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Sample records for level waste hlw

  1. PAIRWISE BLENDING OF HIGH LEVEL WASTE (HLW)

    SciTech Connect (OSTI)

    CERTA, P.J.

    2006-02-22

    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.

  2. HLW Glass Waste Loadings

    Office of Environmental Management (EM)

    HLW Glass Waste Loadings Ian L. Pegg Vitreous State Laboratory The Catholic University of ... (JHCM) technology Factors affecting waste loadings Waste loading requirements ...

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

    Broader source: Energy.gov [DOE]

    GC-52 provides legal advice to DOE regarding the long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF). SNF is nuclear fuel that has been used as fuel in a reactor...

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

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

    HLW is the highly radioactive material resulting from the reprocessing of SNF. Under the Nuclear Waste Policy Act of 1982, the federal government is responsible for the disposal of ...

  5. Collaboration, Automation, and Information Management at Hanford High Level Radioactive Waste (HLW) Tank Farms

    SciTech Connect (OSTI)

    Aurah, Mirwaise Y.; Roberts, Mark A.

    2013-12-12

    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.

  6. 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

    2010-01-04

    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

  7. U.S. Department of Energy (DOE) initiated performance enhancements to the Hanford waste treatment and immobilization plant (WTP) high-level waste vitrification (HLW) system

    SciTech Connect (OSTI)

    Bowan, Bradley [Energy Solutions, LLC (United States); Gerdes, Kurt [United States Department of Energy (United States); Pegg, Ian [Vitreous State Laboratory, Catholic University of America, 400 Hannan Hall 620 Michigan Avenue, NE Washington, DC 20064 (United States); Holton, Langdon [Pacific Northwest National Laboratory, PO Box 999, Richland WA 99352 (United States)

    2007-07-01

    Available in abstract form only. Full text of publication follows: The U.S Department of Energy is currently constructing, at the Hanford, Washington Site, a Waste Treatment and Immobilization Plant (WTP) for the treatment and immobilization, by vitrification, of stored underground tank wastes. The WTP is comprised of four major facilities: a Pretreatment facility to separate the tank waste into high level waste (HLW) and low activity waste (LAW); a HLW vitrification facility to immobilize the HLW fraction; a LAW vitrification facility to immobilize the LAW fraction and an analytical Laboratory to support the treatment facilities. DOE has strategic objectives to optimize the performance of the WTP facilities, and waste forms, in order to reduce the overall schedule and cost for the treatment of the Hanford tank wastes. One key part of this strategy is to maximize the loading of inorganic waste components in the final glass product (waste loading). For the Hanford tank wastes, this is challenging because of the compositional diversity of the wastes generated over several decades. This paper presents the results of an initial series of HLW waste loading enhancement tests, using diverse HLW compositions that are projected for treatment at the WTP. Specifically, results of glass formulation development and melter testing with simulated Hanford HLW containing high concentrations of troublesome components such as bismuth, aluminum, aluminum-sodium, and chromium will be presented. (authors)

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

    SciTech Connect (OSTI)

    Fox, K. M.

    2014-02-27

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

  9. Waste Treatment and Immobilization Plant HLW Waste Vitrification Facility |

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

    Department of Energy HLW Waste Vitrification Facility Waste Treatment and Immobilization Plant HLW Waste Vitrification Facility Full Document and Summary Versions are available for download Waste Treatment and Immobilization Plant HLW Waste Vitrification Facility (742.54 KB) Summary - WTP HLW Waste Vitrification Facility (137.99 KB) More Documents & Publications Waste Treatment and Immobilization Plant (WTP) Analytical Laboratory (LAB), Balance of Facilities (BOF) and Low-Activity Waste

  10. Waste Treatment and Immobilation Plant HLW Waste Vitrification...

    Office of Environmental Management (EM)

    6 Technology Readiness Assessment for the Waste Treatment and Immobilization Plant (WTP) HLW Waste Vitrification Facility L. Holton D. Alexander C. Babel H. Sutter J. Young August ...

  11. 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

    2007-09-17

    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.

  12. 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

    2007-03-30

    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

  13. Summary - WTP HLW Waste Vitrification Facility

    Office of Environmental Management (EM)

    W HLW W DOE is Immob site's t facilitie Facility to iden the HL to be i norma The as along w Level ( * H * H * H Sy * Pu D The Ele Site: H roject: W Report Date: M ited States ...

  14. Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility

    Office of Environmental Management (EM)

    6 Technology Readiness Assessment for the Waste Treatment and Immobilization Plant (WTP) HLW Waste Vitrification Facility L. Holton D. Alexander C. Babel H. Sutter J. Young August 2007 Prepared by the U.S. Department of Energy Office of River Protection Richland, Washington, 99352 07-DESIGN-046 Technology Readiness Assessment for the Waste Treatment and Immobilization Plant (WTP) HLW Waste Vitrification Facility L. Holton D. Alexander C. Babel H. Sutter J. Young August 2007 Prepared by the U.S.

  15. DEVELOPMENT OF GLASS MATRICES FOR HLW RADIOACTIVE WASTES

    SciTech Connect (OSTI)

    Jantzen, C.

    2010-03-18

    Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either borosilicate glass or phosphate glass. One of the primary reasons that glass has become the most widely used immobilization media is the relative simplicity of the vitrification process, e.g. melt waste plus glass forming frit additives and cast. A second reason that glass has become widely used for HLW is that the short range order (SRO) and medium range order (MRO) found in glass atomistically bonds the radionuclides and governs the melt properties such as viscosity, resistivity, sulphate solubility. The molecular structure of glass controls contaminant/radionuclide release by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. The molecular structure is flexible and hence accounts for the flexibility of glass formulations to waste variability. Nuclear waste glasses melt between 1050-1150 C which minimizes the volatility of radioactive components such as Tc{sup 99}, Cs{sup 137}, and I{sup 129}. Nuclear waste glasses have good long term stability including irradiation resistance. Process control models based on the molecular structure of glass have been mechanistically derived and have been demonstrated to be accurate enough to control the world's largest HLW Joule heated ceramic melter in the US since 1996 at 95% confidence.

  16. Amended Record of Decision for the Idaho High-Level Waste (HLW) and Facilities Disposition Final Environmental Impact Statement

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy (DOE) is amending its Record of Decision (ROD) published December 19, 2005 (70 Federal Register (FR) 75165), pursuant to the Idaho HIgh-Level Waste and Facilities...

  17. Immobilized High Level Waste (HLW) Interim Storage Alternative Generation and analysis and Decision Report 2nd Generation Implementing Architecture

    SciTech Connect (OSTI)

    CALMUS, R.B.

    2000-09-14

    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.

  18. 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

    2009-11-05

    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.

  19. The NUMO Strategy for HLW and TRU Waste Disposal

    SciTech Connect (OSTI)

    Kitayama, K.; Oda, Y. [Nuclear Waste Management Organization of Japan (NUMO), Tokyo (Japan)

    2008-07-01

    Shortly after the Nuclear Waste Management Organization of Japan (NUMO) was established, we initiated an open call to all municipalities, requesting volunteers to host a repository for vitrified HLW. The first volunteer applied for a preliminary literature survey phase last January but, unfortunately, it withdrew the application in April. This failure provided an invaluable lesson for both the relevant authorities and NUMO; subsequently the Atomic Energy Commission of Japan and associated organizations are examining a support plan to back up NUMO's open solicitation. On another front, a recent amendment of 'The Specified Radioactive Waste Final Disposal Act' also allocates specific 'TRU' waste for deep geological disposal, requiring a demonstration of safety to a similar level as that for HLW. To implement the radioactive waste disposal project, NUMO has developed a methodology appropriate to our specific boundary conditions - the NUMO Structured Approach. This takes into account, in particular, our need to balance competing goals, such as operational safety, post-closure safety and cost, during repository tailoring to specific sites. The most important challenge for NUMO is, however, to attract volunteers. We believe that our open and structured R and D program is critical to demonstrate technical competence which, in turn, enhances the credibility of our various public relations activities. (authors)

  20. FLUIDIZED BED STEAM REFORMING (FBSR) OF HIGH LEVEL WASTE (HLW) ORGANIC AND NITRATE DESTRUCTION PRIOR TO VITRIFICATION: CRUCIBLE SCALE TO ENGINEERING SCALE DEMONSTRATIONS AND NON-RADIOACTIVE TO RADIOACTIVE DEMONSTRATIONS

    SciTech Connect (OSTI)

    Jantzen, C; Michael Williams, M; Gene Daniel, G; Paul Burket, P; Charles Crawford, C

    2009-02-07

    Over a decade ago, an in-tank precipitation process to remove Cs-137 from radioactive high level waste (HLW) supernates was demonstrated at the Savannah River Site (SRS). The full scale demonstration with actual HLW was performed in SRS Tank 48 (T48). Sodium tetraphenylborate (NaTPB) was added to enable Cs-137 extraction as CsTPB. The CsTPB, an organic, and its decomposition products proved to be problematic for subsequent processing of the Cs-137 precipitate in the SRS HLW vitrification facility for ultimate disposal in a HLW repository. Fluidized Bed Steam Reforming (FBSR) is being considered as a technology for destroying the organics and nitrates in the T48 waste to render it compatible with subsequent HLW vitrification. During FBSR processing the T48 waste is converted into organic-free and nitrate-free carbonate-based minerals which are water soluble. The soluble nature of the carbonate-based minerals allows them to be dissolved and pumped to the vitrification facility or returned to the tank farm for future vitrification. The initial use of the FBSR process for T48 waste was demonstrated with simulated waste in 2003 at the Savannah River National Laboratory (SRNL) using a specially designed sealed crucible test that reproduces the FBSR pyrolysis reactions, i.e. carbonate formation, organic and nitrate destruction. This was followed by pilot scale testing of simulants at the Science Applications International Corporation (SAIC) Science & Technology Application Research (STAR) Center in Idaho Falls, ID by Idaho National Laboratory (INL) and SRNL in 2003-4 and then engineering scale demonstrations by THOR{reg_sign} Treatment Technologies (TTT) and SRS/SRNL at the Hazen Research, Inc. (HRI) test facility in Golden, CO in 2006 and 2008. Radioactive sealed crucible testing with real T48 waste was performed at SRNL in 2008, and radioactive Benchscale Steam Reformer (BSR) testing was performed in the SRNL Shielded Cell Facility (SCF) in 2008.

  1. Idaho High-Level Waste & Facilities Disposition, Final Environmental...

    Office of Environmental Management (EM)

    Copies of the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact ... of alternatives for managing high- level waste (HLW) calcine, mixed transuranic waste...

  2. Cement encapsulation of low-level waste liquids. Final report

    SciTech Connect (OSTI)

    Baker, M.N.; Houston, H.M.

    1999-01-01

    Pretreatment of liquid high-level radioactive waste at the West Valley Demonstration Project (WVDP) was essential to ensuring the success of high-level waste (HLW) vitrification. By chemically separating the HLW from liquid waste, it was possible to achieve a significant reduction in the volume of HLW to be vitrified. In addition, pretreatment made it possible to remove sulfates, which posed several processing problems, from the HLW before vitrification took place.

  3. Idaho High-Level Waste & Facilities Disposition, Final Environmental...

    Office of Environmental Management (EM)

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

  4. High Level Waste System Plan Revision 9

    SciTech Connect (OSTI)

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

    1998-04-01

    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.

  5. 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

    2009-12-30

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

  6. 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...

  7. Optimizing High Level Waste Disposal

    SciTech Connect (OSTI)

    Dirk Gombert

    2005-09-01

    If society is ever to reap the potential benefits of nuclear energy, technologists must close the fuel-cycle completely. A closed cycle equates to a continued supply of fuel and safe reactors, but also reliable and comprehensive closure of waste issues. High level waste (HLW) disposal in borosilicate glass (BSG) is based on 1970s era evaluations. This host matrix is very adaptable to sequestering a wide variety of radionuclides found in raffinates from spent fuel reprocessing. However, it is now known that the current system is far from optimal for disposal of the diverse HLW streams, and proven alternatives are available to reduce costs by billions of dollars. The basis for HLW disposal should be reassessed to consider extensive waste form and process technology research and development efforts, which have been conducted by the United States Department of Energy (USDOE), international agencies and the private sector. Matching the waste form to the waste chemistry and using currently available technology could increase the waste content in waste forms to 50% or more and double processing rates. Optimization of the HLW disposal system would accelerate HLW disposition and increase repository capacity. This does not necessarily require developing new waste forms, the emphasis should be on qualifying existing matrices to demonstrate protection equal to or better than the baseline glass performance. Also, this proposed effort does not necessarily require developing new technology concepts. The emphasis is on demonstrating existing technology that is clearly better (reliability, productivity, cost) than current technology, and justifying its use in future facilities or retrofitted facilities. Higher waste processing and disposal efficiency can be realized by performing the engineering analyses and trade-studies necessary to select the most efficient methods for processing the full spectrum of wastes across the nuclear complex. This paper will describe technologies being

  8. Immobilization and Waste Form Product Acceptance for Low Level and TRU Waste Forms

    SciTech Connect (OSTI)

    Holtzscheiter, E.W. [Westinghouse Savannah River Company, AIKEN, SC (United States); Harbour, J.R.

    1998-05-01

    The Tanks Focus Area is supporting technology development in immobilization of both High Level (HLW) and Low Level (LLW) radioactive wastes. The HLW process development at Hanford and Idaho is patterned closely after that of the Savannah River (Defense Waste Processing Facility) and West Valley Sites (West Valley Demonstration Project). However, the development and options open to addressing Low Level Waste are diverse and often site specific. To start, it is important to understand the breadth of Low Level Wastes categories.

  9. Enterprise Assessments Review of the Hanford Site Waste Treatment...

    Office of Environmental Management (EM)

    Hanford Site Waste Treatment and Immobilization Plant Project Engineering Processes ... Design Description HLW High-Level Waste HMH HLW Melter Handling ITS Important to ...

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

    Office of Environmental Management (EM)

    This EIS also analyzes alternatives for the final disposition of HLW management facilities at the INEEL after their missions are completed. Idaho High-Level Waste and Facilities ...

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

    Broader source: Energy.gov [DOE]

    This report summarizes efforts to simulate coupled thermal-hydrological-chemical (THC) processes occurring within a generic hypothetical high-level waste (HLW) repository in bedded salt.

  12. Widening the envelope of UK HLW vitrification - Experimental studies with high waste loadings and new product formulations on a full scale non-active vitrification plant

    SciTech Connect (OSTI)

    Short, R.; Gribble, N. [Nexia Solutions, Sellafield, Cumbria, CA20 1PG (United Kingdom); Riley, A. [Sellafield Ltd, Sellafield, Seascale, Cumbria, CA20 1PG, UK (United Kingdom)

    2008-07-01

    The Vitrification Test Rig is a full scale waste vitrification plant that processes non-radioactive liquid HLW simulants based on the active waste streams produced by the reprocessing plants in the UK. Previous work on the rig has primarily concerned increasing the operational envelopes for the active waste vitrification plants at Sellafield to accommodate higher throughputs of Blended waste streams, higher waste oxide incorporation rates in the vitrified products, and the incorporation of legacy waste streams from early reactor commissioning and reprocessing operations at Sellafield. Recent operations have focussed on four main areas; dilute liquid feeds, very high Magnox waste stream incorporation levels, alternative base glass formulations and providing an operational envelope for 28 %w/w Magnox waste vitrification. This paper details the work performed and the major findings of that work. In summary: The VTR has been successfully used to determine operational envelopes and product quality for several HLW feed variations that will allow WVP to increase overall plant throughput via increased waste loading in canisters, increased HLW feed rates or a combination of both. The VTR has also demonstrated the ability to go to waste incorporations, feed rates and glass compositions that are currently beyond WVP specified limits, but that are feasible for future vitrification regimes. In addition, the VTR has trialled dilute feeds similar to those that are likely to be received by WVP in the future and the data obtained from these experiments will allow WVP to prepare adequately for the high throughput challenge of such feeds. Furthermore, new equipment has been trialled on the VTR in water feed mode to determine its suitability and operational limitations for WVP. Future operations will, in the short term, be concerned with increasing the throughput of WVP and are likely to focus on HLW decommissioning operations waste streams in the longer term. (authors)

  13. High-Level Waste Melter Review

    SciTech Connect (OSTI)

    Ahearne, J.; Gentilucci, J.; Pye, L. D.; Weber, T.; Woolley, F.; Machara, N. P.; Gerdes, K.; Cooley, C.

    2002-02-26

    The U.S. Department of Energy (DOE) is faced with a massive cleanup task in resolving the legacy of environmental problems from years of manufacturing nuclear weapons. One of the major activities within this task is the treatment and disposal of the extremely large amount of high-level radioactive (HLW) waste stored at the Hanford Site in Richland, Washington. The current planning for the method of choice for accomplishing this task is to vitrify (glassify) this waste for disposal in a geologic repository. This paper describes the results of the DOE-chartered independent review of alternatives for solidification of Hanford HLW that could achieve major cost reductions with reasonable long-term risks, including recommendations on a path forward for advanced melter and waste form material research and development. The potential for improved cost performance was considered to depend largely on increased waste loading (fewer high-level waste canisters for disposal), higher throughput, or decreased vitrification facility size.

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

    SciTech Connect (OSTI)

    Kruger, A. A.; Pegg, Ian L.; Gan, Hao; Kot, Wing K.

    2012-12-13

    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.

  15. EIS-0287: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, EIS-0287 (September 2002)

    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...

  16. Defense High Level Waste Disposal Container System Description

    SciTech Connect (OSTI)

    2000-10-12

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different materials

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

    SciTech Connect (OSTI)

    Marra, James; Kim, Dong -Sang; Maio, Vincent

    2015-09-02

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

  18. Permitting plan for the high-level waste interim storage

    SciTech Connect (OSTI)

    Deffenbaugh, M.L.

    1997-04-23

    This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. 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 of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist.

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

    SciTech Connect (OSTI)

    Burgard, K.C.

    1998-06-02

    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.

  20. FLUIDIZED BED STEAM REFORMING ENABLING ORGANIC HIGH LEVEL WASTE DISPOSAL

    SciTech Connect (OSTI)

    Williams, M

    2008-05-09

    Waste streams planned for generation by the Global Nuclear Energy Partnership (GNEP) and existing radioactive High Level Waste (HLW) streams containing organic compounds such as the Tank 48H waste stream at Savannah River Site have completed simulant and radioactive testing, respectfully, by Savannah River National Laboratory (SRNL). GNEP waste streams will include up to 53 wt% organic compounds and nitrates up to 56 wt%. Decomposition of high nitrate streams requires reducing conditions, e.g. provided by organic additives such as sugar or coal, to reduce NOX in the off-gas to N2 to meet Clean Air Act (CAA) standards during processing. Thus, organics will be present during the waste form stabilization process regardless of the GNEP processes utilized and exists in some of the high level radioactive waste tanks at Savannah River Site and Hanford Tank Farms, e.g. organics in the feed or organics used for nitrate destruction. Waste streams containing high organic concentrations cannot be stabilized with the existing HLW Best Developed Available Technology (BDAT) which is HLW vitrification (HLVIT) unless the organics are removed by pretreatment. The alternative waste stabilization pretreatment process of Fluidized Bed Steam Reforming (FBSR) operates at moderate temperatures (650-750 C) compared to vitrification (1150-1300 C). The FBSR process has been demonstrated on GNEP simulated waste and radioactive waste containing high organics from Tank 48H to convert organics to CAA compliant gases, create no secondary liquid waste streams and create a stable mineral waste form.

  1. Management of nuclear materials and non-HLW | Department of Energy

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

    Management of nuclear materials and non-HLW Management of nuclear materials and non-HLW GC-52 provides legal advice to DOE regarding the consolidation and disposition of nuclear materials, including plutonium, uranium, and nuclear waste in accordance with applicable statutes, DOE Orders and international commitments. Advice encompasses issues related to mixed oxide fuel, waste incidental-to-reprocessing, transuranic waste, low-level waste, greater-than-class C waste and sealed sources.

  2. HLW-OVP-94-00n High Level Waste Management Division HLW System...

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

    ... during a three day period without any significant problems. ... in the area of System Integration and production planning. ... 11) Maintain adequate emergency tank space per the Tank ...

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

    SciTech Connect (OSTI)

    SK Sundaram; ML Elliott; D Bickford

    1999-11-19

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

  4. EIS-0287: Notice of Preferred Sodium Bearing Waste Treatment...

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

    Preferred Sodium Bearing Waste Treatment Technology EIS-0287: Notice of Preferred Sodium Bearing Waste Treatment Technology Idaho High-Level Waste (HLW) and Facilities Disposition...

  5. Tank Closure and Waste Management Environmental Impact Statement...

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

    Secondary-waste-form performance High-level radioactive waste (HLW) disposition (Yucca Mountain issue) Mitigation Exclusion of greater-than-Class C (GTCC) waste in cumulative ...

