from solid · Radionuclide · Separate all but H3/C14 #12;#12;Radioactive Waste;Radioactive Waste H3/C14 solids Type B (non-incinerable) metal glass hazardous materials #12;#12;Radioactive#12;Radioactive Waste at UF Bldg 831 392-8400 #12;Radioactive Waste · Program is designed to

  7. Preliminary total-system analysis of a potential high-level nuclear waste repository at Yucca Mountain

    SciTech Connect (OSTI)

    Eslinger, P.W.; Doremus, L.A.; Engel, D.W.; Miley, T.B.; Murphy, M.T.; Nichols, W.E.; White, M.D. [Pacific Northwest Lab., Richland, WA (United States); Langford, D.W.; Ouderkirk, S.J. [Westinghouse Hanford Co., Richland, WA (United States)

    1993-01-01T23:59:59.000Z

    The placement of high-level radioactive wastes in mined repositories deep underground is considered a disposal method that would effectively isolate these wastes from the environment for long periods of time. This report describes modeling performed at PNL for Yucca Mountain between May and November 1991 addressing the performance of the entire repository system related to regulatory criteria established by the EPA in 40 CFR Part 191. The geologic stratigraphy and material properties used in this study were chosen in cooperation with performance assessment modelers at Sandia National Laboratories (SNL). Sandia modeled a similar problem using different computer codes and a different modeling philosophy. Pacific Northwest Laboratory performed a few model runs with very complex models, and SNL performed many runs with much simpler (abstracted) models.

  8. Laboratory Report on Performance Evaluation of Key Constituents during Pre-Treatment of High Level Waste Direct Feed

    SciTech Connect (OSTI)

    Huber, Heinz J.

    2013-06-24T23:59:59.000Z

    The analytical capabilities of the 222-S Laboratory are tested against the requirements for an optional start up scenario of the Waste Treatment and Immobilization Plant on the Hanford Site. In this case, washed and in-tank leached sludge would be sent directly to the High Level Melter, bypassing Pretreatment. The sludge samples would need to be analyzed for certain key constituents in terms identifying melter-related issues and adjustment needs. The analyses on original tank waste as well as on washed and leached material were performed using five sludge samples from tanks 241-AY-102, 241-AZ-102, 241-AN-106, 241-AW-105, and 241-SY-102. Additionally, solid phase characterization was applied to determine the changes in mineralogy throughout the pre-treatment steps.

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

    SciTech Connect (OSTI)

    Hansen, E

    2005-03-31T23:59:59.000Z

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

  10. Environmental evaluation of alternatives for long-term management of Defense high-level radioactive wastes at the Idaho Chemical Processing Plant

    SciTech Connect (OSTI)

    Not Available

    1982-09-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) is considering the selection of a strategy for the long-term management of the defense high-level wastes at the Idaho Chemical Processing Plant (ICPP). This report describes the environmental impacts of alternative strategies. These alternative strategies include leaving the calcine in its present form at the Idaho National Engineering Laboratory (INEL), or retrieving and modifying the calcine to a more durable waste form and disposing of it either at the INEL or in an offsite repository. This report addresses only the alternatives for a program to manage the high-level waste generated at the ICPP. 24 figures, 60 tables.

  11. Toward Understanding the Effect of Low-Activity Waste Glass Composition on Sulfur Solubility

    SciTech Connect (OSTI)

    Vienna, John D.; Kim, Dong-Sang; Muller, Isabelle S.; Piepel, Gregory F.; Kruger, Albert A.

    2014-10-01T23:59:59.000Z

    The concentration of sulfur in nuclear waste glass melter feed must be maintained below the point where salt accumulates on the melt surface. The allowable concentrations may range from 0.37 to over 2.05 weight percent (of SO3 on a calcined oxide basis). If the amount of sulfur exceeds its tolerance level a molten salt will accumulate and upset melter operations and potentially shorten melter useful life. Therefore relatively conservative limits have been placed on sulfur loading in melter feed which in-turn significantly impacts the amount of glass that will be produced, in particular at the Hanford site. Crucible-scale sulfur solubility data and scaled melter sulfur tolerance data have been collected on simulated Hanford waste glasses over the last 15 years. These data were compiled and analyzed. A model was developed to predict the solubility of SO3 in glass based on 312 individual glass compositions. This model was shown to well represent the data, accounting for over 80% of the variation in data and was well validated. The model was also found to accurately predict the tolerance for sulfur in melter feed based on 19 scaled melter tests. The model is appropriate for control of waste glass processing which includes uncertainty quantification. The model also gives quantitative estimates of component concentration effects on sulfur solubility. The components that most increase sulfur solubility are Li2O > V2O5 ? TiO2 < CaO < P2O5 ? ZnO. The components that most decrease sulfur solubility are Cl > Cr2O3 > SiO2 ? ZrO2 > Al2O3.