  6. Towards increased waste loading in high level waste glasses: Developing a better understanding of crystallization behavior

    SciTech Connect (OSTI)

    Marra, James C.; Kim, Dong -Sang

    2014-12-18

    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. Thus, 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

    2014-12-18

    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. Thus, 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.more » 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.« less

  8. THERMAL ANALYSIS OF GEOLOGIC HIGH-LEVEL RADIOACTIVE WASTE PACKAGES

    SciTech Connect (OSTI)

    Hensel, S.; Lee, S.

    2010-04-20

    The engineering design of disposal of the high level waste (HLW) packages in a geologic repository requires a thermal analysis to provide the temperature history of the packages. Calculated temperatures are used to demonstrate compliance with criteria for waste acceptance into the geologic disposal gallery system and as input to assess the transient thermal characteristics of the vitrified HLW Package. The objective of the work was to evaluate the thermal performance of the supercontainer containing the vitrified HLW in a non-backfilled and unventilated underground disposal gallery. In order to achieve the objective, transient computational models for a geologic vitrified HLW package were developed by using a computational fluid dynamics method, and calculations for the HLW disposal gallery of the current Belgian geological repository reference design were performed. An initial two-dimensional model was used to conduct some parametric sensitivity studies to better understand the geologic system's thermal response. The effect of heat decay, number of co-disposed supercontainers, domain size, humidity, thermal conductivity and thermal emissivity were studied. Later, a more accurate three-dimensional model was developed by considering the conduction-convection cooling mechanism coupled with radiation, and the effect of the number of supercontainers (3, 4 and 8) was studied in more detail, as well as a bounding case with zero heat flux at both ends. The modeling methodology and results of the sensitivity studies will be presented.

  9. 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

    2014-12-18

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHM) 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 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 glasses and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulationsmorehave 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 composition represents a waste group that is waste loading limited primarily due to high concentration of Fe2O3. 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 group.less

  10. Towards increased waste loading in high level waste glasses: developing a better understanding of crystallization behavior

    SciTech Connect (OSTI)

    Marra, James C.; Kim, Dong-Sang

    2014-12-18

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHM) 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 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 glasses 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 composition represents a waste group that is waste loading limited primarily due to high concentration of Fe2O3. 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 group.

  11. New Design for an HLW Repository (for Spent Fuel and Waste from Reprocessing) in a Salt Formation in Germany - 12213

    SciTech Connect (OSTI)

    Bollingerfehr, Wilhelm; Filbert, Wolfgang; Lerch, Christian; Mueller-Hoeppe, Nina; Charlier, Frank

    2012-07-01

    In autumn 2010, after a 10-year moratorium, exploration was resumed in Gorleben, the potential site for a German HLW repository. At the same time, the Federal Government launched a two-year preliminary safety analysis to assess whether the salt dome at Gorleben is suitable to host all heat-generating radioactive waste generated by German NPPs based on the waste amounts expected at that time. The revised Atomic Energy Act of June 2011 now stipulates a gradual phase-out of nuclear energy production by 2022, which is 13 years earlier than expected in 2010. A repository design was developed which took into account an updated set of data on the amounts and types of expected heat-generating waste, the documented results of the exploration of the Gorleben salt dome, and the new 'Safety Requirements Governing the Final Disposal of Heat-Generating Radioactive Waste' of 30 September, 2010. The latter has a strong influence on the conceptual designs as it requires that retrievability of all waste containers is possible within the repository lifetime. One design considered that all waste containers will be disposed of in horizontal drifts of a geologic repository, while the other design considered that all waste containers will be disposed of in deep vertical boreholes. For both options (emplacement in drifts/emplacement in vertical boreholes), the respective design includes a selection of waste containers, the layout of drifts, respectively lined boreholes, a description of emplacement fields, and backfilling and sealing measures. The design results were described and displayed and the differences between the two main concepts were elaborated and discussed. For the first time in both repository designs the requirement was implemented to retrieve waste canisters during the operational phase. The measures to fulfill this requirement and eventually the consequences were highlighted. It was pointed out that there arises the need to keep transport- and storage casks in adequate

  12. High-level waste management technology program plan

    SciTech Connect (OSTI)

    Harmon, H.D.

    1995-01-01

    The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs.

  13. Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    The Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel report assesses the technical options for the safe and permanent disposal of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) managed by the Department of Energy. Specifically, it considers whether DOE-managed HLW and SNF should be disposed of with commercial SNF and HLW in one geologic repository or whether there are advantages to developing separate geologic disposal pathways for some DOE-managed HLW and SNF. The report recommends that the Department begin implementation of a phased, adaptive, and consent-based strategy with development of a separate mined repository for some DOE-managed HLW and cooler DOE-managed SNF.

  14. A pilot test of partitioning for the simulated highly saline high level waste

    SciTech Connect (OSTI)

    Chen, Jing; Wang, Jianchen; Jing, Shan

    2007-07-01

    It is a problem how to treat the highly saline high level waste (HLW). A partitioning process for HLW was developed at INET. The partitioning process includes the removal of actinides by TRPO extraction, the removal of Sr by crown ether extraction, and the removal of Cs by ion exchange. A 72-hour test was carried out in a pilot facility using the simulated HLW. Nd and Zr were used to simulate Am and Pu, respectively. The decontamination factors are >3000, >500, >1000, {approx}150 and {approx}94 for U, Nd, Zr, Sr and Cs, respectively. The results meet the requirement to change the highly saline HLW into a non-{alpha} and intermediate level waste. (authors)

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

    SciTech Connect (OSTI)

    Christian, J. H.

    2015-08-18

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

  16. Classroom simulation of public involvement in H.L.W. issues featuring STS concepts

    SciTech Connect (OSTI)

    Bieniawski, Z.T.

    1996-12-01

    This paper reports on a classroom experiment, now in its third year, conducted by the author for students involvement in High-Level Waste disposal. The project involved designing a methodology for HLW management in the United States.

  17. HIGH ALUMINUM HLW GLASSES FOR HANFORDS WTP

    SciTech Connect (OSTI)

    KRUGER AA; JOSEPH I; BOWMAN BW; GAN H; KOT W; MATLACK KS; PEGG IL

    2009-08-19

    The world's largest radioactive waste vitrification facility is now under construction at the United State Department of Energy's (DOE's) Hanford site. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is designed to treat nearly 53 million gallons of mixed hazardous and radioactive waste now residing in 177 underground storage tanks. This multi-decade processing campaign will be one of the most complex ever undertaken because of the wide chemical and physical variability of the waste compositions generated during the cold war era that are stored at Hanford. The DOE Office of River Protection (ORP) has initiated a program to improve the long-term operating efficiency of the WTP vitrification plants with the objective of reducing the overall cost of tank waste treatment and disposal and shortening the duration of plant operations. Due to the size, complexity and duration of the WTP mission, the lifecycle operating and waste disposal costs are substantial. As a result, gains in High Level Waste (HLW) and Low Activity Waste (LAW) waste loadings, as well as increases in glass production rate, which can reduce mission duration and glass volumes for disposal, can yield substantial overall cost savings. EnergySolutions and its long-term research partner, the Vitreous State Laboratory (VSL) of the Catholic University of America, have been involved in a multi-year ORP program directed at optimizing various aspects of the HLW and LAW vitrification flow sheets. A number of Hanford HLW streams contain high concentrations of aluminum, which is challenging with respect to both waste loading and processing rate. Therefore, a key focus area of the ORP vitrification process optimization program at EnergySolutions and VSL has been development of HLW glass compositions that can accommodate high Al{sub 2}O{sub 3} concentrations while maintaining high processing rates in the Joule Heated Ceramic Melters (JHCMs) used for waste vitrification at the WTP. This paper, reviews

  18. Process Design Concepts for Stabilization of High Level Waste Calcine

    SciTech Connect (OSTI)

    T. R. Thomas; A. K. Herbst

    2005-06-01

    The current baseline assumption is that packaging ¡§as is¡¨ and direct disposal of high level waste (HLW) calcine in a Monitored Geologic Repository will be allowed. The fall back position is to develop a stabilized waste form for the HLW calcine, that will meet repository waste acceptance criteria currently in place, in case regulatory initiatives are unsuccessful. A decision between direct disposal or a stabilization alternative is anticipated by June 2006. The purposes of this Engineering Design File (EDF) are to provide a pre-conceptual design on three low temperature processes under development for stabilization of high level waste calcine (i.e., the grout, hydroceramic grout, and iron phosphate ceramic processes) and to support a down selection among the three candidates. The key assumptions for the pre-conceptual design assessment are that a) a waste treatment plant would operate over eight years for 200 days a year, b) a design processing rate of 3.67 m3/day or 4670 kg/day of HLW calcine would be needed, and c) the performance of waste form would remove the HLW calcine from the hazardous waste category, and d) the waste form loadings would range from about 21-25 wt% calcine. The conclusions of this EDF study are that: (a) To date, the grout formulation appears to be the best candidate stabilizer among the three being tested for HLW calcine and appears to be the easiest to mix, pour, and cure. (b) Only minor differences would exist between the process steps of the grout and hydroceramic grout stabilization processes. If temperature control of the mixer at about 80„aC is required, it would add a major level of complexity to the iron phosphate stabilization process. (c) It is too early in the development program to determine which stabilizer will produce the minimum amount of stabilized waste form for the entire HLW inventory, but the volume is assumed to be within the range of 12,250 to 14,470 m3. (d) The stacked vessel height of the hot process vessels

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

    SciTech Connect (OSTI)

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

    1999-01-01

    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.

  20. Report on Separate Disposal of Defense High- Level Radioactive...

    Office of Environmental Management (EM)

    Radioactive Waste March 2015 This page left blank. i EXECUTIVE SUMMARY Purpose This report considers whether a separate repository for high-level radioactive waste (HLW) ...

  1. Extended Development Work to Validate a HLW Calcine Waste Form via INL's Cold Crucible Induction Melter

    SciTech Connect (OSTI)

    James A. King; Vince Maio

    2011-09-01

    To accomplish calcine treatment objectives, the Idaho Clean-up Project contractor, CWI, has chosen to immobilize the calcine in a glass-ceramic via the use of a Hot-Isostatic-Press (HIP); a treatment selection formally documented in a 2010 Record of Decision (ROD). Even though the HIP process may prove suitable for the calcine as specified in the ROD and validated in a number of past value engineering sessions, DOE is evaluating back-up treatment methods for the calcine as a result of the technical, schedule, and cost risk associated with the HIPing process. Consequently DOE HQ has requested DOE ID to make INL's bench-scale cold-crucible induction melter (CCIM) available for investigating its viability as a process alternate to calcine treatment. The waste form is the key component of immobilization of radioactive waste. Providing a solid, stable, and durable material that can be easily be stored is the rationale for immobilization of radioactive waste material in glass, ceramic, or glass-ceramics. Ceramic waste forms offer an alternative to traditional borosilicate glass waste forms. Ceramics can usually accommodate higher waste loadings than borosilicate glass, leading to smaller intermediate and long-term storage facilities. Many ceramic phases are known to possess superior chemical durability as compared to borosilicate glass. However, ceramics are generally multiphase systems containing many minor phase that make characterization and prediction of performance within a repository challenging. Additionally, the technologies employed in ceramic manufacture are typically more complex and expensive. Thus, many have proposed using glass-ceramics as compromise between in the more inexpensive, easier to characterize glass waste forms and the more durable ceramic waste forms. Glass-ceramics have several advantages over traditional borosilicate glasses as a waste form. Borosilicate glasses can inadvertently devitrify, leading to a less durable product that could crack

  2. 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

    2003-02-27

    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.

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

    SciTech Connect (OSTI)

    Vance, Eric

    2007-07-01

    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)

  4. High-Level Waste Inventory

    Office of Environmental Management (EM)

    Analysis of Alternatives for Disposition of the Idaho Calcined High-Level Waste Inventory ... of the Idaho Calcined High-Level Waste Inventory Volume 1- Summary Report April ...

  5. HLW Canister and Can-In-Canister Drop Calculation

    SciTech Connect (OSTI)

    H. Marr

    1999-09-15

    The purpose of this calculation is to evaluate the structural response of the standard high-level waste (HLW) canister and the HLW canister containing the cans of immobilized plutonium (''can-in-canister'' throughout this document) to the drop event during the handling operation. The objective of the calculation is to provide the structure parameter information to support the canister design and the waste handling facility design. Finite element solution is performed using the commercially available ANSYS Version (V) 5.4 finite element code. Two-dimensional (2-D) axisymmetric and three-dimensional (3-D) finite element representations for the standard HLW canister and the can-in-canister are developed and analyzed using the dynamic solver.

  6. Advanced waste form and Melter development for treatment of troublesome high-level wastes

    SciTech Connect (OSTI)

    Marra, James; Kim, Dong -Sang; Maio, Vincent

    2015-10-01

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (also with high Al2O3 concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both

  7. Advanced High-Level Waste Glass Research and Development Plan

    SciTech Connect (OSTI)

    Peeler, David K.; Vienna, John D.; Schweiger, Michael J.; Fox, Kevin M.

    2015-07-01

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

  8. TWRS HLW interim storage facility search and evaluation

    SciTech Connect (OSTI)

    Calmus, R.B., Westinghouse Hanford

    1996-05-16

    The purpose of this study was to identify and provide an evaluation of interim storage facilities and potential facility locations for the vitrified high-level waste (HLW) from the Phase I demonstration plant and Phase II production plant. In addition, interim storage facilities for solidified separated radionuclides (Cesium and Technetium) generated during pretreatment of Phase I Low-Level Waste Vitrification Plant feed was evaluated.

  9. Stability of High-Level Radioactive Waste Forms

    SciTech Connect (OSTI)

    Besmann, T.M.

    2001-06-22

    High-level waste (HLW) glass compositions, processing schemes, limits on waste content, and corrosion/dissolution release models are dependent on an accurate knowledge of melting temperatures and thermochemical values. Unfortunately, existing models for predicting these temperatures are empirically-based, depending on extrapolations of experimental information. In addition, present models of leaching behavior of glass waste forms use simplistic assumptions or experimentally measured values obtained under non-realistic conditions. There is thus a critical need for both more accurate and more widely applicable models for HLW glass behavior, which this project addressed. Significant progress was made in this project on modeling HLW glass. Borosilicate glass was accurately represented along with the additional important components that contain iron, lithium, potassium, magnesium, and calcium. The formation of crystalline inclusions in the glass, an issue in Hanford HLW formulations, was modeled and shown to be predictive. Thus the results of this work have already demonstrated practical benefits with the ability to map compositional regions where crystalline material forms, and therefore avoid that detrimental effect. With regard to a fundamental understanding, added insights on the behavior of the components of glass have been obtained, including the potential formation of molecular clusters. The EMSP project had very significant effects beyond the confines of Environmental Management. The models developed for glass have been used to solve a very costly problem in the corrosion of refractories for glass production. The effort resulted in another laboratory, Sandia National Laboratories-Livermore, to become conversant in the techniques and to apply those through a DOE Office of Industrial Technologies project joint with PPG Industries. The glass industry as a whole is now cognizant of these capabilities, and there is a Glass Manufacturer's Research Institute proposal

  10. 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.

    2009-10-05

    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.

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

    SciTech Connect (OSTI)

    Christian, J. H.

    2015-08-01

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

  12. In-tank pretreatment of high-level tank wastes: The SIPS system

    SciTech Connect (OSTI)

    Reich, M.; Powell, J.; Barletta, R.

    1996-03-01

    A new approach, termed SIPS (Small In-Tank Processing System), that enables the in-tank processing and separation of high-level tank wastes into high-level waste (HLW) and low-level waste (LLW) streams that are suitable for vitrification, is described. Presently proposed pretreatment systems, such as enhanced sludge washing (ESW) and TRUEX, require that the high-level tank wastes be retrieved and pumped to a large, centralized processing facility, where the various waste components are separated into a relatively small, radioactively concentrated stream (HLW), and a relatively large, predominantly non-radioactive stream (LLW). In SIPS, a small process module, typically on the order of 1 meter in diameter and 4 meters in length, is inserted into a tank. During a period of approximately six months, it processes the solid/liquid materials in the tank, separating them into liquid HLW and liquid LLW output streams that are pumped away in two small diameter (typically 3 cm o.d.) pipes. The SIPS concept appears attractive for pretreating high level wastes, since it would: (1) process waste in-situ in the tanks, (2) be cheaper and more reliable than a larger centralized facility, (3) be quickly demonstrable at full scale, (4) have less technical risk, (5) avoid having to transfer unstable slurries for long distances, and (6) be simple to decommission and dispose of. Further investigation of the SIPS concept appears desirable, including experimental testing and development of subscale demonstration units.

  13. Characteristics of potential repository wastes. Volume 3, Appendix 3A, ORIGEN2 decay tables for immobilized high-level waste; Appendix 3B, Interim high-level waste forms

    SciTech Connect (OSTI)

    Not Available

    1992-07-01

    This appendix presents the results of decay calculations using the ORIGEN2 code to determine the radiological properties of canisters of immobilized high-level waste as a function of decay time for decay times up to one million years. These calculations were made for the four HLW sites (West Valley Demonstration Project, Savannah River Site, Hanford Site, and Idaho National Engineering Laboratory) using the composition data discussed in the HLW section of this report. Calculated ({alpha},n) neutron production rates are also shown.

  14. 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.

    1996-02-01

    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.

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

    SciTech Connect (OSTI)

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

    2013-01-09

    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.

  16. 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.; Marra, S.L.; Herman, C.C.

    2013-07-01

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

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

    SciTech Connect (OSTI)

    Seymour, R.G.

    1995-02-17

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

  18. Low-Level Waste Requirements

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

    1999-07-09

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

  19. High-Level Waste Requirements

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

    1999-07-09

    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.

  20. HLW-OVP-97-0068 High Level Waste Management Division High-Level...

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

    ... Melter pouring problems hampered attainment early in the ... in FY97. The current planning basis indicates that all ... services are installed or emergency measures are taken. ...

  1. SIPS: A small modular process unit for the in-tank pretreatment of high-level wastes

    SciTech Connect (OSTI)

    Reich, M.; Powell, J.; Barletta, R. [Brookhaven National Lab., Upton, NY (United States)

    1996-12-31

    As a result of the U.S. weapons production program, there are now hundreds of large tanks containing highly radioactive wastes. Safe disposal of these wastes requires their processing and separations into a small volume of highly radioactive waste (HLW) and a much larger volume of low-level waste (LLW). The HLW waste would then be vitrified and transported to a geologic repository. To date, the principal approach proposed for the separation envisions a large, centralized process facility. The small in-tank processing system (SIPS) is a proposed new, small modular concept for the in-tank processing and separation of wastes into HLW and LLW output streams suitable for vitrification. Instead of pumping the retrieved tank wastes as a solid/liquid slurry over long distances to a centralized process facility, SIPS would employ a small process module, typically {approximately}1 m in diameter and 4 m long, which would be inserted into the tank. Over a period of {approx} 6 months, the module would process the solid/liquid materials in the tank, producing separated liquid HLW and liquid LLW output streams that are pumped away in two small-diameter ({approx}3-cm outside diameter) pipes. The SIPS module would be serviced by five auxiliary small pipes - a water feed pipe, a water feed pipe containing micron-size ferromagnetic particles, a nitric acid ({approx}3 M) feed pipe, and input/out pipes to hydraulically load/unload ion exchange beads.

  2. Systems study of the feasibility of high-level nuclear-waste fractionation for thermal stress control in a geologic repository: main report

    SciTech Connect (OSTI)

    McKee, R.W.; Elder, H.K.; McCallum, R.F.; Silviera, D.J.; Swanson, J.L.; Wiles, L.E.

    1983-06-01

    This study assesses the benefits and costs of fractionating the cesium and strontium (Cs/Sr) components in commercial high-level waste (HLW) to a separate waste stream for the purpose of reducing geologic-repository thermal stresses in the region of the HLW. System costs are developed for a broad range of conditions comparing the Cs/Sr fractionation concept with disposal of 10-year-old vitrified HLW and vitrified HLW aged to achieve (through decay) the same heat output as the fractionated high-level waste (FHLW). All comparisons are based on a 50,000 metric ton equivalent (MTE) system. The FHLW and the Cs/Sr waste are both disposed of as vitrified waste but emplaced in separate areas of a basalt repository. The FHLW is emplaced in high-integrity packages at relatively high waste loading but low heat loading, while the Cs/Sr waste is emplaced in minimum-integrity packages at relatively high heat loading in a separate region of the repository. System cost comparisons are based on minimum cost combinations of canister diameter, waste concentration, and canister spacing in a basalt repository. The effects on both long- and near-term safety considerations are also addressed. The major conclusion is that the Cs/Sr fractionation concept offers the prospect of a substantial total system cost advantage for HLW disposal if reduced HLW package temperatures in a basalt repository are desired. However, there is no cost advantage if currently designated maximum design temperatures are acceptable. Aging the HLW for 50 to 100 years can accomplish similar results at equivalent or lower costs. 37 figures, 58 tables.

  3. High Level Waste System Impacts from Small Column Ion Exchange Implementation

    SciTech Connect (OSTI)

    McCabe, D. J.; Hamm, L. L.; Aleman, S. E.; Peeler, D. K.; Herman, C. C.; Edwards, T. B.

    2005-08-18

    The objective of this task is to identify potential waste streams that could be treated with the Small Column Ion Exchange (SCIX) and perform an initial assessment of the impact of doing so on the High-Level Waste (HLW) system. Design of the SCIX system has been performed as a backup technology for decontamination of High-Level Waste (HLW) at the Savannah River Site (SRS). The SCIX consists of three modules which can be placed in risers inside underground HLW storage tanks. The pump and filter module and the ion exchange module are used to filter and decontaminate the aqueous tank wastes for disposition in Saltstone. The ion exchange module contains Crystalline Silicotitanate (CST in its engineered granular form is referred to as IONSIV{reg_sign} IE-911), and is selective for removal of cesium ions. After the IE-911 is loaded with Cs-137, it is removed and the column is refilled with a fresh batch. The grinder module is used to size-reduce the cesium-loaded IE-911 to make it compatible with the sludge vitrification system in the Defense Waste Processing Facility (DWPF). If installed at the SRS, this SCIX would need to operate within the current constraints of the larger HLW storage, retrieval, treatment, and disposal system. Although the equipment has been physically designed to comply with system requirements, there is also a need to identify which waste streams could be treated, how it could be implemented in the tank farms, and when this system could be incorporated into the HLW flowsheet and planning. This document summarizes a preliminary examination of the tentative HLW retrieval plans, facility schedules, decontamination factor targets, and vitrified waste form compatibility, with recommendations for a more detailed study later. The examination was based upon four batches of salt solution from the currently planned disposition pathway to treatment in the SCIX. Because of differences in capabilities between the SRS baseline and SCIX, these four batches were

  4. Evaluation of Options for Permanent Geologic Disposal of Spent NuclearFuel and High-Level Radioactive Waste

    Office of Energy Efficiency and Renewable Energy (EERE)

    [In Support of a Comprehensive National Nuclear Fuel Cycle Strategy, Volumes I and II (Appendices)] This study provides a technical basis for informing policy decisions regarding strategies for the management and permanent disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) in the United States requiring geologic isolation.

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

    SciTech Connect (OSTI)

    THIEN MG; WELLS BE; ADAMSON DJ

    2010-01-14

    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.

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

    SciTech Connect (OSTI)

    Dees, L.A.

    1994-08-15

    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.

  7. 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

    2008-01-15

    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.

  8. High-Level Waste Inventory

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

    Analysis of Alternatives for Disposition of the Idaho Calcined High-Level Waste Inventory Volume 1 - Summary Report U.S. Department of Energy Office of Environmental Management April 2016 U.S. DOE-EM Independent Analysis of Alternatives for Disposition of the Idaho Calcined High-Level Waste Inventory Volume 1- Summary Report April 2016 ii THIS PAGE INTENTIONALLY LEFT BLANK U.S. DOE-EM Independent Analysis of Alternatives for Disposition of the Idaho Calcined High-Level Waste Inventory Volume 1-

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

    SciTech Connect (OSTI)

    Herman, C

    2007-02-14

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

  10. Commercial waste treatment program annual progress report for FY 1983

    SciTech Connect (OSTI)

    McElroy, J.L.; Burkholder, H.C. (comps.)

    1984-02-01

    This annual report describes progress during FY 1983 relating to technologies under development by the Commercial Waste Treatment Program, including: development of glass waste form and vitrification equipment for high-level wastes (HLW); waste form development and process selection for transuranic (TRU) wastes; pilot-scale operation of a radioactive liquid-fed ceramic melter (LFCM) system for verifying the reliability of the reference HLW treatment proces technology; evaluation of treatment requirements for spent fuel as a waste form; second-generation waste form development for HLW; and vitrification process control and product quality assurance technologies.

  11. Decomposition of tetraphenylborate precipitates used to isolate Cs-137 from Savannah River Site high-level waste

    SciTech Connect (OSTI)

    Ferrara, D.M.; Bibler, N.E.; Ha, B.C.

    1993-03-01

    This paper presents results of the radioactive demonstration of the Precipitate Hydrolysis Process (PHP) that will be performed in the Defense Waste Processing Facility (DWPF) at the Savannah River Site. The PHP destroys the tetraphenylborate precipitate that is used at SRS to isolate Cs-137 from caustic High-Level Waste (HLW) supernates. This process is necessary to decrease the amount of organic compounds going to the melter in the DWPF. Actual radioactive precipitate containing Cs-137 was used for this demonstration.

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

    SciTech Connect (OSTI)

    G. Radulesscu; J.S. Tang

    2000-06-07

    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

  13. Disposal of defense spent fuel and HLW at the Idaho Chemical Processing Plant

    SciTech Connect (OSTI)

    Ermold, L.F.; Loo, H.H.; Klingler, R.D.; Herzog, J.D.; Knecht, D.A.

    1993-06-01

    Irradiated nuclear fuel has been reprocessed at the Idaho Chemical Processing Plant (ICPP) since 1953 to recover uranium-235 and krypton-85 for the US Department of Energy (DOE). The resulting acidic high-level radioactive waste (HLW) has been solidified to a calcine since 1963 and stored in stainless steel underground bins enclosed by concrete vaults. Several different types of unprocessed irradiated DOE-owned fuels are also in storage at the ICPP. In April, 1992, DOE announced that spent fuel would no longer be reprocessed to recover enriched uranium and called for a shutdown of the reprocessing facilities at the ICPP. A new Spent Fuel and HLW Technology Development program was subsequently initiated to develop technologies for immobilizing ICPP spent fuels and HLW for disposal, in accordance with the Nuclear Waste Policy Act. The Program elements include Systems Analysis, Graphite Fuel Disposal, Other Spent Fuel Disposal, Sodium-Bearing Liquid Waste Processing, Calcine Immobilization, and Metal Recycle/Waste Minimization. This paper presents an overview of the ICPP radioactive wastes and current spent fuels, with an emphasis on the description of HLW and spent fuels requiring repository disposal.

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

    SciTech Connect (OSTI)

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

    2013-11-13

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

  15. US Department of Energy Storage of Spent Fuel and High Level Waste

    SciTech Connect (OSTI)

    Sandra M Birk

    2010-10-01

    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.

  16. Secondary Low-Level Waste Treatment Strategy Analysis

    SciTech Connect (OSTI)

    D.M. LaRue

    1999-05-25

    The objective of this analysis is to identify and review potential options for processing and disposing of the secondary low-level waste (LLW) that will be generated through operation of the Monitored Geologic Repository (MGR). An estimate of annual secondary LLW is generated utilizing the mechanism established in ''Secondary Waste Treatment Analysis'' (Reference 8.1) and ''Secondary Low-Level Waste Generation Rate Analysis'' (Reference 8.5). The secondary LLW quantities are based on the spent fuel and high-level waste (HLW) arrival schedule as defined in the ''Controlled Design Assumptions Document'' (CDA) (Reference 8.6). This analysis presents estimates of the quantities of LLW in its various forms. A review of applicable laws, codes, and standards is discussed, and a synopsis of those applicable laws, codes, and standards and their impacts on potential processing and disposal options is presented. The analysis identifies viable processing/disposal options in light of the existing laws, codes, and standards, and then evaluates these options in regard to: (1) Process and equipment requirements; (2) LLW disposal volumes; and (3) Facility requirements.

  17. RIVER PROTECTION PROJECT MISSION ANALYSIS WASTE BLENDING STUDY

    SciTech Connect (OSTI)

    SHUFORD DH; STEGEN G

    2010-04-19

    Preliminary evaluation for blending Hanford site waste with the objective of minimizing the amount of high-level waste (HLW) glass volumes without major changes to the overall waste retrieval and processing sequences currently planned. The evaluation utilizes simplified spreadsheet models developed to allow screening type comparisons of blending options without the need to use the Hanford Tank Waste Operations Simulator (HTWOS) model. The blending scenarios evaluated are expected to increase tank farm operation costs due to increased waste transfers. Benefit would be derived from shorter operating time period for tank waste processing facilities, reduced onsite storage of immobilized HLW, and reduced offsite transportation and disposal costs for the immobilized HLW.

  18. EM Waste Acceptance Product Specification (WAPS) for Vitrified...

    Office of Environmental Management (EM)

    EM Waste Acceptance Product Specification (WAPS) for Vitrified High-Level Waste Forms Presentation to the HLW Corporate Board July 24, 2008 By Tony KlukKen Picha 2 Background * ...

  19. River Protection Project (RPP) Immobilized High Level Waste (HLW) Interim Storage Plan

    SciTech Connect (OSTI)

    BRIGGS, M.G.

    2000-09-22

    This document replaces HNF-1751, Revision 1. It incorporates updates to reflect changes in programmatic direction associated with the vitrification plant contract and associated DOE-ORP guidance. In addition it includes planning associated with failed/used melter and sample handling and disposition work scope. The document also includes format modifications and section numbering update consistent with CH2M HILL Hanford Group, Inc. procedures.

  20. HLW Melter Control Strategy Without Visual Feedback VSL-12R2500-1 Rev 0

    SciTech Connect (OSTI)

    Kruger, A A.; Joseph, Innocent; Matlack, Keith S.; Callow, Richard A.; Abramowitz, Howard; Pegg, Ian L.; Brandys, Marek; Kot, Wing K.

    2012-11-13

    Plans for the treatment of high level waste (HL W) at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) are based upon the inventory of the tank wastes, the anticipated performance of the pretreatment processes, and current understanding of the capability of the borosilicate glass waste form [I]. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat and mass transfer and increase glass melting rates. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of ~ 1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HL W waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150°C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage.

  1. EMPIRICAL MODEL FOR FORMULATION OF CRYSTAL-TOLERANT HLW GLASSES

    SciTech Connect (OSTI)

    KRUGER AA; MATYAS J; HUCKLEBERRY AR; VIENNA JD; RODRIGUEZ CA

    2012-03-07

    Historically, high-level waste (HLW) glasses have been formulated with a low liquideus temperature (T{sub L}), or temperature at which the equilibrium fraction of spinel crystals in the melt is below 1 vol % (T{sub 0.01}), nominally below 1050 C. These constraints cannot prevent the accumulation of large spinel crystals in considerably cooler regions ({approx} 850 C) of the glass discharge riser during melter idling and significantly limit the waste loading, which is reflected in a high volume of waste glass, and would result in high capital, production, and disposal costs. A developed empirical model predicts crystal accumulation in the riser of the melter as a function of concentration of spinel-forming components in glass, and thereby provides guidance in formulating crystal-tolerant glasses that would allow high waste loadings by keeping the spinel crystals small and therefore suspended in the glass.

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

    SciTech Connect (OSTI)

    Cunnane, J.C.; Bates, J.K.; Bradley, C.R.

    1994-03-01

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

  3. ATW system impact on high-level waste

    SciTech Connect (OSTI)

    Arthur, E.D.

    1992-01-01

    This report discusses the Accelerator Transmutation of Waste (ATW) concept which aims at destruction of key long-lived radionuclides in high-level nuclear waste (HLW), both fission products and actinides. This focus makes it different from most other transmutation concepts which concentrate primarily on actinide burning. The ATW system uses an accelerator-driven, sub-critical assembly to create an intense thermal neutron environment for radionuclide transmutation. This feature allows rapid transmutation under low-inventory system conditions, which in turn, has a direct impact on the size of chemical separations and materials handling components of the system. Inventories in ATW are factors of eight to thirty times smaller than reactor systems of equivalent thermal power. Chemical separations systems are relatively small in scale and can be optimized to achieve high decontamination factors and minimized waste streams. The low-inventory feature also directly impacts material amounts remaining in the system at its end of life. In addition to its low-inventory operation, the accelerator-driven neutron source features of ATW are key to providing a sufficient level of neutrons to allow transmutation of long-lived fission products.

  4. Geological aspects of the nuclear waste disposal problem

    SciTech Connect (OSTI)

    Laverov, N.P.; Omelianenko, B.L.; Velichkin, V.I.

    1994-06-01

    For the successful solution of the high-level waste (HLW) problem in Russia one must take into account such factors as the existence of the great volume of accumulated HLW, the large size and variety of geological conditions in the country, and the difficult economic conditions. The most efficient method of HLW disposal consists in the maximum use of protective capacities of the geological environment and in using inexpensive natural minerals for engineered barrier construction. In this paper, the principal trends of geological investigation directed toward the solution of HLW disposal are considered. One urgent practical aim is the selection of sites in deep wells in regions where the HLW is now held in temporary storage. The aim of long-term investigations into HLW disposal is to evaluate geological prerequisites for regional HLW repositories.

  5. Examining the role of canister cooling conditions on the formation of nepheline from nuclear waste glasses

    SciTech Connect (OSTI)

    Christian, J. H.

    2015-09-01

    Nepheline (NaAlSiO₄) crystals can form during slow cooling of high-level waste (HLW) glass after it has been poured into a waste canister. Formation of these crystals can adversely affect the chemical durability of the glass. The tendency for nepheline crystallization to form in a HLW glass increases with increasing concentrations of Al₂O₃ and Na₂O.

  6. Examining the role of canister cooling conditions on the formation of nepheline from nuclear waste glasses

    SciTech Connect (OSTI)

    Christian, J. H.

    2015-09-01

    Nepheline (NaAlSiO?) crystals can form during slow cooling of high-level waste (HLW) glass after it has been poured into a waste canister. Formation of these crystals can adversely affect the chemical durability of the glass. The tendency for nepheline crystallization to form in a HLW glass increases with increasing concentrations of Al?O? and Na?O.

  7. Waste Treatment and Immobilization Plant Progress

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

    Waste Treatment and Immobilization Plant Progress Hanford Advisory Board requested action:  Based on progress discussions, the Hanford Advisory Board will develop and advocate an effective public communication strategy for use by the Waste Treatment and Immobilization Plant Assistant Manager/Federal Project Director Progress discussions on the following:  High-level waste (HLW) authorization to proceed with full production engineering:  HLW Safety Design Strategy approval and

  8. Preliminary waste form characteristics report Version 1.0. Revision 1

    SciTech Connect (OSTI)

    Stout, R.B.; Leider, H.R.

    1991-10-11

    This report focuses on radioactive waste form characteristics that will be used to design a waste package and an engineered barrier system (EBS) for a suitable repository as part of the Yucca Mountain Project. The term waste form refers to irradiated reactor fuel, other high-level waste (HLW) in various physical forms, and other radioactive materials (other than HLW) which are received for emplacement in a geologic repository. Any encapsulating of stabilizing matrix is also referred to as a waste form.

  9. Low-level waste program technical strategy

    SciTech Connect (OSTI)

    Bledsoe, K.W.

    1994-10-01

    The Low-Level Waste Technical Strategy document describes the mechanisms which the Low-Level Waste Program Office plans to implement to achieve its mission. The mission is to manage the receipt, immobilization, packaging, storage/disposal and RCRA closure (of the site) of the low-level Hanford waste (pretreated tank wastes) in an environmentally sound, safe and cost-effective manner. The primary objective of the TWRS Low-level waste Program office is to vitrify the LLW fraction of the tank waste and dispose of it onsite.

  10. 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

    2006-10-02

    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.

  11. 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.

    2002-02-26

    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.

  12. Assessment of degradation concerns for spent fuel, high-level wastes, and transuranic wastes in monitored retrievalbe storage

    SciTech Connect (OSTI)

    Guenther, R.J.; Gilbert, E.R.; Slate, S.C.; Partain, W.L.; Divine, J.R.; Kreid, D.K.

    1984-01-01

    It has been concluded that there are no significant degradation mechanisms that could prevent the design, construction, and safe operation of monitored retrievable storage (MRS) facilities. However, there are some long-term degradation mechanisms that could affect the ability to maintain or readily retrieve spent fuel (SF), high-level wastes (HLW), and transuranic wastes (TRUW) several decades after emplacement. Although catastrophic failures are not anticipated, long-term degradation mechanisms have been identified that could, under certain conditions, cause failure of the SF cladding and/or failure of TRUW storage containers. Stress rupture limits for Zircaloy-clad SF in MRS range from 300 to 440/sup 0/C, based on limited data. Additional tests on irradiated Zircaloy (3- to 5-year duration) are needed to narrow this uncertainty. Cladding defect sizes could increase in air as a result of fuel density decreases due to oxidation. Oxidation tests (3- to 5-year duration) on SF are also needed to verify oxidation rates in air and to determine temperatures below which monitoring of an inert cover gas would not be required. Few, if any, changes in the physical state of HLW glass or canisters or their performance would occur under projected MRS conditions. The major uncertainty for HLW is in the heat transfer through cracked glass and glass devitrification above 500/sup 0/C. Additional study of TRUW is required. Some fraction of present TRUW containers would probably fail within the first 100 years of MRS, and some TRUW would be highly degraded upon retrieval, even in unfailed containers. One possible solution is the design of a 100-year container. 93 references, 28 figures, 17 tables.

  13. Chemical Speciation of Americium, Curium and Selected Tetravalent Actinides in High Level Waste

    SciTech Connect (OSTI)

    Felmy, Andrew R.

    2006-06-01

    Large volumes of high-level waste (HLW) currently stored in tanks at DOE sites contain both sludges and supernatants. The sludges are composed of insoluble precipitates of actinides, radioactive fission products, and nonradioactive components. The supernatants are alkaline carbonate solutions, which can contain soluble actinides, fission products, metal ions, and high concentrations of major electrolytes including sodium hydroxide, nitrate, nitrite, phosphate, carbonate, aluminate, sulfate, and organic complexants. The organic complexants include several compounds that can form strong aqueous complexes with actinide species and fission products including ethylenediaminetetraacetic acid (EDTA), N-(2-hydroxyethyl)ethylenediaminetriacetic acid (HEDTA), nitrilotriacetic acid (NTA), iminodiacetic acid (IDA), citrate, glycolate, gluconate, and degradation products, formate and oxalate.

  14. Chemical Speciation of Americium, Curium and Selected Tetravalent Actinides in High Level Waste

    SciTech Connect (OSTI)

    Felmy, Andrew R.

    2005-06-01

    Large volumes of high-level waste (HLW) currently stored in tanks at DOE sites contain both sludges and supernatants. The sludges are composed of insoluble precipitates of actinides, radioactive fission products, and nonradioactive components. The supernatants are alkaline carbonate solutions, which can contain soluble actinides, fission products, metal ions, and high concentrations of major electrolytes including sodium hydroxide, nitrate, nitrite, phosphate, carbonate, aluminate, sulfate, and organic complexants. The organic complexants include several compounds that can form strong aqueous complexes with actinide species and fission products including ethylenediaminetetraacetic acid (EDTA), N-(2-hydroxyethyl)ethylenediaminetriacetic acid (HEDTA), nitrilotriacetic acid (NTA), iminodiacetic acid (IDA), citrate, glycolate, gluconate, and degradation products, formate and oxalate.

  15. HLW System Integrated Project Team

    Office of Environmental Management (EM)

    l W S Hi h l W S High Level Waste System High Level Waste System Integrated Project Team ... and skilled kf Developing and deploying t h l i This document is intended for planning ...

  16. 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)

    2008-07-01

    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

  17. Low-level-waste-form criteria

    SciTech Connect (OSTI)

    Barletta, R.E.; Davis, R.E.

    1982-01-01

    Efforts in five areas are reported: technical considerations for a high-integrity container for resin wastes; permissible radionuclide loadings for organic ion exchange resin wastes; technical factors affecting low-level waste form acceptance requirements of the proposed 10 CFR 61 and draft BTP; modeling of groundwater transport; and analysis of soils from low-level waste disposal sites (Barnwell, Hanford, and Sheffield). (DLC)

  18. EIS-0356: Retrieval, Treatment and Disposal of Tank Wastes and Closure of Single-Shell Tanks at the Hanford Site, Richland, WA

    Broader source: Energy.gov [DOE]

    This EIS analyzes DOE's proposed retrieval, treatment, and disposal of the waste being managed in the high-level waste (HLW) tank farms at the Hanford Site near Richland, Washington, and closure of the 149 single-shell tanks (SSTs) and associated facilities in the HLW tank farms.

  19. ACTUAL-WASTE TESTS OF ENHANCED CHEMICAL CLEANING FOR RETRIEVAL OF SRS HLW SLUDGE TANK HEELS AND DECOMPOSITION OF OXALIC ACID

    SciTech Connect (OSTI)

    Martino, C.; King, W.; Ketusky, E.

    2012-01-12

    Savannah River National Laboratory conducted a series of tests on the Enhanced Chemical Cleaning (ECC) process using actual Savannah River Site waste material from Tanks 5F and 12H. Testing involved sludge dissolution with 2 wt% oxalic acid, the decomposition of the oxalates by ozonolysis (with and without the aid of ultraviolet light), the evaporation of water from the product, and tracking the concentrations of key components throughout the process. During ECC actual waste testing, the process was successful in decomposing oxalate to below the target levels without causing substantial physical or chemical changes in the product sludge.