  12. INCONEL 690 CORROSION IN WTP (WASTE TREATMENT PLANT) HLW (HIGH LEVEL WASTE) GLASS MELTS RICH IN ALUMINUM & BISMUTH & CHROMIUM OR ALUMINUM/SODIUM

    SciTech Connect (OSTI)

    KRUGER AA; FENG Z; GAN H; PEGG IL

    2009-11-05T23:59:59.000Z

    Metal corrosion tests were conducted with four high waste loading non-Fe-limited HLW glass compositions. The results at 1150 C (the WTP nominal melter operating temperature) show corrosion performance for all four glasses that is comparable to that of other typical borosilicate waste glasses, including HLW glass compositions that have been developed for iron-limited WTP streams. Of the four glasses tested, the Bi-limited composition shows the greatest extent of corrosion, which may be related to its higher phosphorus content. Tests at higher suggest that a moderate elevation of the melter operating temperature (up to 1200 C) should not result in any significant increase in Inconel corrosion. However, corrosion rates did increase significantly at yet higher temperatures (1230 C). Very little difference was observed with and without the presence of an electric current density of 6 A/inch{sup 2}, which is the typical upper design limit for Inconel electrodes. The data show a roughly linear relationship between the thickness of the oxide scale on the coupon and the Cr-depletion depth, which is consistent with the chromium depletion providing the material source for scale growth. Analysis of the time dependence of the Cr depletion profiles measured at 1200 C suggests that diffusion of Cr in the Ni-based Inconel alloy controls the depletion depth of Cr inside the alloy. The diffusion coefficient derived from the experimental data agrees within one order of magnitude with the published diffusion coefficient data for Cr in Ni matrices; the difference is likely due to the contribution from faster grain boundary diffusion in the tested Inconel alloy. A simple diffusion model based on these data predicts that Inconel 690 alloy will suffer Cr depletion damage to a depth of about 1 cm over a five year service life at 1200 C in these glasses.

  13. Public acceptance activities for the development of new commercial low-level radioactive waste disposal capacity in the United States of America

    SciTech Connect (OSTI)

    Ozaki, C.B.; Scott, R.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1993-12-31T23:59:59.000Z

    In the US, the states are responsible for providing disposal capability for commercial low-level radioactive waste generated within their borders. Public acceptance of state activities toward developing this capability is a key factor in the ultimate success of state efforts. The states are using several different approaches to gain public acceptance for the location and development of new low-level radioactive waste disposal facilities. This presentation describes state efforts to gain public acceptance for siting and developing activities and discusses the lessons learned from these state experiences.

  14. Use of Novel Highly Selective Ion Exchange Media for Minimizing the Waste Arising from Different NPP and Other Liquids

    SciTech Connect (OSTI)

    Tusa, Esko; Harjula, Risto; Lehto, Jukka

    2003-02-25T23:59:59.000Z

    Highly selective inorganic ion exchangers give new possibilities to implement and operate new innovative treatment systems for radioactive liquids. Because of high selectivity these ion exchangers can be used even in liquids of high salt concentrations. Only selected target nuclides will be separated and inactive salts are left in the liquid, which can be released or recategorized. Thus, it is possible to reduce the volume of radioactive waste dramatically. On the other hand, only a small volume of highly selective material is required in applications, which makes it possible to design totally new types of compact treatment systems. The major benefit of selective ion exchange media comes from the very large volume reduction of radioactive waste in final disposal. It is also possible to save in investment costs, because small ion exchanger volumes can be used and handled in a very small facility. This paper describes different applications of these highly selective ion exchangers, both commercial fullscale applications and laboratory tests, to give the idea of their efficiency for different liquids.