  20. Qualification of Innovative High Level Waste Pipeline Unplugging Technologies

    SciTech Connect (OSTI)

    McDaniel, D.; Gokaltun, S.; Varona, J.; Awwad, A.; Roelant, D.; Srivastava, R.

    2008-07-01

    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

  1. World first in high level waste vitrification - A review of French vitrification industrial achievements

    SciTech Connect (OSTI)

    Brueziere, J.; Chauvin, E. [AREVA, 1 place Jean Millier, 92084 Paris La Defense (France); Piroux, J.C. [Joint Vitrification Laboratory - LCV, Marcoule, BP171, 30207 Bagnols sur Ceze (France)

    2013-07-01

    AREVA has more than 30 years experience in operating industrial HLW (High Level radioactive Waste) vitrification facilities (AVM - Marcoule Vitrification Facility, R7 and T7 facilities). This vitrification technology was based on borosilicate glasses and induction-heating. AVM was the world's first industrial HLW vitrification facility to operate in-line with a reprocessing plant. The glass formulation was adapted to commercial Light Water Reactor fission products solutions, including alkaline liquid waste concentrates as well as platinoid-rich clarification fines. The R7 and T7 facilities were designed on the basis of the industrial experience acquired in the AVM facility. The AVM vitrification process was implemented at a larger scale in order to operate the R7 and T7 facilities in-line with the UP2 and UP3 reprocessing plants. After more than 30 years of operation, outstanding record of operation has been established by the R7 and T7 facilities. The industrial startup of the CCIM (Cold Crucible Induction Melter) technology with enhanced glass formulation was possible thanks to the close cooperation between CEA and AREVA. CCIM is a water-cooled induction melter in which the glass frit and the waste are melted by direct high frequency induction. This technology allows the handling of highly corrosive solutions and high operating temperatures which permits new glass compositions and a higher glass production capacity. The CCIM technology has been implemented successfully at La Hague plant.

  2. Technical assessment of processes to enable recycling of low-level contaminated metal waste

    SciTech Connect (OSTI)

    Reimann, G.A.

    1991-10-01

    Accumulations of metal waste exhibiting low levels of radioactivity (LLCMW) have become a national burden, both financially and environmentally. Much of this metal could be considered as a resource. The Department of Energy was assigned the task of inventorying and classifying LLCMW, identifying potential applications, and applying and/or developing the technology necessary to enable recycling. One application for recycled LLCMW is high-quality canisters for permanent repository storage of high-level waste (HLW). As many as 80,000 canisters will be needed by 2035. Much of the technology needed to decontaminate LLCMW has already been developed, but no integrated process has been described, even on a pilot scale, for recycling LLCMW into HLW canisters. This report reviews practices for removal of radionuclides and for producing low carbon stainless steel. Contaminants that readily form oxides may be reduced to below de minimis levels and combined with a slag. Most of the radioactivity remaining in the ingot is concentrated in the inclusions. Radionuclides that chemically resemble the elements that comprise stainless steel can not be removed effectively. Slag compositions, current melting practices, and canister fabrication techniques were reviewed.

  3. The Production of Advanced Glass Ceramic HLW Forms using Cold Crucible Induction Melter

    SciTech Connect (OSTI)

    Veronica J Rutledge; Vince Maio

    2013-10-01

    Cold Crucible Induction Melters (CCIMs) will favorably change how High-Level radioactive Waste (from nuclear fuel recovery) is treated in the 21st century. Unlike the existing Joule-Heated Melters (JHMs) currently in operation for the glass-based immobilization of High-Level Waste (HLW), CCIMs offer unique material features that will increase melt temperatures, increase throughput, increase mixing, increase loading in the waste form, lower melter foot prints, eliminate melter corrosion and lower costs. These features not only enhance the technology for producing HLW forms, but also provide advantageous attributes to the waste form by allowing more durable alternatives to glass. This paper discusses advantageous features of the CCIM, with emphasis on features that overcome the historical issues with the JHMs presently utilized, as well as the benefits of glass ceramic waste forms over borosilicate glass waste forms. These advantages are then validated based on recent INL testing to demonstrate a first-of-a-kind formulation of a non-radioactive ceramic-based waste form utilizing a CCIM.

  4. Vitrification of high level nuclear waste inside ambient temperature disposal containers using inductive heating: The SMILE system

    SciTech Connect (OSTI)

    Powell, J.; Reich, M.; Barletta, R.

    1996-03-01

    A new approach, termed SMILE (Small Module Inductively Loaded Energy), for the vitrification of high level nuclear wastes (HLW) is described. Present vitrification systems liquefy the HLW solids and associated frit material in large high temperature melters. The molten mix is then poured into small ({approximately}1 m{sup 3}) disposal canisters, where it solidifies and cools. SMILE eliminates the separate, large high temperature melter. Instead, the BLW solids and frit melt inside the final disposal containers, using inductive heating. The contents then solidify and cool in place. The SMILE modules and the inductive heating process are designed so that the outer stainless can of the module remains at near ambient temperature during the process cycle. Module dimensions are similar to those of present disposal containers. The can is thermally insulated from the high temperature inner container by a thin layer of refractory alumina firebricks. The inner container is a graphite crucible lined with a dense alumina refractory that holds the HLW and fiit materials. After the SMILE module is loaded with a slurry of HLW and frit solids, an external multi-turn coil is energized with 30-cycle AC current. The enclosing external coil is the primary of a power transformer, with the graphite crucible acting as a single turn ``secondary.`` The induced current in the ``secondary`` heats the graphite, which in turn heats the HLW and frit materials. The first stage of the heating process is carried out at an intermediate temperature to drive off remnant liquid water and water of hydration, which takes about 1 day. The small fill/vent tube to the module is then sealed off and the interior temperature raised to the vitrification range, i.e., {approximately}1200C. Liquefaction is complete after approximately 1 day. The inductive heating then ceases and the module slowly loses heat to the environment, allowing the molten material to solidify and cool down to ambient temperature.

  5. High Level Waste Management Division High. Level Waste System...

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

    ... The loose waste sludge was then immobilized by blowing in dry powdered grout. The dry ... melter pouring is suspended, at least one steam atomized scrubber operates all the time. ...

  6. Cost estimate of high-level radioactive waste containers for the Yucca Mountain Site Characterization Project

    SciTech Connect (OSTI)

    Russell, E.W.; Clarke, W.; Domian, H.A.; Madson, A.A.

    1991-08-01

    This report summarizes the bottoms-up cost estimates for fabrication of high-level radioactive waste disposal containers based on the Site Characterization Plan Conceptual Design (SCP-CD). These estimates were acquired by Babcock and Wilcox (B&S) under sub-contract to Lawrence Livermore National Laboratory (LLNL) for the Yucca Mountain Site Characterization Project (YMP). The estimates were obtained for two leading container candidate materials (Alloy 825 and CDA 715), and from other three vendors who were selected from a list of twenty solicited. Three types of container designs were analyzed that represent containers for spent fuel, and for vitrified high-level waste (HLW). The container internal structures were assumed to be AISI-304 stainless steel in all cases, with an annual production rate of 750 containers. Subjective techniques were used for estimating QA/QC costs based on vendor experience and the specifications derived for the LLNL-YMP Quality Assurance program. In addition, an independent QA/QC analysis is reported which was prepared by Kasier Engineering. Based on the cost estimates developed, LLNL recommends that values of $825K and $62K be used for the 1991 TSLCC for the spent fuel and HLW containers, respectively. These numbers represent the most conservative among the three vendors, and are for the high-nickel anstenitic steel (Alloy 825). 6 refs., 7 figs.

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

    SciTech Connect (OSTI)

    Stefanovsky, Sergey; Marra, James; Lebedev, Vladimir

    2014-01-13

    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.%.

  8. Physical, Chemical and Structural Evolution of Zeolite-Containing Waste Forms Produced from Metakaolinite and Calcined Sodium Bearing Waste (HLW and/or LLW)

    SciTech Connect (OSTI)

    Grutzeck, Michael W.

    2005-06-27

    Zeolites are extremely versatile. They can adsorb liquids and gases and serve as cation exchange media. They occur in nature as well cemented deposits. The ancient Romans used blocks of zeolitized tuff as a building material. Using zeolites for the management of radioactive waste is not a new idea, but a process by which the zeolites can be made to act as a cementing agent is. Zeolitic materials are relatively easy to synthesize from a wide range of both natural and man-made substances. The process under study is derived from a well known method in which metakaolin (an impure thermally dehydroxylated kaolinite heated to {approx}700 C containing traces of quartz and mica) is mixed with sodium hydroxide (NaOH) and reacted in slurry form (for a day or two) at mildly elevated temperatures. The zeolites form as finely divided powders containing micrometer ({micro}m) sized crystals. However, if the process is changed slightly and only just enough concentrated sodium hydroxide solution is added to the metakaolinite to make a thick crumbly paste and then the paste is compacted and cured under mild hydrothermal conditions (60-200 C), the mixture will form a hard ceramic-like material containing distinct crystalline tectosilicate minerals (zeolites and feldspathoids) imbedded in an X-ray amorphous hydrated sodium aluminosilicate matrix. Due to its lack of porosity and vitreous appearance we have chosen to call this composite a ''hydroceramic''.

  9. A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests

    SciTech Connect (OSTI)

    Thien, Mike G.; Barnes, Steve M.

    2013-01-17

    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.

  10. Supplemental Immobilization Cast Stone Technology Development and Waste Form Qualification Testing Plan

    SciTech Connect (OSTI)

    Westsik, Joseph H.; Serne, R. Jeffrey; Pierce, Eric M.; Cozzi, Alex; Chung, Chul-Woo; Swanberg, David J.

    2013-05-31

    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). The pretreatment facility will have the capacity to separate all of the tank wastes into the HLW and LAW fractions, and the HLW Vitrification Facility will have the capacity to vitrify all of the HLW. However, a second immobilization facility will be needed for the expected volume of LAW requiring immobilization. A number of alternatives, including Cast Stone—a cementitious waste form—are being considered to provide the additional LAW immobilization capacity.

  11. PUREX low-level waste radionuclide characterization

    SciTech Connect (OSTI)

    Ellis, M.W.; LeBaron, G.J.

    1995-01-16

    The PUREX low-level waste (LLW) radionuclide characterization document describes the methodology for the characterization of solid LLW and solid low-level mixed waste (MW) with the respect to radiological characteristics. This document only serves as an overview of the PUREX radionuclide characterization methodology and provides specific examples for how the radionuclide distribution is derived. It would be impractical to provide all background information in this document. If further clarification and background information is required, consult the PUREX Regulatory Compliance group files. This document applies to only that waste generated in or is the responsibility of the PUREX facilities. The US Department of Energy (DOE) establishes the requirements for radioactive solid waste in DOE Order 5820.2A Radioactive Waste Management. Chapters 2 and 3 from DOE Order 5820.2A requires that generators of solid wastes in the LLW categories and the radioactive mixed waste subcategories: (1) identify the major radionuclides in each solid waste matrix and (2) determine the radionuclide concentrations and waste classes of their solid wastes. In addition, the Order also requires each generator to carry out a compliance program that ensures the proper certification of the solid waste generated.

  12. ESTIMATING HIGH LEVEL WASTE MIXING PERFORMANCE IN HANFORD DOUBLE SHELL TANKS

    SciTech Connect (OSTI)

    THIEN MG; GREER DA; TOWNSON P

    2011-01-13

    The ability to effectively mix, sample, certify, and deliver consistent batches of high level waste (HLW) feed from the Hanford double shell tanks (DSTs) 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. The Department of Energy's (DOE's) Tank Operations Contractor (TOC), Washington River Protection Solutions (WRPS) is currently demonstrating mixing, sampling, and batch transfer performance in two different sizes of small-scale DSTs. The results of these demonstrations will be used to estimate full-scale DST mixing performance and provide the key input to a programmatic decision on the need to build a dedicated feed certification facility. This paper discusses the results from initial mixing demonstration activities and presents data evaluation techniques that allow insight into the performance relationships of the two small tanks. The next steps, sampling and batch transfers, of the small scale demonstration activities are introduced. A discussion of the integration of results from the mixing, sampling, and batch transfer tests to allow estimating full-scale DST performance is presented.

  13. Tank Waste Remediation System optimized processing strategy

    SciTech Connect (OSTI)

    Slaathaug, E.J.; Boldt, A.L.; Boomer, K.D.; Galbraith, J.D.; Leach, C.E.; Waldo, T.L.

    1996-03-01

    This report provides an alternative strategy evolved from the current Hanford Site Tank Waste Remediation System (TWRS) programmatic baseline for accomplishing the treatment and disposal of the Hanford Site tank wastes. This optimized processing strategy performs the major elements of the TWRS Program, but modifies the deployment of selected treatment technologies to reduce the program cost. The present program for development of waste retrieval, pretreatment, and vitrification technologies continues, but the optimized processing strategy reuses a single facility to accomplish the separations/low-activity waste (LAW) vitrification and the high-level waste (HLW) vitrification processes sequentially, thereby eliminating the need for a separate HLW vitrification facility.

  14. Waste Segregation Based on Derived Clearance Levels

    SciTech Connect (OSTI)

    Garisto, N.C.; Parhizgari, Z.

    2008-07-01

    This paper describes the methodology and results of a radiological modeling in support of an application to release very low level radiologically contaminated waste from regulatory control and allow its haulage and disposal in a hazardous waste landfill. The Canadian regulatory body responsible for licensing operations involving nuclear materials (the Canadian Nuclear Safety Commission), has not yet formally defined clearance levels for free release of low level radiologically contaminated waste. The IAEA clearance levels have been derived for certain situations and receptor characteristics, which might be too conservative for an actual case. A site-specific pathways analysis was therefore completed to define conditional clearance levels using the concept of de minimis dose limit. Derived Conditional Clearance Levels were calculated for each radionuclide based on the maximally exposed hypothetical individuals to determine whether each waste stream can be 'cleared' from regulatory controls. The results showed that haulage of the waste from the station to the haulage/processing facility and transportation of waste or sludge from the haulage/processing facility to the disposal facility, handling of the waste or sludge at the haulage/processing facility, and incineration and/or disposal of waste or sludge at the disposal facility would not expose the workers to doses above 0.1 {mu}Sv/yr., which is less than the de minimis dose limit of 10 {mu}Sv/yr. (authors)

  15. 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.

    2013-11-13

    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.

  16. Data quality objectives for TWRS privatization phase 1: confirm tank T is an appropriate feed source for high-level waste feed batch X

    SciTech Connect (OSTI)

    NGUYEN, D.M.

    1999-06-01

    The U.S. Department of Energy-Richland Operations Office (DOE-RL) has initiated Phase 1 of a two-phase privatization strategy for treatment and immobilization of high-level waste (HLW) that is currently managed by the Hanford Tank Waste Remediation System (TWRS) Project. In this strategy, DOE will purchase services from a contractor-owned and operated facility under a fixed price. The Phase 1 TWRS privatization contract requires that the Project Hanford Management Contract (PHMC) contractors, on behalf of DOE, deliver HLW feed in specified quantities and composition to the Privatization Contractor in a timely manner (DOE-RL 1996). Additional requirements are imposed by the interface control document (ICD) for HLW feed (PHMC 1997). In response to these requirements, the Tank Waste Remediation System Operation and Utilization Plan (TWRSO and UP) (Kirkbride et al. 1997) was prepared by the PHMC. The TWRSO and UP, as updated by the Readiness-To-Proceed (RTP) deliverable (Payne et al. 1998), establishes the baseline operating scenario for the delivery of HLW feed to the Privatization Contractor. The scenario specifies tanks from which HLW will be provided for each feed batch, the operational activities needed to prepare and deliver each batch, and the timing of these activities. The operating scenario was developed based on current knowledge of waste composition and chemistry, waste transfer methods, and operating constraints such as tank farm logistics and availability of tank space. A project master baseline schedule (PMBS) has been developed to implement the operating scenario. The PMBS also includes activities aimed at reducing programmatic risks. One of the activities, ''Confirm Tank TI is Acceptable for Feed,'' was identified to verify the basis used to develop the scenario Additional data on waste quantity, physical and chemical characteristics, and transfer properties will be needed to support this activity. This document describes the data quality objective

  17. Russian low-level waste disposal program

    SciTech Connect (OSTI)

    Lehman, L.

    1993-03-01

    The strategy for disposal of low-level radioactive waste in Russia differs from that employed in the US. In Russia, there are separate authorities and facilities for wastes generated by nuclear power plants, defense wastes, and hospital/small generator/research wastes. The reactor wastes and the defense wastes are generally processed onsite and disposed of either onsite, or nearby. Treating these waste streams utilizes such volume reduction techniques as compaction and incineration. The Russians also employ methods such as bitumenization, cementation, and vitrification for waste treatment before burial. Shallow land trench burial is the most commonly used technique. Hospital and research waste is centrally regulated by the Moscow Council of Deputies. Plans are made in cooperation with the Ministry of Atomic Energy. Currently the former Soviet Union has a network of low-level disposal sites located near large cities. Fifteen disposal sites are located in the Federal Republic of Russia, six are in the Ukraine, and one is located in each of the remaining 13 republics. Like the US, each republic is in charge of management of the facilities within their borders. The sites are all similarly designed, being modeled after the RADON site near Moscow.

  18. Tank farms compacted low-level waste

    SciTech Connect (OSTI)

    Hetzer, D.C.

    1997-08-01

    This report describes the process of Low-Level Waste (LLW) volume reduction by compaction. Also included is the data used for characterization of LLW destined for compaction. Scaling factors (ratios) are formed based on data contained in this report.

  19. Tank farms compacted low level waste

    SciTech Connect (OSTI)

    Waters, M.S., Westinghouse Hanford

    1996-07-01

    This report describes the process of Low Level Waste (LLW) volume reduction by compaction. Also included is the data used for characterization of LLW destined for compaction. Scaling factors (ratios) are formed based on data contained in this report.

  20. A Prototype Performance Assessment Model for Generic Deep Borehole Repository for High-Level Nuclear Waste - 12132

    SciTech Connect (OSTI)

    Lee, Joon H.; Arnold, Bill W.; Swift, Peter N.; Hadgu, Teklu; Freeze, Geoff; Wang, Yifeng

    2012-07-01

    A deep borehole repository is one of the four geologic disposal system options currently under study by the U.S. DOE to support the development of a long-term strategy for geologic disposal of commercial used nuclear fuel (UNF) and high-level radioactive waste (HLW). The immediate goal of the generic deep borehole repository study is to develop the necessary modeling tools to evaluate and improve the understanding of the repository system response and processes relevant to long-term disposal of UNF and HLW in a deep borehole. A prototype performance assessment model for a generic deep borehole repository has been developed using the approach for a mined geological repository. The preliminary results from the simplified deep borehole generic repository performance assessment indicate that soluble, non-sorbing (or weakly sorbing) fission product radionuclides, such as I-129, Se-79 and Cl-36, are the likely major dose contributors, and that the annual radiation doses to hypothetical future humans associated with those releases may be extremely small. While much work needs to be done to validate the model assumptions and parameters, these preliminary results highlight the importance of a robust seal design in assuring long-term isolation, and suggest that deep boreholes may be a viable alternative to mined repositories for disposal of both HLW and UNF. (authors)

  1. Steam stripping of polycyclic aromatics from simulated high-level radioactive waste

    SciTech Connect (OSTI)

    Lambert, D.P.; Shah, H.B.; Young, S.R.; Edwards, R.E.; Carter, J.T.

    1992-12-31

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) will be the United States` first facility to process High Level radioactive Waste (HLW) into a borosilicate glass matrix. The removal of aromatic precipitates by hydrolysis, evaporation, liquid-liquid extraction and decantation will be a key step in the processing of the HLW. This step, titled the Precipitate Hydrolysis Process, has been demonstrated by the Savannah River Technology Center with the Precipitate Hydrolysis Experimental Facility (PHEF). The mission of the PHEF is to demonstrate processing of simulated high level radioactive waste which contains tetraphenylborate precipitates and nitrite. Aqueous washing or nitrite destruction is used to reduce nitrite. Formic acid with a copper catalyst is used to hydrolyze tetraphenylborate (TPB). The primary offgases are benzene, carbon dioxide, nitrous oxide, and nitric oxide. Hydrolysis of TPB in the presence of nitrite results in the production of polycyclic aromatics and aromatic amines (referred as high boiling organics) such as biphenyl, diphenylamine, terphenyls etc. The decanter separates the organic (benzene) and aqueous phase, but the high boiling organic separation is difficult. This paper focuses on the evaluation of the operating strategies, including steam stripping, to maximize the removal of the high boiling organics from the aqueous stream. Two areas were investigated, (1) a stream stripping comparison of the late wash flowsheet to the HAN flowsheet and (2) the extraction performance of the original decanter to the new decanter. The focus of both studies was to minimize the high boiling organic content of the Precipitate Hydrolysis Aqueous (PHA) product in order to minimize downstream impacts caused by organic deposition.

  2. Low-Level Waste Disposal Facility Federal Review Group Manual...

    Office of Environmental Management (EM)

    Low-Level Waste Disposal Facility Federal Review Group Manual Low-Level Waste Disposal Facility Federal Review Group Manual This Revision 3 of the Low-Level Waste Disposal Facility ...

  3. West Valley Demonstration Project Low-Level Waste Shipment |...

    Office of Environmental Management (EM)

    Low-Level Waste Shipment West Valley Demonstration Project Low-Level Waste Shipment West Valley Demonstration Project Low-Level Waste Shipment (406.16 KB) More Documents & ...

  4. 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

    2012-07-01

    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)

  5. LOW-LEVEL WASTE DISPOSAL FACILITY FEDERAL REVIEW GROUP EXECUTION...

    Office of Environmental Management (EM)

    LOW-LEVEL WASTE DISPOSAL FACILITY FEDERAL REVIEW GROUP EXECUTION PLAN Los Alamos National ... Safety and Security LFRG Low-Level Waste Disposal Facility Federal Review Group LLW ...

  6. High-Level Waste Melter Study Report

    SciTech Connect (OSTI)

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

    2001-07-13

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

  7. Low-level waste feed staging plan

    SciTech Connect (OSTI)

    Certa, P.J.; Grams, W.H.; McConville, C.M.; L. W. Shelton, L.W.; Slaathaug, E.J., Westinghouse Hanford

    1996-08-12

    The `Preliminary Low-Level Waste Feed Staging Plan` was updated to reflect the latest requirement in the Tank Waste Remediation Privatization Request for Proposals (RFP) and amendments. The updated plan develops the sequence and transfer schedule for retrieval of DST supernate by the management and integration contractor and delivery of the staged supernate to the private low-activity waste contractors for treatment. Two DSTs are allocated as intermediate staging tanks. A transfer system conflict analysis provides part of the basis for determining transfer system upgrade requirements to support both low-activity and high-level waste feed delivery. The intermediate staging tank architecture and retrieval system equipment are provided as a planning basis until design requirements documents are prepared. The actions needed to successfully implement the plan are identified. These include resolution of safety issues and changes to the feed envelope limits, minimum order quantities, and desired batch sizes.

  8. High Level Waste Remote Handling Equipment in the Melter Cave Support Handling System at the Hanford Waste Treatment Plant

    SciTech Connect (OSTI)

    Bardal, M.A.; Darwen, N.J.

    2008-07-01

    Cold war plutonium production led to extensive amounts of radioactive waste stored in tanks at the Department of Energy's (DOE) Hanford site. Bechtel National, Inc. is building the largest nuclear Waste Treatment Plant in the world located at the Department of Energy's Hanford site to immobilize the millions of gallons of radioactive waste. The site comprises five main facilities; Pretreatment, High Level Waste vitrification, Low Active Waste vitrification, an Analytical Lab and the Balance of Facilities. The pretreatment facilities will separate the high and low level waste. The high level waste will then proceed to the HLW facility for vitrification. Vitrification is a process of utilizing a melter to mix molten glass with radioactive waste to form a stable product for storage. The melter cave is designated as the High Level Waste Melter Cave Support Handling System (HSH). There are several key processes that occur in the HSH cell that are necessary for vitrification and include: feed preparation, mixing, pouring, cooling and all maintenance and repair of the process equipment. Due to the cell's high level radiation, remote handling equipment provided by PaR Systems, Inc. is required to install and remove all equipment in the HSH cell. The remote handling crane is composed of a bridge and trolley. The trolley supports a telescoping tube set that rigidly deploys a TR 4350 manipulator arm with seven degrees of freedom. A rotating, extending, and retracting slewing hoist is mounted to the bottom of the trolley and is centered about the telescoping tube set. Both the manipulator and slewer are unique to this cell. The slewer can reach into corners and the manipulator's cross pivoting wrist provides better operational dexterity and camera viewing angles at the end of the arm. Since the crane functions will be operated remotely, the entire cell and crane have been modeled with 3-D software. Model simulations have been used to confirm operational and maintenance

  9. HIGH-LEVEL WASTE GLASS FORMULATION MODEL SENSITIVITY STUDY 2009 GLASS FORMULATION MODEL VERSUS 1996 GLASS FORMULATION MODEL

    SciTech Connect (OSTI)

    BELSHER JD; MEINERT FL

    2009-12-07

    This document presents the differences between two HLW glass formulation models (GFM): The 1996 GFM and 2009 GFM. A glass formulation model is a collection of glass property correlations and associated limits, as well as model validity and solubility constraints; it uses the pretreated HLW feed composition to predict the amount and composition of glass forming additives necessary to produce acceptable HLW glass. The 2009 GFM presented in this report was constructed as a nonlinear optimization calculation based on updated glass property data and solubility limits described in PNNL-18501 (2009). Key mission drivers such as the total mass of HLW glass and waste oxide loading are compared between the two glass formulation models. In addition, a sensitivity study was performed within the 2009 GFM to determine the effect of relaxing various constraints on the predicted mass of the HLW glass.

  10. Twelfth annual US DOE low-level waste management conference

    SciTech Connect (OSTI)

    Not Available

    1990-01-01

    The papers in this document comprise the proceedings of the Department of Energy's Twelfth Annual Low-Level Radioactive Waste Management Conference, which was held in Chicago, Illinois, on August 28 and 29, 1990. General subjects addressed during the conference included: mixed waste, low-level radioactive waste tracking and transportation, public involvement, performance assessment, waste stabilization, financial assurance, waste minimization, licensing and environmental documentation, below-regulatory-concern waste, low-level radioactive waste temporary storage, current challenges, and challenges beyond 1990.

  11. High-level waste inventory, characteristics, generation, and facility assessment for treatment, storage, and disposal alternatives considered in the US Department of Energy eenvironmental management programmatic environmental impact statement

    SciTech Connect (OSTI)

    Folga, S.M.; Conzelmann, G.; Gillette, J.L.; Kier, P.H.; Poch, L.A.

    1996-12-01

    This report provides data and information needed to support the risk and impact assessments of high-level waste (HLW) management alternatives in the U.S. Department of Energy Waste Management (WM) Programmatic Environmental Impact Statement (PEIS). Available data on the physical form, chemical and isotopic composition, storage locations, and other waste characteristics of interest are presented. High-level waste management follows six implementation phases: current storage, retrieval, pretreatment, treatment, interim canister storage, and geologic repository disposal; pretreatment, treatment, and repository disposal are outside the scope of the WM PEIS. Brief descriptions of current and planned HLW management facilities are provided, including information on the type of waste managed in the facility, costs, product form, resource requirements, emissions, and current and future status. Data sources and technical and regulatory assumptions are identified. The range of HLW management alternatives (including decentralized, regionalized, and centralized approaches) is described. The required waste management facilities include expanded interim storage facilities under the various alternatives. Resource requirements for construction (e.g., land and materials) and operation (e.g., energy and process chemicals), work force, costs, effluents, design capacities, and emissions are presented for each alternative.

  12. Grout Placement and Property Evaluation for Closing Hanford High-Level Waste Tanks - Scale-Up Testing

    SciTech Connect (OSTI)

    LANGTON, CHRISTINE

    2003-12-15

    Hanford has 149 single-shell high level waste (HLW) tanks that were constructed between 1943 and 1964. Many of these tanks have leaked or are suspected of leaking HLW into the soil above the ground water. Consequently, a major effort is ongoing to transfer the liquid portion of the waste to the 28 newer, double-shell tanks. Savannah River National Laboratory (SRNL) was tasked to develop grout formulations for the three-layer closure concept selected by CH2M HILL for closing Tank C-106. These grout formulations were also evaluated for use as fill materials in the next six tanks scheduled to be closed. The overall scope consisted of both bench-scale testing to confirm mix designs and scale-up testing to confirm placement properties. This report provides results of the scale-up testing for the three-phase tank closure strategy. It also contains information on grouts for equipment and riser filling. The three-phase fill strategy is summarized as follows: Phase I fill encapsulates and minimizes dispersion of the residual waste in the tank. This fill is referred to as the Stabilization Layer and consists of the Stabilization Grout. The Phase II fill provides structural stability to the tank system and prevents subsidence. It is referred to as the Structural Layer and consists of the Structural Grout. A final Phase III fill consists of a grout designed to provide protection against intrusion and is referred to as the Capping Layer or Capping Grout.

  13. 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.

    2011-09-19

    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

  14. Amended Record of Decision for the Idaho High-Level Waste (HLW) and Facilities Disposition Final Environmental Impact Statement

    Office of Environmental Management (EM)

    All About Solar and Real Estate With nearly one million solar energy systems installed on residential rooftops across the country, buying or selling a solar home is getting more and more common. Just like a renovated kitchen or a finished basement increases a home's value, solar has been shown to boost home valuation and shorten a home's time on the market. In addition, new borrowing tools give solar-interested homeowners access to financing programs that can lower the cost of capital and unlock

  15. High-level radioactive wastes. Supplement 1

    SciTech Connect (OSTI)

    McLaren, L.H.

    1984-09-01

    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.

  16. High Level Waste Corporate Board Newsletter - 06/03/08

    Office of Environmental Management (EM)

    2008 UPCOMING EVENTS: Next High-Level Waste Corporate Board meeting will be held at ... Needs Collection Prioritization * Waste Acceptance Product Specification This ...

  17. 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.

    1980-08-01

    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.

  18. Some Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository Study (The Yucca Mountain Project)

    SciTech Connect (OSTI)

    F. Hua; P. Pasupathi; N. Brown; K. Mon

    2005-09-19

    The safe disposal of radioactive waste requires that the waste be isolated from the environment until radioactive decay has reduced its toxicity to innocuous levels for plants, animals, and humans. All of the countries currently studying the options for disposing of high-level nuclear waste (HLW) have selected deep geologic formations to be the primary barrier for accomplishing this isolation. In U.S.A., the Nuclear Waste Policy Act of 1982 (as amended in 1987) designated Yucca Mountain in Nevada as the potential site to be characterized for high-level nuclear waste (HLW) disposal. Long-term containment of waste and subsequent slow release of radionuclides into the geosphere will rely on a system of natural and engineered barriers including a robust waste containment design. The waste package design consists of a highly corrosion resistant Ni-based Alloy 22 cylindrical barrier surrounding a Type 316 stainless steel inner structural vessel. The waste package is covered by a mailbox-shaped drip shield composed primarily of Ti Grade 7 with Ti Grade 24 structural support members. The U.S. Yucca Mountain Project has been studying and modeling the degradation issues of the relevant materials for some 20 years. This paper reviews the state-of-the-art understanding of the degradation processes based on the past 20 years studies on Yucca Mountain Project (YMP) materials degradation issues with focus on interaction between the in-drift environmental conditions and long-term materials degradation of waste packages and drip shields within the repository system during the 10,000 years regulatory period. This paper provides an overview of the current understanding of the likely degradation behavior of the waste package and drip shield in the repository after the permanent closure of the facility. The degradation scenario discussed in this paper include aging and phase instability, dry oxidation, general and localized corrosion, stress corrosion cracking and hydrogen induced

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

    SciTech Connect (OSTI)

    Reboul, S. H.

    2013-09-30

    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.

  20. Mixed Low-Level Radioactive Waste (MLLW) Primer

    SciTech Connect (OSTI)

    W. E. Schwinkendorf

    1999-04-01

    This document presents a general overview of mixed low-level waste, including the regulatory definitions and drivers, the manner in which the various kinds of mixed waste are regulated, and a discussion of the waste treatment options.

  1. Low level tank waste disposal study

    SciTech Connect (OSTI)

    Mullally, J.A.

    1994-09-29

    Westinghouse Hanford Company (WHC) contracted a team consisting of Los Alamos Technical Associates (LATA), British Nuclear Fuel Laboratories (BNFL), Southwest Research Institute (SwRI), and TRW through the Tank Waste Remediation System (TWRS) Technical Support Contract to conduct a study on several areas concerning vitrification and disposal of low-level-waste (LLW). The purpose of the study was to investigate how several parameters could be specified to achieve full compliance with regulations. The most restrictive regulation governing this disposal activity is the National Primary Drinking Water Act which sets the limits of exposure to 4 mrem per year for a person drinking two liters of ground water daily. To fully comply, this constraint would be met independently of the passage of time. In addition, another key factor in the investigation was the capability to retrieve the disposed waste during the first 50 years as specified in Department of Energy (DOE) Order 5820.2A. The objective of the project was to develop a strategy for effective long-term disposal of the low-level waste at the Hanford site.

  2. Waste Processing Annual Technology Development Report 2007 |...

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

    More Documents & Publications System Planning for Low-Activity Waste at Hanford Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility Caustic Recovery Technology

  3. LIFE ESTIMATION OF HIGH LEVEL WASTE TANK STEEL FOR F-TANK FARM CLOSURE PERFORMANCE ASSESSMENT - 9310

    SciTech Connect (OSTI)

    Subramanian, K; Bruce Wiersma, B; Stephen Harris, S

    2009-01-12

    High level radioactive waste (HLW) is stored in underground carbon steel storage tanks at the Savannah River Site. The underground tanks will be closed by removing the bulk of the waste, chemical cleaning, heel removal, stabilizing remaining residuals with tailored grout formulations, and severing/sealing external penetrations. The life of the carbon steel materials of construction in support of the performance assessment has been completed. The estimation considered general and localized corrosion mechanisms of the tank steel exposed to grouted conditions. A stochastic approach was followed to estimate the distributions of failures based upon mechanisms of corrosion accounting for variances in each of the independent variables. The methodology and results used for one-type of tank is presented.

  4. Tank waste remediation system optimized processing strategy with an altered treatment scheme

    SciTech Connect (OSTI)

    Slaathaug, E.J.

    1996-03-01

    This report provides an alternative strategy evolved from the current Hanford Site Tank Waste Remediation System (TWRS) programmatic baseline for accomplishing the treatment and disposal of the Hanford Site tank wastes. This optimized processing strategy with an altered treatment scheme performs the major elements of the TWRS Program, but modifies the deployment of selected treatment technologies to reduce the program cost. The present program for development of waste retrieval, pretreatment, and vitrification technologies continues, but the optimized processing strategy reuses a single facility to accomplish the separations/low-activity waste (LAW) vitrification and the high-level waste (HLW) vitrification processes sequentially, thereby eliminating the need for a separate HLW vitrification facility.

  5. 12/2000 Low-Level Waste Disposal Capacity Report Version 2 |...

    Office of Environmental Management (EM)

    Waste Management Waste Disposition 122000 Low-Level Waste Disposal Capacity Report Version 2 122000 Low-Level Waste Disposal Capacity Report Version 2 The purpose of this ...

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

    SciTech Connect (OSTI)

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

    2014-05-15

    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

  7. CRAD, Low-Level Radioactive Waste Management - April 30, 2015...

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

    Low-Level Radioactive Waste Management - April 30, 2015 (EA CRAD 31-11, Rev. 0) CRAD, Low-Level Radioactive Waste Management - April 30, 2015 (EA CRAD 31-11, Rev. 0) April 2015...

  8. Maintenance Guide for DOE Low-Level Waste Disposal Facility ...

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

    Maintenance Guide for U.S. Department of Energy Low-Level Waste Disposal Facility Performance Assessments and Composite Analyses Maintenance Guide for DOE Low-Level Waste Disposal ...

  9. Maintenance Guide for DOE Low-Level Waste Disposal Facility

    Office of Environmental Management (EM)

    Maintenance Guide for U.S. Department of Energy Low-Level Waste Disposal Facility ... for U.S. Department of Energy Low-Level Waste Disposal Facility Performance Assessments ...

  10. Low-Level Waste Disposal Facility Federal Review Group Manual

    Office of Environmental Management (EM)

    LEVEL WASTE DISPOSAL FACILITY FEDERAL REVIEW GROUP MANUAL REVISION 3 JUNE 2008 (This page ... 3, June 200S Concurrence The Low-Level Waste Disposal Facility Federal Review Group ...

  11. HLW Glass Studies: Development of Crystal-Tolerant HLW Glasses

    SciTech Connect (OSTI)

    Matyas, Josef; Huckleberry, Adam R.; Rodriguez, Carmen P.; Lang, Jesse B.; Owen, Antionette T.; Kruger, Albert A.

    2012-04-02

    In our study, a series of lab-scale crucible tests were performed on designed glasses of different compositions to further investigate and simulate the effect of Cr, Ni, Fe, Al, Li, and RuO2 on the accumulation rate of spinel crystals in the glass discharge riser of the HLW melter. The experimental data were used to expand the compositional region covered by an empirical model developed previously (Maty et al. 2010b), improving its predictive performance. We also investigated the mechanism for agglomeration of particles and impact of agglomerates on accumulation rate. In addition, the TL was measured as a function of temperature and composition.

  12. Lid design for low level waste container

    DOE Patents [OSTI]

    Holbrook, Richard H.; Keener, Wendell E.

    1995-01-01

    A container for low level waste includes a shell and a lid. The lid has a frame to which a planar member is welded. The lid frame includes a rectangular outer portion made of square metal tubing, a longitudinal beam extending between axial ends of the rectangular outer portion, and a transverse beam extending between opposite lateral sides of the rectangular outer portion. Two pairs of diagonal braces extend between the longitudinal beam and the four corners of the rectangular outer portion of the frame.

  13. Lid design for low level waste container

    DOE Patents [OSTI]

    Holbrook, R.H.; Keener, W.E.

    1995-02-28

    A container for low level waste includes a shell and a lid. The lid has a frame to which a planar member is welded. The lid frame includes a rectangular outer portion made of square metal tubing, a longitudinal beam extending between axial ends of the rectangular outer portion, and a transverse beam extending between opposite lateral sides of the rectangular outer portion. Two pairs of diagonal braces extend between the longitudinal beam and the four corners of the rectangular outer portion of the frame. 6 figs.

  14. Initiating the Validation of CCIM Processability for Multi-phase all Ceramic (SYNROC) HLW Form: Plan for Test BFY14CCIM-C

    SciTech Connect (OSTI)

    Vince Maio

    2014-08-01

    This plan covers test BFY14CCIM-C which will be a first–of–its-kind demonstration for the complete non-radioactive surrogate production of multi-phase ceramic (SYNROC) High Level Waste Forms (HLW) using Cold Crucible Induction Melting (CCIM) Technology. The test will occur in the Idaho National Laboratory’s (INL) CCIM Pilot Plant and is tentatively scheduled for the week of September 15, 2014. The purpose of the test is to begin collecting qualitative data for validating the ceramic HLW form processability advantages using CCIM technology- as opposed to existing ceramic–lined Joule Heated Melters (JHM) currently producing BSG HLW forms. The major objectives of BFY14CCIM-C are to complete crystalline melt initiation with a new joule-heated resistive starter ring, sustain inductive melting at temperatures between 1600 to 1700°C for two different relatively high conductive materials representative of the SYNROC ceramic formation inclusive of a HLW surrogate, complete melter tapping and pouring of molten ceramic material in to a preheated 4 inch graphite canister and a similar canister at room temperature. Other goals include assessing the performance of a new crucible specially designed to accommodate the tapping and pouring of pure crystalline forms in contrast to less recalcitrant amorphous glass, assessing the overall operational effectiveness of melt initiation using a resistive starter ring with a dedicated power source, and observing the tapped molten flow and subsequent relatively quick crystallization behavior in pans with areas identical to standard HLW disposal canisters. Surrogate waste compositions with ceramic SYNROC forming additives and their measured properties for inductive melting, testing parameters, pre-test conditions and modifications, data collection requirements, and sampling/post-demonstration analysis requirements for the produced forms are provided and defined.