  15. Hanford high level waste (HLW) tank mixer pump safe operating envelope reliability assessment

    SciTech Connect (OSTI)

    Fischer, S.R. [Los Alamos National Lab., NM (United States); Clark, J. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)

    1993-10-01T23:59:59.000Z

    The US Department of Energy and its contractor, Westinghouse Corp., are responsible for the management and safe storage of waste accumulated from processing defense reactor irradiated fuels for plutonium recovery at the Hanford Site. These wastes, which consist of liquids and precipitated solids, are stored in underground storage tanks pending final disposition. Currently, 23 waste tanks have been placed on a safety watch list because of their potential for generating, storing, and periodically releasing various quantities of hydrogen and other gases. Tank 101-SY in the Hanford SY Tank Farm has been found to release hydrogen concentrations greater than the lower flammable limit (LFL) during periodic gas release events. In the unlikely event that an ignition source is present during a hydrogen release, a hydrogen burn could occur with a potential to release nuclear waste materials. To mitigate the periodic gas releases occurring from Tank 101-SY, a large mixer pump currently is being installed in the tank to promote a sustained release of hydrogen gas to the tank dome space. An extensive safety analysis (SA) effort was undertaken and documented to ensure the safe operation of the mixer pump after it is installed in Tank 101-SY.1 The SA identified a need for detailed operating, alarm, and abort limits to ensure that analyzed safety limits were not exceeded during pump operations.

  16. COMBINED RETENTION OF MOLYBDENUM AND SULFUR IN SIMULATED HIGH LEVEL WASTE GLASS

    SciTech Connect (OSTI)

    Fox, K.

    2009-10-16T23:59:59.000Z

    This study was undertaken to investigate the effect of elevated sulfate and molybdenum concentrations in nuclear waste glasses. A matrix of 24 glasses was developed and the glasses were tested for acceptability based on visual observations, canister centerline-cooled heat treatments, and chemical composition analysis. Results from the chemical analysis of the rinse water from each sample were used to confirm the presence of SO{sup 2-}{sub 4} and MoO{sub 3} on the surface of glasses as well as other components which might form water soluble compounds with the excess sulfur and molybdenum. A simple, linear model was developed to show acceptable concentrations of SO{sub 4}{sup 2-} and MoO{sub 3} in an example waste glass composition. This model was constructed for scoping studies only and is not ready for implementation in support of actual waste vitrification. Several other factors must be considered in determining the limits of sulfate and molybdenum concentrations in the waste vitrification process, including but not limited to, impacts on refractory and melter component corrosion, effects on the melter off-gas system, and impacts on the chemical durability and crystallization of the glass product.

  17. Enzyme-polymer composites with high biocatalytic activity and stability

    SciTech Connect (OSTI)

    Kim, Jungbae; Kosto, Timothy J.; Manimala, Joseph C.; Nauman, E B.; Dordick, Jonathan S.

    2004-08-22T23:59:59.000Z

    We have applied vacuum-spraying and electrospinning to incorporate an enzyme into a polymer matrix, creating a novel and highly active biocatalytic composite. As a unique technical approach, enzymes were co-dissolved in toluene with polymers, and the solvent was then rapidly removed by injecting the mixture into a vacuum chamber or by electrospinning. Subsequent crosslinking of the enzyme with glutaraldehyde resulted in stable entrapped enzyme within the polymeric matrices. For example, an amorphous composite of alpha-chymotrypsin and polyethylene showed no significant loss of enzymatic activity in aqueous buffer for one month. Nanofibers of alpha-chymotrypsin and polystyrene also showed no decrease in activity for more than two weeks. The normalized activity of amorphous composite in organic solvents was 3-13 times higher than that of native alpha-chymotrypsin. The activity of nanofibers was 5-7 times higher than that of amorphous composite in aqueous buffer solution. The composites of alpha-chymotrypsin and polymers demonstrate the feasibility of obtaining a wide variety of active and stable biocatalytic materials with many combinations of enzymes and polymers.