  15. Melter Throughput Enhancements for High-Iron HLW

    SciTech Connect (OSTI)

    Kruger, A. A.; Gan, Hoa; Joseph, Innocent; Pegg, Ian L.; Matlack, Keith S.; Chaudhuri, Malabika; Kot, Wing

    2012-12-26

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

  16. Performance Enhancements to the Hanford Waste Treatment and Immobilization Plant Low-Activity Waste Vitrification System

    SciTech Connect (OSTI)

    Hamel, W. F. [Office of River Protection, U.S. Department of Energy, 2400 Stevens Drive, Richland, WA 99354 (United States); Gerdes, K. [U.S. Department of Energy, 19901 Germantown Road, Germantown, MD 20874 (United States); Holton, L. K. [Pacific Northwest National Laboratory, PO Box 999, Richland, WA 99352 (United States); Pegg, I.L. [Vitreous State Laboratory, The Catholic University of America, 620 Michigan Avenue NE, Washington, DC 20064 (United States); Bowan, B.W. [Duratek, Inc., 10100 Old Columbia Road, Columbia, Maryland 21046 (United States)

    2006-07-01

    The U.S Department of Energy Office of River Protection (DOE-ORP) is constructing a Waste Treatment and Immobilization Plant (WTP) for the treatment and vitrification of underground tank wastes stored at the Hanford Site in Washington State. The WTP comprises four major facilities: a pretreatment facility to separate the tank waste into high level waste (HLW) and low-activity waste (LAW) process streams, a HLW vitrification facility to immobilize the HLW fraction; a LAW vitrification facility to immobilize the LAW fraction, and an analytical laboratory to support the operations of all four treatment facilities. DOE has established strategic objectives to optimize the performance of the WTP facilities and the LAW and HLW waste forms to reduce the overall schedule and cost for treatment and vitrification of the Hanford tank wastes. This strategy has been implemented by establishing performance expectations in the WTP contract for the facilities and waste forms. In addition, DOE, as owner-operator of the WTP facilities, continues to evaluate 1) the design, to determine the potential for performance above the requirements specified in the WTP contract; and 2) improvements in production of the LAW and HLW waste forms. This paper reports recent progress directed at improving production of the LAW waste form. DOE's initial assessment, which is based on the work reported in this paper, is that the treatment rate of the WTP LAW vitrification facility can be increased by a factor of 2 to 4 with a combination of revised glass formulations, modest increases in melter glass operating temperatures, and a second-generation LAW melter with a larger surface area. Implementing these improvements in the LAW waste immobilization capability can benefit the LAW treatment mission by reducing the cost of waste treatment. (authors)

  17. Redox Control For Hanford HLW Feeds VSL-12R2530-1, REV 0

    SciTech Connect (OSTI)

    Kruger, A. A.; Matlack, Keith S.; Pegg, Ian L.; Kot, Wing K.; Joseph, Innocent

    2012-12-13

    The principal objectives of this work were to investigate the effects of processing simulated Hanford HLW at the estimated maximum concentrations of nitrates and oxalates and to identify strategies to mitigate any processing issues resulting from high concentrations of nitrates and oxalates. This report provides results for a series of tests that were performed on the DM10 melter system with simulated C-106/AY-102 HLW. The tests employed simulated HLW feeds containing variable amounts of nitrates and waste organic compounds corresponding to maximum concentrations proj ected for Hanford HLW streams in order to determine their effects on glass production rate, processing characteristics, glass redox conditions, melt pool foaming, and the tendency to form secondary phases. Such melter tests provide information on key process factors such as feed processing behavior, dynamic effects during processing, processing rates, off-gas amounts and compositions, foaming control, etc., that cannot be reliably obtained from crucible melts.

  18. Application of EPA regulations to low-level radioactive waste

    SciTech Connect (OSTI)

    Bowerman, B.S.; Piciulo, P.L.

    1985-01-01

    The survey reported here was conducted with the intent of identifying categories of low-level radioactive wastes which would be classified under EPA regulations 40 CFR Part 261 as hazardous due to the chemical properties of the waste. Three waste types are identified under these criteria as potential radioactive mixed wastes: wastes containing organic liquids; wastes containing lead metal; and wastes containing chromium. The survey also indicated that certain wastes, specific to particular generators, may also be radioactive mixed wastes. Ultimately, the responsibility for determining whether a facility's wastes are mixed wastes rest with the generator. However, the uncertainties as to which regulations are applicable, and the fact that no legal definition of mixed wastes exists, make such a determination difficult. In addition to identifying mixed wastes, appropriate methods for the management of mixed wastes must be defined. In an ongoing study, BNL is evaluating options for the management of mixed wastes. These options will include segregation, substitution, and treatments to reduce or eliminate chemical hazards associated with the wastes listed above. The impacts of the EPA regulations governing hazardous wastes on radioactive mixed waste cannot be assessed in detail until the applicability of these regulations is agreed upon. This issue is still being discussed by EPA and NRC and should be resolved in the near future. Areas of waste management which may affect generators of mixed wastes include: monitoring/tracking of wastes before shipment; chemical testing of wastes; permits for treatment of storage of wastes; and additional packaging requirements. 3 refs., 1 fig., 2 tabs.

  19. Idaho High-Level Waste & Facilities Disposition, Final Environmental...

    Office of Environmental Management (EM)

    A.1 Introduction A-1 A.2 Methodology A-1 A.3 High-Level Waste Treatment and Interim ... Evaluation Process A-6 A.4 Low-Activity Waste Disposal Site Selection A-6 A.4.1 ...

  20. High Level Waste Corporate Board Newsletter - 09/11/08

    Office of Environmental Management (EM)

    Waste Federal Review Group (LFRG) in Washington, DC on 16-18 September 2008. Contact Maureen O'Dell for details (MAUREEN.O'DELL@hq.doe.gov) Next High-Level Waste Corporate ...

  1. Draft low level waste technical summary

    SciTech Connect (OSTI)

    Powell, W.J.; Benar, C.J.; Certa, P.J.; Eiholzer, C.R.; Kruger, A.A.; Norman, E.C.; Mitchell, D.E.; Penwell, D.E.; Reidel, S.P.; Shade, J.W.

    1995-09-01

    The purpose of this document is to present an outline of the Hanford Site Low-Level Waste (LLW) disposal program, what it has accomplished, what is being done, and where the program is headed. This document may be used to provide background information to personnel new to the LLW management/disposal field and to those individuals needing more information or background on an area in LLW for which they are not familiar. This document should be appropriate for outside groups that may want to learn about the program without immediately becoming immersed in the details. This document is not a program or systems engineering baseline report, and personnel should refer to more current baseline documentation for critical information.

  2. Enterprise Assessments Operational Awareness Record, Waste Treatment and

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

    Immobilization Plant - December 2014 | Department of Energy December 2014 Enterprise Assessments Operational Awareness Record, Waste Treatment and Immobilization Plant - December 2014 December 2014 Operational Awareness Record for the Observation of Waste Treatment and Immobilization Plant High Level Waste Facility Radioactive Liquid Waste Disposal System Hazards Analysis Activities (EA-WTP-HLW-2014-08-18(a)) The Office of Nuclear Safety and Environmental Assessments, within the U.S.

  3. Reference commercial high-level waste glass and canister definition.

    SciTech Connect (OSTI)

    Slate, S.C.; Ross, W.A.; Partain, W.L.

    1981-09-01

    This report presents technical data and performance characteristics of a high-level waste glass and canister intended for use in the design of a complete waste encapsulation package suitable for disposal in a geologic repository. The borosilicate glass contained in the stainless steel canister represents the probable type of high-level waste product that will be produced in a commercial nuclear-fuel reprocessing plant. Development history is summarized for high-level liquid waste compositions, waste glass composition and characteristics, and canister design. The decay histories of the fission products and actinides (plus daughters) calculated by the ORIGEN-II code are presented.

  4. Vitrified waste option study report

    SciTech Connect (OSTI)

    Lopez, D.A.; Kimmitt, R.R.

    1998-02-01

    A {open_quotes}Settlement Agreement{close_quotes} between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste (HLW) now stored at the Idaho Chemical Processing Plant (ICPP) will be treated so that it is ready to be moved out of Idaho for disposal by a target date of 2035. This report investigates vitrification treatment of all ICPP calcine, including the existing and future HLW calcine resulting from calcining liquid Sodium-Bearing Waste (SBW). Currently, the SBW is stored in the tank farm at the ICPP. Vitrification of these wastes is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the calcined waste and casting the vitrified mass into stainless steel canisters that will be ready to be moved out of the Idaho for disposal by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory (INEEL) until they are sent to a HLW national repository. The operating period for vitrification treatment will be from 2013 through 2032; all HLW will be treated and in storage by the end of 2032.

  5. REGIONAL BINNING FOR CONTINUED STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTES

    SciTech Connect (OSTI)

    W. Lee Poe, Jr

    1998-10-01

    In the Continued Storage Analysis Report (CSAR) (Reference 1), DOE decided to analyze the environmental consequences of continuing to store the commercial spent nuclear fuel (SNF) at 72 commercial nuclear power sites and DOE-owned spent nuclear fuel and high-level waste at five Department of Energy sites by region rather than by individual site. This analysis assumes that three commercial facilities pairs--Salem and Hope Creek, Fitzpatrick and Nine-Mile Point, and Dresden and Moms--share common storage due to their proximity to each other. The five regions selected for this analysis are shown on Figure 1. Regions 1, 2, and 3 are the same as those used by the Nuclear Regulatory Commission in their regulatory oversight of commercial power reactors. NRC Region 4 was subdivided into two regions to more appropriately define the two different climates that exist in NRC Region 4. A single hypothetical site in each region was assumed to store all the SNF and HLW in that region. Such a site does not exist and has no geographic location but is a mathematical construct for analytical purposes. To ensure that the calculated results for the regional analyses reflect appropriate inventory, facility and material degradation, and radionuclide transport, the waste inventories, engineered barriers, and environmental conditions for the hypothetical sites were developed from data for each of the existing sites within the given region. Weighting criteria to account for the amount and types of SNF and HLW at each site were used in the development of the environmental data for the regional site, such that the results of the analyses for the hypothetical site were representative of the sum of the results of each actual site if they had been modeled independently. This report defines the actual site data used in development of this hypothetical site, shows how the individual site data was weighted to develop the regional site, and provides the weighted data used in the CSAR analysis. It is

  6. Decontamination and melting of low-level waste

    SciTech Connect (OSTI)

    Clements, D.W.

    1997-03-01

    This article describes the decommissioning project of the Capenhurst Diffusion Plant in Europe. Over 99 percent of the low-level waste was successfully treated and recycled. Topics include the following: decommissioning philosophy; specialized techniques including plant pretreatment, plant dismantling, size reduction, decontamination, melting, and encapsulation of waste; recycled materials and waste stream; project safety; cost drivers and savings. 5 refs., 5 figs.

  7. Identification of radioactive mixed wastes in commercial low-level wastes

    SciTech Connect (OSTI)

    Bowerman, B.S.; Kempf, C.R.; MacKenzie, D.R.; Siskind, B.; Piciulo, P.L.

    1986-01-01

    A literature review and survey were conducted on behalf of the US NRC Division of Waste Management to determine whether any commercial low-level radioactive wastes (LLW) could be considered hazardous as defined by EPA under 40 CFR Part 261. The purpose of the study was to identify broad categories of LLW which may require special management as radioactive mixed waste, and to help address uncertainties regarding the regulation of such wastes. Of 239 questionnaires sent out to reactor and non-reactor LLW generators, there were 91 responses representing 29% by volume of all low-level wastes disposed of at commercial disposal sites in 1984. The analysis of the survey results indicated that the following waste types generic to commercial LLW may be potential radioactive mixed wastes: Wastes containing oil, disposed of by reactors and industrial facilities, and representing 4.2% of the total LLW volume reported in the survey. Wastes containing organic liquids, disposed of by all types of generators, and representing 2.3% by volume of all wastes reported. Wastes containing lead metal, i.e., discarded shielding and lead containers, representing <0.1% by volume of all wastes reported. Wastes containing chromium, i.e., process wastes from nuclear power plants which use chromates as corrosion inhibitors; these represent 0.6% of the total volume reported in the survey. Certain wastes, specific to particular generators, were identified as potential mixed wastes as well.

  8. High Level Waste Corporate Board Charter

    Office of Environmental Management (EM)

    Board: * Deputy Assistant Secretary (DAS) for Engineering and Technology, EM-20 (Chair) * Director for the Office of Waste Processing, EM-21 (Deputy Chair and Executive Secretary)...

  9. Stability testing of low-level waste forms

    SciTech Connect (OSTI)

    Piciulo, P.L.; Shea, C.E.; Barletta, R.E.

    1983-01-01

    The NRC Technical Position on Waste Form identifies methods for thermal cycle testing and biodegradation testing of low-level waste forms. These tests were carried out on low-level waste forms to establish whether the tests are reasonable and can be achieved. The thermal-cycle test is believed adequate for demonstrating the thermal stability of solidified waste forms. The biodegradation tests are sufficient for distinguishing materials that are susceptible to biodegradation. However, failure of either of these tests should not be regarded of itself as an indication that the waste form will biodegrade to an extent that the form does not meet the stability requirements of 10 CFR Part 61.

  10. Idaho High-Level Waste & Facilities Disposition, Final Environmental...

    Office of Environmental Management (EM)

    must prepare an Environmental Impact Statement (EIS). Copies of the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement are available at the...

  11. Characterization of High Level Waste from a Hybrid LIFE Engine...

    Office of Scientific and Technical Information (OSTI)

    Title: Characterization of High Level Waste from a Hybrid LIFE Engine for Enhanced Repository Performance Authors: Beckett, E ; Fratoni, M Publication Date: 2010-08-25 OSTI ...

  12. Mixing Processes in High-Level Waste Tanks (Technical Report...

    Office of Scientific and Technical Information (OSTI)

    with transport in free and wall jets modeled using standard integral techniques. ... of fuel and oxygen concentrations in DOE high-level waste tanks following loss of ...

  13. TESTS WITH HIGH-BISMUTH HLW GLASSES FINAL REPORT VSL-10R1780-1 REV 0 12/13/10

    SciTech Connect (OSTI)

    MATLACK KS; KRUGER AA; JOSEPH I; GAN H; KOT WK; CHAUDHURI M; MOHR RK; MCKEOWN DA; BARDAKEI T; GONG W; BUECCHELE AC; PEGG IL

    2011-01-05

    This Final Report describes the testing of glass formulations developed for Hanford High Level Waste (HLW) containing high concentrations of bismuth. In previous work on high-bismuth HLW streams specified by the Office of River Protection (ORP), fully compliant, high waste loading compositions were developed and subjected to melter testing on the DM100 vitrification system. However, during heat treatment according to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW canister centerline cooling (CCC) curves, crucible melts of the high-bismuth glasses were observed to foam. Clearly, such an occurrence during cooling of actual HLW canisters would be highly undesirable. Accordingly, the present work involves larger-scale testing to determine whether this effect occurs under more prototypical conditions, as well as crucible-scale tests to determine the causes and potentially remediate the observed foaming behavior. The work included preparation and characterization of crucible melts designed to determine the underlying causes of the foaming behavior as well as to assess potential mitigation strategies. Testing was also conducted on the DM1200 HLW Pilot melter with a composition previously tested on the DM100 and shown to foam during crucible-scale CCC heat treatment. The DM1200 tests evaluated foaming of glasses over a range of bismuth concentrations poured into temperature-controlled, 55-gallon drums which have a diameter that is close to that of the full-scale WTP HLW canisters. In addition, the DM1200 tests provided the first large-scale melter test data on high-bismuth WTP HLW compositions, including information on processing rates, cold cap behavior and off-gas characteristics, and data from this waste composition on the prototypical DM1200 off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for ORP on the same waste composition. The scope of this study was outlined in a Test Plan that was

  14. Identification of radioactive mixed wastes in commercial low-level wastes

    SciTech Connect (OSTI)

    Bowerman, B.S.; Kempf, C.R.; MacKenzie, D.R.; Siskind, B.; Piciulo, P.L.

    1985-01-01

    A literature review and survey were conducted on behalf of the US NRC Division of Waste Management to determine whether any commercial low-level radioactive wastes (LLW) could be considered hazardous as defined by EPA under 40 CFR Part 261. The purpose of the study was to identify broad categories of LLW which may require special management as radioactive mixed waste, and to help address uncertainties regarding the regulation of such wastes. Of 239 questionnaires sent out to reactor and non-reactor LLW generators, there were 91 responses representing 29% by volume of all low-level wastes disposed of at commercial disposal sites in 1984. The analysis of the survey results indicated that three waste streams generic to commercial LLW may be potential radioactive mixed wastes. These are as follows: (1) wastes containing organic liquids, disposed of by all types of generators and representing approx. =2.3% by volume of all wastes reported; (2) wastes containing lead metal, i.e., discarded shielding and lead containers, representing <0.1% by volume of all wastes reported; and (3) wastes containing chromium, i.e., process wastes from nuclear power plants which use chromates as corrosion inhibitors; these represent 0.6% of the total volume reported in the survey. Certain wastes, specific to particular generators, were identified as potential mixed wastes as well. 4 refs., 5 tabs.

  15. USE OF AN EQUILIBRIUM MODEL TO FORECAST DISSOLUTION EFFECTIVENESS, SAFETY IMPACTS, AND DOWNSTREAM PROCESSABILITY FROM OXALIC ACID AIDED SLUDGE REMOVAL IN SAVANNAH RIVER SITE HIGH LEVEL WASTE TANKS 1-15

    SciTech Connect (OSTI)

    KETUSKY, EDWARD

    2005-10-31

    This thesis details a graduate research effort written to fulfill the Magister of Technologiae in Chemical Engineering requirements at the University of South Africa. The research evaluates the ability of equilibrium based software to forecast dissolution, evaluate safety impacts, and determine downstream processability changes associated with using oxalic acid solutions to dissolve sludge heels in Savannah River Site High Level Waste (HLW) Tanks 1-15. First, a dissolution model is constructed and validated. Coupled with a model, a material balance determines the fate of hypothetical worst-case sludge in the treatment and neutralization tanks during each chemical adjustment. Although sludge is dissolved, after neutralization more is created within HLW. An energy balance determines overpressurization and overheating to be unlikely. Corrosion induced hydrogen may overwhelm the purge ventilation. Limiting the heel volume treated/acid added and processing the solids through vitrification is preferred and should not significantly increase the number of glass canisters.

  16. Directions in low-level radioactive waste management: A brief history of commercial low-level radioactive waste disposal

    SciTech Connect (OSTI)

    Not Available

    1990-10-01

    This report presents a history of commercial low-level radioactive waste management in the United States, with emphasis on the history of six commercially operated low-level radioactive waste disposal facilities. The report includes a brief description of important steps that have been taken during the 1980s to ensure the safe disposal of low-level waste in the 1990s and beyond. These steps include the issuance of Title 10 Code of Federal Regulations Part 61, Licensing Requirements for the Land Disposal of Radioactive Waste, the Low-Level Radioactive Waste Policy Act of 1980, the Low-Level Radioactive Waste Policy Amendments Act of 1985, and steps taken by states and regional compacts to establish additional disposal sites. 42 refs., 13 figs., 1 tab.

  17. Small Scale Mixing Demonstration Batch Transfer and Sampling Performance of Simulated HLW - 12307

    SciTech Connect (OSTI)

    Jensen, Jesse; Townson, Paul; Vanatta, Matt

    2012-07-01

    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 Plant (WTP) has been recognized as a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. At the end of 2009 DOE's Tank Operations Contractor, Washington River Protection Solutions (WRPS), awarded a contract to EnergySolutions to design, fabricate and operate a demonstration platform called the Small Scale Mixing Demonstration (SSMD) to establish pre-transfer sampling capacity, and batch transfer performance data at two different scales. This data will be used to examine the baseline capacity for a tank mixed via rotational jet mixers to transfer consistent or bounding batches, and provide scale up information to predict full scale operational performance. This information will then in turn be used to define the baseline capacity of such a system to transfer and sample batches sent to WTP. The Small Scale Mixing Demonstration (SSMD) platform consists of 43'' and 120'' diameter clear acrylic test vessels, each equipped with two scaled jet mixer pump assemblies, and all supporting vessels, controls, services, and simulant make up facilities. All tank internals have been modeled including the air lift circulators (ALCs), the steam heating coil, and the radius between the wall and floor. The test vessels are set up to simulate the transfer of HLW out of a mixed tank, and collect a pre-transfer sample in a manner similar to the proposed baseline configuration. The collected material is submitted to an NQA-1 laboratory for chemical analysis. Previous work has been done to assess tank mixing performance at both scales. This work involved a combination of unique instruments to understand the three dimensional distribution of solids using a combination of Coriolis meter measurements, in situ chord length distribution measurements, and electro

  18. Concept and Design of the JAEA KMS for Geological Disposal of HLW

    SciTech Connect (OSTI)

    Makino, Hitoshi; Osawa, Hideaki; Nakano, Katsushi; Naito, Morimasa; Umeki, Hiroyuki; Takase, Hiroyasu; McKinley, Ian G.

    2007-07-01

    The information explosion resulting from modern technology is identified as a critical problem for deep geological disposal of high-level radioactive waste (HLW). A paradigm shift is needed in the basic concept for information management. This recognition had led to the development of a 'next generation' Knowledge Management System (the JAEA KMS) that makes maximum use of recent developments in Information Technology (IT) and the methodology of Knowledge Engineering (KE) as applied in other technical fields. This paper provides a brief outline of the key concepts of the JAEA KMS and then overviews recent progress towards development of an operational system, including a 'wish list' of expected functions of the JAEA KMS, a perspective on applicability of existing methodologies and an introduction to the concept of an 'intelligent assistant'. (authors)

  19. Process for solidifying high-level nuclear waste

    DOE Patents [OSTI]

    Ross, Wayne A.

    1978-01-01

    The addition of a small amount of reducing agent to a mixture of a high-level radioactive waste calcine and glass frit before the mixture is melted will produce a more homogeneous glass which is leach-resistant and suitable for long-term storage of high-level radioactive waste products.