  18. Symmetric Active/Active Metadata Service for High Availability Parallel File Systems

    SciTech Connect (OSTI)

    He, X. [Tennessee Technological University; Ou, Li [Tennessee Technological University; Engelmann, Christian [ORNL; Chen, Xin [Tennessee Technological University; Scott, Stephen L [ORNL

    2009-01-01T23:59:59.000Z

    High availability data storage systems are critical for many applications as research and business become more data-driven. Since metadata management is essential to system availability, multiple metadata services are used to improve the availability of distributed storage systems. Past research focused on the active/standby model, where each active service has at least one redundant idle backup. However, interruption of service and even some loss of service state may occur during a fail-over depending on the used replication technique. In addition, the replication overhead for multiple metadata services can be very high. The research in this paper targets the symmetric active/active replication model, which uses multiple redundant service nodes running in virtual synchrony. In this model, service node failures do not cause a fail-over to a backup and there is no disruption of service or loss of service state. We further discuss a fast delivery protocol to reduce the latency of the needed total order broadcast. Our prototype implementation shows that metadata service high availability can be achieved with an acceptable performance trade-off using our symmetric active/active metadata service solution.

  19. ROLE OF MANGANESE REDUCTION/OXIDATION (REDOX) ON FOAMING AND MELT RATE IN HIGH LEVEL WASTE (HLW) MELTERS (U)

    SciTech Connect (OSTI)

    Jantzen, C; Michael Stone, M

    2007-03-30T23:59:59.000Z

    High-level nuclear waste is being immobilized at the Savannah River Site (SRS) by vitrification into borosilicate glass at the Defense Waste Processing Facility (DWPF). Control of the Reduction/Oxidation (REDOX) equilibrium in the DWPF melter is critical for processing high level liquid wastes. Foaming, cold cap roll-overs, and off-gas surges all have an impact on pouring and melt rate during processing of high-level waste (HLW) glass. All of these phenomena can impact waste throughput and attainment in Joule heated melters such as the DWPF. These phenomena are caused by gas-glass disequilibrium when components in the melter feeds convert to glass and liberate gases such as H{sub 2}O vapor (steam), CO{sub 2}, O{sub 2}, H{sub 2}, NO{sub x}, and/or N{sub 2}. During the feed-to-glass conversion in the DWPF melter, multiple types of reactions occur in the cold cap and in the melt pool that release gaseous products. The various gaseous products can cause foaming at the melt pool surface. Foaming should be avoided as much as possible because an insulative layer of foam on the melt surface retards heat transfer to the cold cap and results in low melt rates. Uncontrolled foaming can also result in a blockage of critical melter or melter off-gas components. Foaming can also increase the potential for melter pressure surges, which would then make it difficult to maintain a constant pressure differential between the DWPF melter and the pour spout. Pressure surges can cause erratic pour streams and possible pluggage of the bellows as well. For these reasons, the DWPF uses a REDOX strategy and controls the melt REDOX between 0.09 {le} Fe{sup 2+}/{summation}Fe {le} 0.33. Controlling the DWPF melter at an equilibrium of Fe{sup +2}/{summation}Fe {le} 0.33 prevents metallic and sulfide rich species from forming nodules that can accumulate on the floor of the melter. Control of foaming, due to deoxygenation of manganic species, is achieved by converting oxidized MnO{sub 2} or Mn{sub 2}O{sub 3} species to MnO during melter preprocessing. At the lower redox limit of Fe{sup +2}/{summation}Fe {approx} 0.09 about 99% of the Mn{sup +4}/Mn{sup +3} is converted to Mn{sup +2}. Therefore, the lower REDOX limits eliminates melter foaming from deoxygenation.

  20. APPLICATION OF A THIN FILM EVAPORATOR SYSTEM FOR MANAGEMENT OF LIQUID HIGH-LEVEL WASTES AT HANFORD

    SciTech Connect (OSTI)

    TEDESCHI AR; WILSON RA

    2010-01-14T23:59:59.000Z

    A modular, transportable evaporator system, using thin film evaporative technology, is planned for deployment at the Hanford radioactive waste storage tank complex. This technology, herein referred to as a wiped film evaporator (WFE), will be located at grade level above an underground storage tank to receive pumped liquids, concentrate the liquid stream from 1.1 specific gravity to approximately 1.4 and then return the concentrated solution back into the tank. Water is removed by evaporation at an internal heated drum surface exposed to high vacuum. The condensed water stream will be shipped to the site effluent treatment facility for final disposal. This operation provides significant risk mitigation to failure of the aging 242-A Evaporator facility; the only operating evaporative system at Hanford maximizing waste storage. This technology is being implemented through a development and deployment project by the tank farm operating contractor, Washington River Protection Solutions (WRPS), for the Office of River Protection/Department of Energy (ORP/DOE), through Columbia Energy & Environmental Services, Inc. (Columbia Energy). The project will finalize technology maturity and install a system at one of the double-shell tank farms. This paper discusses results of pre-project pilot-scale testing by Columbia Energy and ongoing technology maturation development scope through fiscal year 2012, including planned additional pilot-scale and full-scale simulant testing and operation with actual radioactive tank waste.