  20. Generalized Test Plan for the Vitrification of Simulated High-Level -Waste Calcine in the Idaho National Laboratory‘s Bench -Scale Cold Crucible Induction Melter

    SciTech Connect (OSTI)

    Vince Maio

    2011-08-01

    This Preliminary Idaho National Laboratory (INL) Test Plan outlines the chronological steps required to initially evaluate the validity of vitrifying INL surrogate (cold) High-Level-Waste (HLW) solid particulate calcine in INL's Cold Crucible Induction Melter (CCIM). Its documentation and publication satisfies interim milestone WP-413-INL-01 of the DOE-EM (via the Office of River Protection) sponsored work package, WP 4.1.3, entitled 'Improved Vitrification' The primary goal of the proposed CCIM testing is to initiate efforts to identify an efficient and effective back-up and risk adverse technology for treating the actual HLW calcine stored at the INL. The calcine's treatment must be completed by 2035 as dictated by a State of Idaho Consent Order. A final report on this surrogate/calcine test in the CCIM will be issued in May 2012-pending next fiscal year funding In particular the plan provides; (1) distinct test objectives, (2) a description of the purpose and scope of planned university contracted pre-screening tests required to optimize the CCIM glass/surrogate calcine formulation, (3) a listing of necessary CCIM equipment modifications and corresponding work control document changes necessary to feed a solid particulate to the CCIM, (4) a description of the class of calcine that will be represented by the surrogate, and (5) a tentative tabulation of the anticipated CCIM testing conditions, testing parameters, sampling requirements and analytical tests. Key FY -11 milestones associated with this CCIM testing effort are also provided. The CCIM test run is scheduled to be conducted in February of 2012 and will involve testing with a surrogate HLW calcine representative of only 13% of the 4,000 m3 of 'hot' calcine residing in 6 INL Bin Sets. The remaining classes of calcine will have to be eventually tested in the CCIM if an operational scale CCIM is to be a feasible option for the actual INL HLW calcine. This remaining calcine's make-up is HLW containing

  1. System for chemically digesting low level radioactive, solid waste material

    DOE Patents [OSTI]

    Cowan, Richard G.; Blasewitz, Albert G.

    1982-01-01

    An improved method and system for chemically digesting low level radioactive, solid waste material having a high through-put. The solid waste material is added to an annular vessel (10) substantially filled with concentrated sulfuric acid. Concentrated nitric acid or nitrogen dioxide is added to the sulfuric acid within the annular vessel while the sulfuric acid is reacting with the solid waste. The solid waste is mixed within the sulfuric acid so that the solid waste is substantilly fully immersed during the reaction. The off gas from the reaction and the products slurry residue is removed from the vessel during the reaction.

  2. Design requirements document for the interim store phase I solidified high-level waste function 4.2.4.1.2

    SciTech Connect (OSTI)

    Calmus, R.B.

    1996-09-30

    The U.S. Department of Energy (DOE) has embarked upon a course to acquire Hanford site.t,,nk waste tr:atment and immobilization services using privatized facilities. This plan contains a two-phased approach. Phase I is a ``proof-of-principle/commercial demonstration-scale`` effort and Phase II is a full-scale production effort. In accordance with the planned approach, interim storage and disposal of various products from privatized facilities are to be DOE furnished. The path forward adopted for Phase I solidified high-level waste (HLW) interim storage entails use of Vaut 2 and 3 in the Hanford Site Spent Nuclear Fuels Canister Storage Building (CSB), to be located in the Hanford Site 200 East Area. This design requirements document establishes the functions, with associated requirements, allocated to the Phase I solidified HLW interim storage system. These requirements will be used as the basis for conceptual design of the CSB and supporting systems. This document will also provide the basis for preparation of a performance specification for design and construction activities necessary to achieve the overall project mission.

  3. Review of high-level waste form properties. [146 bibliographies

    SciTech Connect (OSTI)

    Rusin, J.M.

    1980-12-01

    This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison.

  4. Exploratory study of complexant concentrate waste processing

    SciTech Connect (OSTI)

    Lumetta, G.J.; Bray, L.A.; Kurath, D.E.; Morrey, J.R.; Swanson, J.L.; Wester, D.W.

    1993-02-01

    The purpose of this exploratory study, conducted by Pacific Northwest Laboratory for Westinghouse Hanford Company, was to determine the effect of applying advanced chemical separations technologies to the processing and disposal of high-level wastes (HLW) stored in underground tanks. The major goals of this study were to determine (1) if the wastes can be partitioned into a small volume of HLW plus a large volume of low-level waste (LLW), and (2) if the activity in the LLW can be lowered enough to meet NRC Class LLW criteria. This report presents the results obtained in a brief scouting study of various processes for separating radionuclides from Hanford complexant concentrate (CC) waste.

  5. A preliminary evaluation of alternatives for disposal of INEL low-level waste and low-level mixed waste

    SciTech Connect (OSTI)

    Smith, T.H.; Roesener, W.S.; Jorgenson-Waters, M.J.

    1993-07-01

    The Mixed and Low-Level Waste Disposal Facility (MLLWDF) project was established in 1992 by the US Department of Energy Idaho Operations Office to provide enhanced disposal capabilities for Idaho National Engineering Laboratory (INEL) low-level mixed waste and low-level waste. This Preliminary Evaluation of Alternatives for Disposal of INEL Low-Level Waste and Low-Level Mixed Waste identifies and evaluates-on a preliminary, overview basis-the alternatives for disposal of that waste. Five disposal alternatives, ranging from of no-action`` to constructing and operating the MLLWDF, are identified and evaluated. Several subalternatives are formulated within the MLLWDF alternative. The subalternatives involve various disposal technologies as well as various scenarios related to the waste volumes and waste forms to be received for disposal. The evaluations include qualitative comparisons of the projected isolation performance for each alternative, and facility, health and safety, environmental, institutional, schedule, and rough order-of-magnitude life-cycle cost comparisons. The performance of each alternative is evaluated against lists of ``musts`` and ``wants.`` Also included is a discussion of other key considerations for decisionmaking. The analysis of results indicated further study is necessary to obtain the best estimate of long-term future waste volume and characteristics from the INEL Environmental Restoration activities and the expanded INEL Decontamination and Decommissioning Program.

  6. IMPACT OF NOBLE METALS AND MERCURY ON HYDROGEN GENERATION DURING HIGH LEVEL WASTE PRETREATMENT AT THE SAVANNAH RIVER SITE

    SciTech Connect (OSTI)

    Stone, M; Tommy Edwards, T; David Koopman, D

    2009-03-03

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site vitrifies radioactive High Level Waste (HLW) for repository internment. The process consists of three major steps: waste pretreatment, vitrification, and canister decontamination/sealing. HLW consists of insoluble metal hydroxides (primarily iron, aluminum, calcium, magnesium, manganese, and uranium) and soluble sodium salts (carbonate, hydroxide, nitrite, nitrate, and sulfate). The pretreatment process in the Chemical Processing Cell (CPC) consists of two process tanks, the Sludge Receipt and Adjustment Tank (SRAT) and the Slurry Mix Evaporator (SME) as well as a melter feed tank. During SRAT processing, nitric and formic acids are added to the sludge to lower pH, destroy nitrite and carbonate ions, and reduce mercury and manganese. During the SME cycle, glass formers are added, and the batch is concentrated to the final solids target prior to vitrification. During these processes, hydrogen can be produced by catalytic decomposition of excess formic acid. The waste contains silver, palladium, rhodium, ruthenium, and mercury, but silver and palladium have been shown to be insignificant factors in catalytic hydrogen generation during the DWPF process. A full factorial experimental design was developed to ensure that the existence of statistically significant two-way interactions could be determined without confounding of the main effects with the two-way interaction effects. Rh ranged from 0.0026-0.013% and Ru ranged from 0.010-0.050% in the dried sludge solids, while initial Hg ranged from 0.5-2.5 wt%, as shown in Table 1. The nominal matrix design consisted of twelve SRAT cycles. Testing included: a three factor (Rh, Ru, and Hg) study at two levels per factor (eight runs), three duplicate midpoint runs, and one additional replicate run to assess reproducibility away from the midpoint. Midpoint testing was used to identify potential quadratic effects from the three factors. A single sludge

  7. Mixed Waste Management Options: 1995 Update. National Low-Level Waste Management Program

    SciTech Connect (OSTI)

    Kirner, N.; Kelly, J.; Faison, G.; Johnson, D.

    1995-05-01

    In the original mixed Waste Management Options (DOE/LLW-134) issued in December 1991, the question was posed, ``Can mixed waste be managed out of existence?`` That study found that most, but not all, of the Nation`s mixed waste can theoretically be managed out of existence. Four years later, the Nation is still faced with a lack of disposal options for commercially generated mixed waste. However, since publication of the original Mixed Waste Management Options report in 1991, limited disposal capacity and new technologies to treat mixed waste have become available. A more detailed estimate of the Nation`s mixed waste also became available when the US Environmental Protection Agency (EPA) and the US Nuclear Regulatory Commission (NRC) published their comprehensive assessment, titled National Profile on Commercially Generated Low-Level Radioactive Mixed Waste (National Profile). These advancements in our knowledge about mixed waste inventories and generation, coupled with greater treatment and disposal options, lead to a more applied question posed for this updated report: ``Which mixed waste has no treatment option?`` Beyond estimating the volume of mixed waste requiring jointly regulated disposal, this report also provides a general background on the Atomic Energy Act (AEA) and the Resource Conservation and Recovery Act (RCRA). It also presents a methodical approach for generators to use when deciding how to manage their mixed waste. The volume of mixed waste that may require land disposal in a jointly regulated facility each year was estimated through the application of this methodology.

  8. HLW-OVP-96 C

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

    ... CIF will generate two waste streams: ash formed in the rotary kiln and scrubber blowdown ... continued storage, where it may soften the dry sludge in those tanks: Tank 17 currently ...

  9. LIFE ESTIMATION OF HIGH LEVEL WASTE TANK STEEL FOR F-TANK FARM CLOSURE PERFORMANCE ASSESSMENT

    SciTech Connect (OSTI)

    Subramanian, K

    2007-10-01

    High level radioactive waste (HLW) is stored in underground storage tanks at the Savannah River Site. The SRS is proceeding with closure of the 22 tanks located in F-Area. Closure consists of removing the bulk of the waste, chemical cleaning, heel removal, stabilizing remaining residuals with tailored grout formulations and severing/sealing external penetrations. A performance assessment is being performed in support of closure of the F-Tank Farm. Initially, the carbon steel construction materials of the high level waste tanks will provide a barrier to the leaching of radionuclides into the soil. However, the carbon steel liners will degrade over time, most likely due to corrosion, and no longer provide a barrier. The tank life estimation in support of the performance assessment has been completed. The estimation considered general and localized corrosion mechanisms of the tank steel exposed to the contamination zone, grouted, and soil conditions. The estimation was completed for Type I, Type III, and Type IV tanks in the F-Tank Farm. The tank life estimation in support of the F-Tank Farm closure performance assessment has been completed. The estimation considered general and localized corrosion mechanisms of the tank steel exposed to the contamination zone, grouted, and soil conditions. The estimation was completed for Type I, Type III, and Type IV tanks in the F-Tank Farm. Consumption of the tank steel encased in grouted conditions was determined to occur either due to carbonation of the concrete leading to low pH conditions, or the chloride-induced de-passivation of the steel leading to accelerated corrosion. A deterministic approach was initially followed to estimate the life of the tank liner in grouted conditions or in soil conditions. The results of this life estimation are shown in Table 1 and Table 2 for grouted and soil conditions respectively. The tank life has been estimated under conservative assumptions of diffusion rates. However, the same process of

  10. Low-level waste vitrification contact maintenance viability study

    SciTech Connect (OSTI)

    Leach, C.E., Westinghouse Hanford

    1996-07-12

    This study investigates the economic viability of contact maintenance in the Low-Level Waste Vitrification Facility, which is part of the Hanford Site Tank Waste Remediation System. This document was prepared by Flour Daniel, Inc., and transmitted to Westinghouse Hanford Company in September 1995.

  11. Report on Separate Disposal of Defense High-Level Radioactive Waste

    Broader source: Energy.gov [DOE]

    This is a report on the separate disposal of defense high-level radioactive waste and commercial nuclear waste.

  12. Alpha low-level stored waste systems design study

    SciTech Connect (OSTI)

    Feizollahi, F.; Teheranian, B.; Quapp, W.J.

    1992-08-01

    The Stored Waste System Design Study (SWSDS), commissioned by the Waste Technology Development Department at the Idaho National Engineering Laboratory (INEL), examines relative life-cycle costs associated with three system concepts for processing the alpha low-level waste (alpha-LLW) stored at the Radioactive Waste Management Complex`s Transuranic Storage Area at the INEL. The three system concepts are incineration/melting; thermal treatment/solidification; and sort, treat, and repackage. The SWSDS identifies system functional and operational requirements and assesses implementability; effectiveness; cost; and demonstration, testing, and evaluation (DT&E) requirements for each of the three concepts.

  13. Alpha low-level stored waste systems design study

    SciTech Connect (OSTI)

    Feizollahi, F.; Teheranian, B. . Environmental Services Div.); Quapp, W.J. )

    1992-08-01

    The Stored Waste System Design Study (SWSDS), commissioned by the Waste Technology Development Department at the Idaho National Engineering Laboratory (INEL), examines relative life-cycle costs associated with three system concepts for processing the alpha low-level waste (alpha-LLW) stored at the Radioactive Waste Management Complex's Transuranic Storage Area at the INEL. The three system concepts are incineration/melting; thermal treatment/solidification; and sort, treat, and repackage. The SWSDS identifies system functional and operational requirements and assesses implementability; effectiveness; cost; and demonstration, testing, and evaluation (DT E) requirements for each of the three concepts.

  14. Seventh annual DOE LLWMP participants' information meeting. DOE Low-Level Waste Management Program. Abstracts

    SciTech Connect (OSTI)

    Not Available

    1985-08-01

    The following sessions were held: International Low-Level Waste Management Activities; Low-Level Waste Disposal; Characteristics and Treatment of Low-Level Waste; Environmental Monitoring and Performance; Greater Confinement and Alternative Disposal Methods; Low-Level Waste Management; Corrective Measures; Performance Prediction and Assessment; and Siting New Defense and Commercial Low-Level Waste Disposal Facilities.

  15. Development of Effective Solvent Modifiers for the Solvent Extraction of Cesium from Alkaline High-Level Tank Waste.

    SciTech Connect (OSTI)

    Bonnesen, Peter V.; Delmau, Laetitia H.; Moyer, Bruce A.; Lumetta, Gregg J. )

    2003-01-01

    A series of novel alkylphenoxy fluorinated alcohols were prepared and investigated for their effectiveness as modifiers in solvents containing calix[4]arene-bis-(tert-octylbenzo)-crown-6 for extracting cesium from alkaline nitrate media. A modifier that contained a terminal 1,1,2,2-tetrafluoroethoxy group was found to decompose following long-term exposure to warm alkaline solutions. However, replacement of the tetrafluoroethoxy group with a 2,2,3,3-tetrafluoropropoxy group led to a series of modifiers that possessed the alkaline stability required for a solvent extraction process. Within this series of modifiers, the structure of the alkyl substituent (tert-octyl, tert-butyl, tert-amyl, and sec-butyl) of the alkylphenoxy moiety was found to have a profound impact on the phase behavior of the solvent in liquid-liquid contacting experiments, and hence on the overall suitability of the modifier for a solvent extraction process. The sec-butyl derivative[1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol] (Cs-7SB) was found to possess the best overall balance of properties with respect to third phase and coalescence behavior, cleanup following degradation, resistance to solids formation, and cesium distribution behavior. Accordingly, this modifier was selected for use as a component of the solvent employed in the Caustic-Side Solvent Extraction (CSSX) process for removing cesium from high level nuclear waste (HLW) at the U.S. Department of Energy?s (DOE) Savannah River Site. In batch equilibrium experiments, this solvent has also been successfully shown to extract cesium from both simulated and actual solutions generated from caustic leaching of HLW tank sludge stored in tank B-110 at the DOE?s Hanford Site.

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

    SciTech Connect (OSTI)

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

    1996-08-01

    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.

  17. RETENTION OF SULFATE IN HIGH LEVEL RADIOACTIVE WASTE GLASS

    SciTech Connect (OSTI)

    Fox, K.

    2010-09-07

    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.

  18. Neptunium estimation in dissolver and high-level-waste solutions

    SciTech Connect (OSTI)

    Pathak, P.N.; Prabhu, D.R.; Kanekar, A.S.; Manchanda, V.K.

    2008-07-01

    This papers deals with the optimization of the experimental conditions for the estimation of {sup 237}Np in spent-fuel dissolver/high-level waste solutions using thenoyltrifluoroacetone as the extractant. (authors)

  19. Low-level waste vitrification plant environmental permitting plan

    SciTech Connect (OSTI)

    Gretsinger, W.T.; Colby, J.M.

    1994-10-03

    This document presents projected environmental permitting and approval requirements for the treatment and disposal of low-level Hanford tank waste by vitrification. Applicability, current status, and strategy are discussed for each potential environmental permit or approval.

  20. Bibliographic Data on Low-Level Radioactive Waste Documents

    Energy Science and Technology Software Center (OSTI)

    1995-11-10

    The purpose of the system is to allow users (researchers, policy makers, etc) to identify existing documents on a range of subjects related to low-level radioactive waste management. The software is menu driven.

  1. Hanford Tank Waste Residuals

    Office of Environmental Management (EM)

    Hanford Tank Waste Residuals DOE HLW Corporate Board November 6, 2008 Chris Kemp, DOE ORP Bill Hewitt, YAHSGS LLC Hanford Tanks & Tank Waste * Single-Shell Tanks (SSTs) - 27 million ...

  2. Characteristics of potential repository wastes. Volume 1

    SciTech Connect (OSTI)

    Not Available

    1992-07-01

    This document, and its associated appendices and microcomputer (PC) data bases, constitutes the reference OCRWM data base of physical and radiological characteristics data of radioactive wastes. This Characteristics Data Base (CDB) system includes data on spent nuclear fuel and high-level waste (HLW), which clearly require geologic disposal, and other wastes which may require long-term isolation, such as sealed radioisotope sources. The data base system was developed for OCRWM by the CDB Project at Oak Ridge National Laboratory. Various principal or official sources of these data provided primary information to the CDB Project which then used the ORIGEN2 computer code to calculate radiological properties. The data have been qualified by an OCRWM-sponsored peer review as suitable for quality-affecting work meeting the requirements of OCRWM`s Quality Assurance Program. The wastes characterized in this report include: light-water reactor (LWR) spent fuel and immobilized HLW.

  3. US DOE Initiated Performance Enhancements to the Hanford Waste Treatment and Immobilization Plant (WTP) Low-activity Waste Vitrification (LAW) System

    SciTech Connect (OSTI)

    Hamel, William F.; Gerdes, Kurt D.; Holton, Langdon K.; Pegg, Ian L.; Bowen, Brad W.

    2006-03-03

    The U.S Department of Energy Office of River Protection (DOE-ORP) is constructing a Waste Treatment and Immobilization Plant (WTP) for the treatment and vitrification of underground tank wastes stored at the Hanford Site in Washington State. The WTP comprises four major facilities: a pretreatment facility to separate the tank waste into high level waste (HLW) and low-activity waste (LAW) process streams, a HLW vitrification facility to immobilize the HLW fraction; a LAW vitrification facility to immobilize the LAW fraction, and an analytical laboratory to support the operations of all four treatment facilities. DOE has established strategic objectives to optimize the performance of the WTP facilities and the LAW and HLW waste forms to reduce the overall schedule and cost for treatment and vitrification of the Hanford tank wastes. This strategy has been implemented by establishing performance expectations in the WTP contract for the facilities and waste forms. In addition, DOE, as owner-operator of the WTP facilities, continues to evaluate 1) the design, to determine the potential for performance above the requirements specified in the WTP contract; and 2) improvements in production of the LAW and HLW waste forms. This paper reports recent progress directed at improving production of the LAW waste form. DOEs initial assessment, which is based on the work reported in this paper, is that the capacity of the WTP LAW vitrification facility can be increased by a factor of 2 to 4 with a combination of revised glass formulations, modest increases in melter glass operating temperatures, and a second-generation LAW melter with a larger surface area. Implementing these improvements in the LAW waste immobilization capability can benefit the LAW treatment mission by reducing both processing time and cost.

  4. Immobilized low-level waste disposal options configuration study

    SciTech Connect (OSTI)

    Mitchell, D.E.

    1995-02-01

    This report compiles information that supports the eventual conceptual and definitive design of a disposal facility for immobilized low-level waste. The report includes the results of a joint Westinghouse/Fluor Daniel Inc. evaluation of trade-offs for glass manufacturing and product (waste form) disposal. Though recommendations for the preferred manufacturing and disposal option for low-level waste are outside the scope of this document, relative ranking as applied to facility complexity, safety, remote operation concepts and ease of retrieval are addressed.