  1. UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    and issues related to the waste- management system, including transportation of spent nuclear fuel and high of Energy, including the possible development of a repository for disposing of spent nuclear fuel and high activities related to the possible disposal, packaging, and transportation of the country's spent nuclear

  2. Test Plan for Field Experiments to Support the Immobilized Low-Activity Waste Disposal Performance Assessment at the Hanford Site

    SciTech Connect (OSTI)

    Meyer, Philip D.; McGrail, B. Peter; Bacon, Diana H.

    2001-09-01T23:59:59.000Z

    Much of the data collected to support the Immobilized Low-Activity Waste Performance Assessment (ILAW PA) simulations have been obtained in the laboratory on a relatively small scale (less than 10 cm). In addition, the PA simulations themselves are currently the only means available to integrate the chemical and hydrologic processes involved in the transport of contaminants from the disposal facility into the environment. This report describes the test plan for field experiments to provide data on the hydraulic, transport, and geochemical characteristics of the near-field materials on a more representative (i.e., larger) scale than the laboratory data currently available. The experiments will also provide results that encompass a variety of transport processes likely to occur within the actual disposal facility. These experiments will thus provide the first integrated data on the ILAW facility performance and will provide a crucial dataset to evaluate the simulation-based estimates of overall facility performance used in the PA.

  3. Tank waste remediation system operational scenario

    SciTech Connect (OSTI)

    Johnson, M.E.

    1995-05-01T23:59:59.000Z

    The Tank Waste Remediation System (TWRS) mission is to store, treat, and immobilize highly radioactive Hanford waste (current and future tank waste and the strontium and cesium capsules) in an environmentally sound, safe, and cost-effective manner (DOE 1993). This operational scenario is a description of the facilities that are necessary to remediate the Hanford Site tank wastes. The TWRS Program is developing technologies, conducting engineering analyses, and preparing for design and construction of facilities necessary to remediate the Hanford Site tank wastes. An Environmental Impact Statement (EIS) is being prepared to evaluate proposed actions of the TWRS. This operational scenario is only one of many plausible scenarios that would result from the completion of TWRS technology development, engineering analyses, design and construction activities and the TWRS EIS. This operational scenario will be updated as the development of the TWRS proceeds and will be used as a benchmark by which to evaluate alternative scenarios.

  4. Plasma Mass Filters For Nuclear Waste Reprocessing

    SciTech Connect (OSTI)

    Abraham J. Fetterman and Nathaniel J. Fisch

    2011-05-26T23:59:59.000Z

    Practical disposal of nuclear waste requires high-throughput separation techniques. The most dangerous part of nuclear waste is the fission product, which contains the most active and mobile radioisotopes and produces most of the heat. We suggest that the fission products could be separated as a group from nuclear waste using plasma mass filters. Plasmabased processes are well suited to separating nuclear waste, because mass rather than chemical properties are used for separation. A single plasma stage can replace several stages of chemical separation, producing separate streams of bulk elements, fission products, and actinoids. The plasma mass filters may have lower cost and produce less auxiliary waste than chemical processing plants. Three rotating plasma configurations are considered that act as mass filters: the plasma centrifuge, the Ohkawa filter, and the asymmetric centrifugal trap.

  5. Plasma Mass Filters For Nuclear Waste Reprocessing

    SciTech Connect (OSTI)

    Abraham J. Fetterman and Nathaniel J. Fisch

    2011-05-25T23:59:59.000Z

    Practical disposal of nuclear waste requires high-throughput separation techniques. The most dangerous part of nuclear waste is the fission product, which contains the most active and mobile radioisotopes and produces most of the heat. We suggest that the fission products could be separated as a group from nuclear waste using plasma mass filters. Plasmabased processes are well suited to separating nuclear waste, because mass rather than chemical properties are used for separation. A single plasma stage can replace several stages of chemical separation, producing separate streams of bulk elements, fission products, and actinoids. The plasma mass filters may have lower cost and produce less auxiliary waste than chemical processing plants. Three rotating plasma configurations are considered that act as mass filters: the plasma centrifuge, the Ohkawa filter, and the asymmetric centrifugal trap.

  6. Radioactive Waste Management Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1999-07-09T23:59:59.000Z

    This Manual further describes the requirements and establishes specific responsibilities for implementing DOE O 435.1, Radioactive Waste Management, for the management of DOE high-level waste, transuranic waste, low-level waste, and the radioactive component of mixed waste. The purpose of the Manual is to catalog those procedural requirements and existing practices that ensure that all DOE elements and contractors continue to manage DOE's radioactive waste in a manner that is protective of worker and public health and safety, and the environment. Does not cancel other directives.

  7. SOLUBILITY OF URANIUM AND PLUTONIUM IN ALKALINE SAVANNAH RIVER SITE HIGH LEVEL WASTE SOLUTIONS

    SciTech Connect (OSTI)

    King, W.; Hobbs, D.; Wilmarth, B.; Edwards, T.

    2010-03-10T23:59:59.000Z

    Five actual Savannah River Site tank waste samples and three chemically-modified samples were tested to determine solubility limits for uranium and plutonium over a one year time period. Observed final uranium concentrations ranged from 7 mg U/L to 4.5 g U/L. Final plutonium concentrations ranged from 4 {micro}g Pu/L to 12 mg Pu/L. Actinide carbonate complexation is believed to result in the dramatic solubility increases observed for one sample over long time periods. Clarkeite, NaUO{sub 2}(O)OH {center_dot} H{sub 2}O, was found to be the dominant uranium solid phase in equilibrium with the waste supernate in most cases.

  8. Disposal of high-level nuclear waste above the water table in arid regions

    SciTech Connect (OSTI)

    Roseboom, E.H. Jr.

    1983-12-31T23:59:59.000Z

    Locating a repository in the unsaturated zone of arid regions eliminates or simplifies many of the technological problems involved in designing a repository for operation below the water table and predicting its performance. It also offers possible accessibility and ease of monitoring throughout the operational period and possible retrieval of waste long after. The risks inherent in such a repository appear to be no greater than in one located in the saturated zone; in fact, many aspects of such a repository`s performance will be much easier to predict and the uncertainties will be reduced correspondingly. A major new concern would be whether future climatic changes could produce significant consequences due to possible rise of the water table or increased flux of water through the repository. If spent fuel were used as a waste form, a second new concern would be the rates of escape of gaseous {sup 129}I and {sup 14}C to the atmosphere.

  9. IMPACT OF URANIUM AND THORIUM ON HIGH TIO2 CONCENTRATION NUCLEAR WASTE GLASSES

    SciTech Connect (OSTI)

    Fox, K.; Edwards, T.

    2012-01-11T23:59:59.000Z

    This study focused on the potential 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. MST from the Salt Waste Processing Facility (SWPF) is also considered in the study. The KT08-series of glasses was designed to evaluate any impacts of the inclusion of uranium and thorium in glasses containing the SCIX components. All but one of the study glasses were found to be amorphous by X-ray diffraction (XRD). One of the slowly cooled glasses contained a small amount of trevorite, which is typically found in DWPF-type glasses and had no practical impact on the durability of the glass. The measured Product Consistency Test (PCT) responses for the study glasses and the viscosities of the glasses were well predicted by the current DWPF models. No unexpected issues were encountered when uranium and thorium were added to the glasses with SCIX components.

  10. Midwestern High-Level Radioactive Waste Transportation Project. Highway infrastructure report

    SciTech Connect (OSTI)

    Sattler, L.R.