  5. Modeling and low-level waste management: an interagency workshop

    SciTech Connect (OSTI)

    Little, C.A.; Stratton, L.E.

    1980-01-01

    The interagency workshop on Modeling and Low-Level Waste Management was held on December 1-4, 1980 in Denver, Colorado. Twenty papers were presented at this meeting which consisted of three sessions. First, each agency presented its point of view concerning modeling and the need for models in low-level radioactive waste applications. Second, a larger group of more technical papers was presented by persons actively involved in model development or applications. Last of all, four workshops were held to attempt to reach a consensus among participants regarding numerous waste modeling topics. Abstracts are provided for the papers presented at this workshop.

  6. Low-level radioactive waste disposal facility closure

    SciTech Connect (OSTI)

    White, G.J.; Ferns, T.W.; Otis, M.D.; Marts, S.T.; DeHaan, M.S.; Schwaller, R.G.; White, G.J. )

    1990-11-01

    Part I of this report describes and evaluates potential impacts associated with changes in environmental conditions on a low-level radioactive waste disposal site over a long period of time. Ecological processes are discussed and baselines are established consistent with their potential for causing a significant impact to low-level radioactive waste facility. A variety of factors that might disrupt or act on long-term predictions are evaluated including biological, chemical, and physical phenomena of both natural and anthropogenic origin. These factors are then applied to six existing, yet very different, low-level radioactive waste sites. A summary and recommendations for future site characterization and monitoring activities is given for application to potential and existing sites. Part II of this report contains guidance on the design and implementation of a performance monitoring program for low-level radioactive waste disposal facilities. A monitoring programs is described that will assess whether engineered barriers surrounding the waste are effectively isolating the waste and will continue to isolate the waste by remaining structurally stable. Monitoring techniques and instruments are discussed relative to their ability to measure (a) parameters directly related to water movement though engineered barriers, (b) parameters directly related to the structural stability of engineered barriers, and (c) parameters that characterize external or internal conditions that may cause physical changes leading to enhanced water movement or compromises in stability. Data interpretation leading to decisions concerning facility closure is discussed. 120 refs., 12 figs., 17 tabs.

  7. SECONDARY WASTE MANAGEMENT FOR HANFORD EARLY LOW ACTIVITY WASTE VITRIFICATION

    SciTech Connect (OSTI)

    UNTERREINER BJ

    2008-07-18

    More than 200 million liters (53 million gallons) of highly radioactive and hazardous waste is stored at the U.S. Department of Energy's Hanford Site in southeastern Washington State. The DOE's Hanford Site River Protection Project (RPP) mission includes tank waste retrieval, waste treatment, waste disposal, and tank farms closure activities. This mission will largely be accomplished by the construction and operation of three large treatment facilities at the Waste Treatment and Immobilization Plant (WTP): (1) a Pretreatment (PT) facility intended to separate the tank waste into High Level Waste (HLW) and Low Activity Waste (LAW); (2) a HLW vitrification facility intended to immobilize the HLW for disposal at a geologic repository in Yucca Mountain; and (3) a LAW vitrification facility intended to immobilize the LAW for shallow land burial at Hanford's Integrated Disposal Facility (IDF). The LAW facility is on target to be completed in 2014, five years prior to the completion of the rest of the WTP. In order to gain experience in the operation of the LAW vitrification facility, accelerate retrieval from single-shell tank (SST) farms, and hasten the completion of the LAW immobilization, it has been proposed to begin treatment of the low-activity waste five years before the conclusion of the WTP's construction. A challenge with this strategy is that the stream containing the LAW vitrification facility off-gas treatment condensates will not have the option of recycling back to pretreatment, and will instead be treated by the Hanford Effluent Treatment Facility (ETF). Here the off-gas condensates will be immobilized into a secondary waste form; ETF solid waste.

  8. 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)

    2013-07-01

    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)

  9. Flammable gas tank waste level reconciliation for 241-SX-105

    SciTech Connect (OSTI)

    Brevick, C.H.; Gaddie, L.A.

    1997-06-23

    Fluor Daniel Northwest was authorized to address flammable gas issues by reconciling the unexplained surface level increases in Tank 241-SX-105 (SX-105, typical). The trapped gas evaluation document states that Tank SX-105 exceeds the 25% of the lower flammable limit criterion, based on a surface level rise evaluation. The Waste Storage Tank Status and Leak Detection Criteria document, commonly referred to as the Welty Report is the basis for this letter report. The Welty Report is also a part of the trapped gas evaluation document criteria. The Welty Report contains various tank information, including: physical information, status, levels, and dry wells. The unexplained waste level rises were attributed to the production and retention of gas in the column of waste corresponding to the unaccounted for surface level rise. From 1973 through 1980, the Welty Report tracked Tank SX-105 transfers and reported a net cumulative change of 20.75 in. This surface level increase is from an unknown source or is unaccounted for. Duke Engineering and Services Hanford and Lockheed Martin Hanford Corporation are interested in determining the validity of unexplained surface level changes reported in the Welty Report based upon other corroborative sources of data. The purpose of this letter report is to assemble detailed surface level and waste addition data from daily tank records, logbooks, and other corroborative data that indicate surface levels, and to reconcile the cumulative unaccounted for surface level changes as shown in the Welty Report from 1973 through 1980. Tank SX-105 initially received waste from REDOX starting the second quarter of 1955. After June 1975, the tank primarily received processed waste (slurry) from the 242-S Evaporator/Crystallizer and transferred supernate waste to Tanks S-102 and SX-102. The Welty Report shows a cumulative change of 20.75 in. from June 1973 through December 1980.

  10. End of Year 2010 SNF & HLW Inventories | Department of Energy

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

    End of Year 2010 SNF & HLW Inventories End of Year 2010 SNF & HLW Inventories Map of the United States of America that shows the location of approximately 64,000 MTHM of Spent ...

  11. Disposal of Low-Level Waste at the Nevada National Security Site...

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

    of Low-Level Waste at the Nevada National Security Site Disposal of Low-Level Waste at the Nevada National Security Site Disposal of Low-Level Waste at the Nevada National Security ...

  12. Low-Level Waste Disposal Alternatives Analysis Report

    SciTech Connect (OSTI)

    Timothy Carlson; Kay Adler-Flitton; Roy Grant; Joan Connolly; Peggy Hinman; Charles Marcinkiewicz

    2006-09-01

    This report identifies and compares on-site and off-site disposal options for the disposal of contract-handled and remote-handled low-level waste generated by the Idaho National Laboratory and its tenants. Potential disposal options are screened for viability by waste type resulting in a short list of options for further consideration. The most crediable option are selected after systematic consideration of cost, schedule constraints, and risk. In order to holistically address the approach for low-level waste disposal, options are compiled into comprehensive disposal schemes, that is, alternative scenarios. Each alternative scenario addresses the disposal path for all low-level waste types over the period of interest. The alternative scenarios are compared and ranked using cost, risk and complexity to arrive at the recommended approach. Schedule alignment with disposal needs is addressed to ensure that all waste types are managed appropriately. The recommended alternative scenario for the disposal of low-level waste based on this analysis is to build a disposal facility at the Idaho National Laboratory Site.

  13. Disposal of low-level and low-level mixed waste: audit report

    SciTech Connect (OSTI)

    1998-09-03

    The Department of Energy (Department) is faced with the legacy of thousands of contaminated areas and buildings and large volumes of `backlog` waste requiring disposal. Waste management and environmental restoration activities have become central to the Department`s mission. One of the Department`s priorities is to clean up former nuclear weapons sites and find more effective and timely methods for disposing of nuclear waste. This audit focused on determining if the Department was disposing of low-level and low-level mixed waste in the most cost-effective manner.

  14. Radionuclide release from low-level waste in field lysimeters

    SciTech Connect (OSTI)

    Oblath, S B

    1986-01-01

    A field program has been in operation for 8 years at the Savannah River Plant (SRP) to determine the leaching/migration behavior of low-level radioactive waste using lysimeters. The lysimeters are soil-filled caissons containing well characterized wastes, with each lysimeter serving as a model of a shallow land burial trench. Sampling and analysis of percolate water and vegetation from the lysimeters provide a determination of the release rates of the radionuclides from the waste/soil system. Vegetative uptake appears to be a major pathway for migration. Fractional release rates from the waste/soil system are less than 0.01% per year. Waste-to-soil leach rates up to 10% per year have been determined by coring several of the lysimeters. The leaching of solidified wasteforms under unsaturated field conditions has agreed well with static, immersion leaching of the same type waste in the laboratory. However, releases from the waste/soil system in the lysimeter may be greater than predicted based on leaching alone, due to complexation of the radionuclides by other components leached from the wastes to form mobile, anionic species.

  15. DOE/EIS-0287 Idaho High-Level Waste & Facilities Disposition...

    Office of Environmental Management (EM)

    State of Idaho Title: Idaho High-Level Waste and Facilities Disposition Draft ... or call: Abstract: This Idaho High-Level Waste and Facilities Disposition Draft EIS ...

  16. DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel

    Office of Environmental Management (EM)

    Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent ... level radioactive waste and spent nuclear fuel in a single repository or repositories. ...

  17. Introduction to DOE Order 435.1 Low Level Radioactive Waste Disposal...

    Office of Environmental Management (EM)

    Introduction to DOE Order 435.1 Low Level Radioactive Waste Disposal Requirements Introduction to DOE Order 435.1 Low Level Radioactive Waste Disposal Requirements Christine ...

  18. Corrosion and failure processes in high-level waste tanks

    SciTech Connect (OSTI)

    Mahidhara, R.K.; Elleman, T.S.; Murty, K.L.

    1992-11-01

    A large amount of radioactive waste has been stored safely at the Savannah River and Hanford sites over the past 46 years. The aim of this report is to review the experimental corrosion studies at Savannah River and Hanford with the intention of identifying the types and rates of corrosion encountered and indicate how these data contribute to tank failure predictions. The compositions of the High-Level Wastes, mild steels used in the construction of the waste tanks and degradation-modes particularly stress corrosion cracking and pitting are discussed. Current concerns at the Hanford Site are highlighted.

  19. Evaluation of solid-based separation materials for the pretreatment of radioactive wastes

    SciTech Connect (OSTI)

    Lumetta, G.J.; Wagner, M.J.; Wester, D.W.; Morrey, J.R.

    1993-05-01

    Separation science will play an important role in pretreating nuclear wastes stored at various US Department of Energy Sites. The application of separation processes offers potential economic and environmental benefits with regards to remediating these sites. For example, at the Hanford Site, the sizeable volume of radioactive wastes stored in underground tanks could be partitioned into a small volume of high-level waste (HLW) and a relatively large volume of low-level waste (LLW). After waste separation, only the smaller volume of HLW would require costly vitrification and geologic disposal. Furthermore, the quality of the remaining LLW form (e.g., grout) would be improved due to the lower inventory of radionuclides present in the LLW stream. This report investigates extraction chromatography as a possible separation process for Hanford wastes.

  20. B Plant low level waste system integrity assessment report

    SciTech Connect (OSTI)

    Walter, E.J.

    1995-09-01

    This document provides the report of the integrity assessment activities for the B Plant low level waste system. The assessment activities were in response to requirements of the Washington State Dangerous Waste Regulations, Washington Administrative Code (WAC), 173-303-640. This integrity assessment report supports compliance with Hanford Federal Facility Agreement and Consent Order interim milestone target action M-32-07-T03.

  1. Chemical digestion of low level nuclear solid waste material

    DOE Patents [OSTI]

    Cooley, Carl R.; Lerch, Ronald E.

    1976-01-01

    A chemical digestion for treatment of low level combustible nuclear solid waste material is provided and comprises reacting the solid waste material with concentrated sulfuric acid at a temperature within the range of 230.degree.-300.degree.C and simultaneously and/or thereafter contacting the reacting mixture with concentrated nitric acid or nitrogen dioxide. In a special embodiment spent ion exchange resins are converted by this chemical digestion to noncombustible gases and a low volume noncombustible residue.

  2. Technical area status report for low-level mixed waste final waste forms. Volume 2, Appendices

    SciTech Connect (OSTI)

    Mayberry, J.L.; Huebner, T.L.; Ross, W.; Nakaoka, R.; Schumacher, R.; Cunnane, J.; Singh, D.; Darnell, R.; Greenhalgh, W.

    1993-08-01

    This report presents information on low-level mixed waste forms.The descriptions of the low-level mixed waste (LLMW) streams that are considered by the Mixed Waste Integrated Program (MWIP) are given in Appendix A. This information was taken from descriptions generated by the Mixed Waste Treatment Program (MWTP). Appendix B provides a list of characteristic properties initially considered by the Final Waste Form (FWF) Working Group (WG). A description of facilities available to test the various FWFs discussed in Volume I of DOE/MWIP-3 are given in Appendix C. Appendix D provides a summary of numerous articles that were reviewed on testing of FWFS. Information that was collected by the tests on the characteristic properties considered in this report are documented in Appendix D. The articles reviewed are not a comprehensive list, but are provided to give an indication of the data that are available.

  3. Low-level radioactive waste technology: a selected, annotated bibliography

    SciTech Connect (OSTI)

    Fore, C.S.; Vaughan, N.D.; Hyder, L.K.

    1980-10-01

    This annotated bibliography of 447 references contains scientific, technical, economic, and regulatory information relevant to low-level radioactive waste technology. The bibliography focuses on environmental transport, disposal site, and waste treatment studies. The publication covers both domestic and foreign literature for the period 1952 to 1979. Major chapters selected are Chemical and Physical Aspects; Container Design and Performance; Disposal Site; Environmental Transport; General Studies and Reviews; Geology, Hydrology and Site Resources; Regulatory and Economic Aspects; Transportation Technology; Waste Production; and Waste Treatment. Specialized data fields have been incorporated into the data file to improve the ease and accuracy of locating pertinent references. Specific radionuclides for which data are presented are listed in the Measured Radionuclides field, and specific parameters which affect the migration of these radionuclides are presented in the Measured Parameters field. In addition, each document referenced in this bibliography has been assigned a relevance number to facilitate sorting the documents according to their pertinence to low-level radioactive waste technology. The documents are rated 1, 2, 3, or 4, with 1 indicating direct applicability to low-level radioactive waste technology and 4 indicating that a considerable amount of interpretation is required for the information presented to be applied. The references within each chapter are arranged alphabetically by leading author, corporate affiliation, or title of the document. Indexes are provide for (1) author(s), (2) keywords, (3) subject category, (4) title, (5) geographic location, (6) measured parameters, (7) measured radionuclides, and (8) publication description.

  4. Mixed and Low-Level Waste Treatment Facility Project

    SciTech Connect (OSTI)

    Not Available

    1992-04-01

    Mixed and low-level wastes generated at the Idaho National Engineering Laboratory (INEL) are required to be managed according to applicable State and Federal regulations, and Department of Energy Orders that provide for the protection of human health and the environment. The Mixed and Low-Level Waste Treatment Facility Project was chartered in 1991, by the Department of Energy to provide treatment capability for these mixed and low-level waste streams. The first project task consisted of conducting engineering studies to identify the waste streams, their potential treatment strategies, and the requirements that would be imposed on the waste streams and the facilities used to process them. This report documents those studies so the project can continue with an evaluation of programmatic options, system tradeoff studies, and the conceptual design phase of the project. This report, appendix B, comprises the engineering design files for this project study. The engineering design files document each waste steam, its characteristics, and identified treatment strategies.

  5. Mixed and Low-Level Waste Treatment Facility project

    SciTech Connect (OSTI)

    Not Available

    1992-04-01

    Mixed and low-level wastes generated at the Idaho National Engineering Laboratory (INEL) are required to be managed according to applicable State and Federal regulations, and Department of Energy Orders that provide for the protection of human health and the environment. The Mixed and Low-Level Waste Treatment Facility Project was chartered in 1991, by the Department of Energy to provide treatment capability for these mixed and low-level waste streams. The first project task consisted of conducting engineering studies to identify the waste streams, their potential treatment strategies, and the requirements that would be imposed on the waste streams and the facilities used to process them. This report, Appendix A, Environmental Regulatory Planning Documentation, identifies the regulatory requirements that would be imposed on the operation or construction of a facility designed to process the INEL's waste streams. These requirements are contained in five reports that discuss the following topics: (1) an environmental compliance plan and schedule, (2) National Environmental Policy Act requirements, (3) preliminary siting requirements, (4) regulatory justification for the project, and (5) health and safety criteria.

  6. Waste Management Facilities cost information for mixed low-level waste. Revision 1

    SciTech Connect (OSTI)

    Shropshire, D.; Sherick, M.; Biadgi, C.

    1995-06-01

    This report contains preconceptual designs and planning level life-cycle cost estimates for managing mixed low-level waste. The report`s information on treatment, storage, and disposal modules can be integrated to develop total life-cycle costs for various waste management options. A procedure to guide the US Department of Energy and its contractor personnel in the use of cost estimation data is also summarized in this report.

  7. Interim report: Waste management facilities cost information for mixed low-level waste

    SciTech Connect (OSTI)

    Feizollahi, F.; Shropshire, D.

    1994-03-01

    This report contains preconceptual designs and planning level life-cycle cost estimates for treating alpha and nonalpha mixed low-level radioactive waste. This report contains information on twenty-seven treatment, storage, and disposal modules that can be integrated to develop total life cycle costs for various waste management options. A procedure to guide the US Department of Energy and its contractor personnel in the use of estimating data is also summarized in this report.

  8. Cementitious waste option scoping study report

    SciTech Connect (OSTI)

    Lee, A.E.; Taylor, D.D.

    1998-02-01

    A Settlement Agreement between the Department of Energy (DOE) and the State of Idaho mandates that all high-level radioactive waste (HLW) now stored at the Idaho Chemical Processing Plant (ICPP) on the Idaho National Engineering and Environmental Laboratory (INEEL) will be treated so that it is ready to be moved out of Idaho for disposal by a target date of 2035. This study investigates the nonseparations Cementitious Waste Option (CWO) as a means to achieve this goal. Under this option all liquid sodium-bearing waste (SBW) and existing HLW calcine would be recalcined with sucrose, grouted, canisterized, and interim stored as a mixed-HLW for eventual preparation and shipment off-Site for disposal. The CWO waste would be transported to a Greater Confinement Disposal Facility (GCDF) located in the southwestern desert of the US on the Nevada Test Site (NTS). All transport preparation, shipment, and disposal facility activities are beyond the scope of this study. CWO waste processing, packaging, and interim storage would occur over a 5-year period between 2013 and 2017. Waste transport and disposal would occur during the same time period.

  9. Remote-Handled Low Level Waste Disposal Project Alternatives Analysis

    SciTech Connect (OSTI)

    David Duncan

    2010-10-01

    This report identifies, evaluates, and compares alternatives for meeting the U.S. Department of Energy’s mission need for management of remote-handled low-level waste generated by the Idaho National Laboratory and its tenants. Each alternative identified in the Mission Need Statement for the Remote-Handled Low-Level Waste Treatment Project is described and evaluated for capability to fulfill the mission need. Alternatives that could meet the mission need are further evaluated and compared using criteria of cost, risk, complexity, stakeholder values, and regulatory compliance. The alternative for disposal of remote-handled low-level waste that has the highest confidence of meeting the mission need and represents best value to the government is to build a new disposal facility at the Idaho National Laboratory Site.

  10. Perspectives of Future R and D on HLW Disposal in Germany

    SciTech Connect (OSTI)

    Steininger, W.J. [Forschungszentrum Karlsruhe GmbH, Project Management Agency Forschungszentrum Karlsruhe (PTKA-WTE), Eggenstein-Leopoldshafen (Germany)

    2008-07-01

    The 5. Energy Research Program of the Federal Government 'Innovation and New Technology' is the general framework for R and D activities in radioactive waste disposal. The Ministry of Economics and Technology (BMWi), the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) and the Ministry of Education and Research (BMBF) apply the Research Program concerning their respective responsibilities and competences. With regard to the Government's obligation to provide repositories for HLW (spent fuel and vitrified HAW) radioactive waste basic and applied R and D is needed in order to make adequate knowledge available to implementers, decision makers and stakeholders in general. Non-site specific R and D projects are funded by BMWi on the basis of its Research Concept. In the first stage (1998 -2001) most R and D issues were focused on R and D activities related to HLW disposal in rock salt. By that time the R and D program had to be revised and some prioritization was demanded due to changes in politics. In the current version (2001 -2006) emphasize was put on non-saline rocks. The current Research Concept of BMWi is presently subjected to a sort of revision, evaluation, and discussion, inter alia, by experts from several German research institutions. This activity is of special importance against the background of streamlining and focusing the research activities to future demands, priorities and perspectives with regard to the salt concept and the option of disposing of HLW in argillaceous media. Because the status of knowledge on disposal in rock salt is well advanced, it is necessary to take stock of the current state-of-the-art. In this framework some key projects are being currently carried out. The results may contribute to future decisions to be made in Germany with respect to HLW disposal. The first project deals with the development of an advanced safety concept for a HLW waste repository in rock salt. The second project (also