    1992-02-01T23:59:59.000Z

    In addition to arranging for storage and disposal of radioactive waste, the US Department of Energy (DOE) must develop a safe and efficient transportation system in order to deliver the material that has accumulated at various sites throughout the country. The ability to transport radioactive waste safely has been demonstrated during the past 20 years: DOE has made over 2,000 shipments of spent fuel and other wastes without any fatalities or environmental damage related to the radioactive nature of the cargo. To guarantee the efficiency of the transportation system, DOE must determine the optimal combination of rail transport (which allows greater payloads but requires special facilities) and truck transport Utilizing trucks, in turn, calls for decisions as to when to use legal weight trucks or, if feasible, overweight trucks for fewer but larger shipments. As part of the transportation system, the Facility Interface Capability Assessment (FICA) study contributes to DOE`s development of transportation plans for specific facilities. This study evaluates the ability of different facilities to receive, load and ship the special casks in which radioactive materials will be housed during transport In addition, the DOE`s Near-Site Transportation Infrastructure (NSTI) study (forthcoming) will evaluate the rail, road and barge access to 76 reactor sites from which DOE is obligated to begin accepting spent fuel in 1998. The NSTI study will also assess the existing capabilities of each transportation mode and route, including the potential for upgrade.

  11. Comments on a paper tilted `The sea transport of vitrified high-level radioactive wastes: Unresolved safety issues`

    SciTech Connect (OSTI)

    Sprung, J.L.; McConnell, P.E.; Nigrey, P.J.; Ammerman, D.J. [and others

    1997-05-01T23:59:59.000Z

    The cited paper estimates the consequences that might occur should a purpose-built ship transporting Vitrified High Level Waste (VHLW) be involved in a severe collision that causes the VHLW canisters in one Type-B package to spill onto the floor of a major ocean fishing region. Release of radioactivity from VHLW glass logs, failure of elastomer cask seals, failure of VHLW canisters due to stress corrosion cracking (SCC), and the probabilities of the hypothesized accident scenario, of catastrophic cask failure, and of cask recovery from the sea are all discussed.

  12. Development of integraded mechanistically-based degradation-mode models for performance assessment of high-level waste containers

    SciTech Connect (OSTI)

    Farmer, J. C., LLNL

    1998-06-01T23:59:59.000Z

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological repository at Yucca Mountain, Nevada is a two-tayer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A516 Gr 55 or Monel 400. At the present time, Alloy C- 22 and A516 Gr 55 are favored.

  13. Comprehensive data base of high-level nuclear waste glasses: September 1987 status report: Volume 1, Discussion and glass durability data

    SciTech Connect (OSTI)

    Kindle, C.H.; Kreiter, M.R.

    1987-12-01T23:59:59.000Z

    The Materials Characterization Center (MCC) at Pacific Northwest Laboratory is assembling a comprehensive data base (CDB) of experimental data collected for high-level nuclear waste package components. Data collected throughout the world are included in the data base; current emphasis is on waste glasses and their properties. The goal is to provide a data base of properties and compositions and an analysis of dominant property trends as a function of composition. This data base is a resource that nuclear waste producers, disposers, and regulators can use to compare properties of a particular high-level nuclear waste glass product with the properties of other glasses of similar compositions. Researchers may use the data base to guide experimental tests to fill gaps in the available knowledge or to refine empirical models. The data are incorporated into a computerized data base that will allow the data to be extracted based on, for example, glass composition or test duration. 3 figs.

  14. Characteristics of potential repository wastes

    SciTech Connect (OSTI)

    Notz, K.J.

    1989-01-01T23:59:59.000Z

    The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the spent fuels and other wastes that will be disposed of in a geologic repository. The two major sources of these materials are commercial light-water reactor (LWR) spent fuel and immobilized high-level waste (HLW). Other wastes that may require long-term isolation include non-LWR spent fuels and miscellaneous sources such as activated metals. Detailed characterizations are required for all of these potential repository wastes. These characterizations include physical, chemical, and radiological properties. The latter must take into account decay as a function of time. This information has been extracted from primary data sources, evaluated, and assembled in a Characteristics Data Base which provides data in four formats: hard copy standard reports, menu-driven personal computer (PC) data bases, program-level PC data bases, and mainframe computer files. The Characteristics Data Base provides a standard set of self-consistent data to the various areas of responsibility including systems integration and waste stream analysis, storage, transportation, and geologic disposal. The data will be used for design studies, evaluation of alternatives, and system optimization by OCRWM and supporting contractors. 7 refs., 5 figs., 7 tabs.

  15. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM 2010

    SciTech Connect (OSTI)

    West, B.; Waltz, R.

    2011-06-23T23:59:59.000Z

    Aqueous radioactive wastes from Savannah River Site (SRS) separations and vitrification processes are contained in large underground carbon steel tanks. Inspections made during 2010 to evaluate these vessels and other