National Library of Energy BETA

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

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

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

  3. 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 ...

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

    Energy Savers [EERE]

    nuclear fuel (SNF) | Department of Energy Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) 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 to generate nuclear energy but that has been removed from the reactor as no

  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. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat transfer and glass melting rate. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of {approx}1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150 C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage. The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP Contract terms. The WTP's overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in wasteloading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.

  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 observed in any of the pour stream glass samples. Spinel was observed at the bottom of DWPF Melter 1 as a result of K-3 refractory corrosion. Issues have occurred with accumulation of spinel in the pour spout during periods of operation at higher waste loadings. Given that both DWPF melters were or have been in operation for greater than 8 years, the service life of the melters has far exceeded design expectations. It is possible that the DWPF liquidus temperature approach is conservative, in that it may be possible to successfully operate the melter with a small degree of allowable crystallization in the glass. This could be a viable approach to increasing waste loading in the glass assuming that the crystals are suspended in the melt and swept out through the riser and pour spout. Additional study is needed, and development work for WTP might be leveraged to support a different operating limit for the DWPF. Several recommendations are made regarding considerations that need to be included as part of the WTP crystal tolerant strategy based on the DWPF development work and operational data reviewed here. These include: Identify and consider the impacts of potential heat sinks in the WTP melter and glass pouring system; Consider the contributions of refractory corrosion products, which may serve to nucleate additional crystals leading to further accumulation; Consider volatilization of components from the melt (e.g., boron, alkali, halides, etc.) and determine their impacts on glass crystallization behavior; Evaluate the impacts of glass REDuction/OXidation (REDOX) conditions and the distribution of temperature within the WTP melt pool and melter pour chamber on crystal accumulation rate; Consider the impact of precipitated crystals on glass viscosity; Consider the impact of an accumulated crystalline layer on thermal convection currents and bubbler effectiveness within the melt pool; Evaluate the impact of spinel accumulation on Joule heating of the WTP melt pool; and Include noble metals in glass melt experiments because of their potential to act as nucleation site

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

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

    Department of Energy HLW Waste Vitrification Facility PDF icon Summary - WTP HLW Waste Vitrification Facility More Documents & Publications Waste Treatment and Immobilization Plant (WTP) Analytical Laboratory (LAB), Balance of Facilities (BOF) and Low-Activity Waste Vitrification Facilities (LAW) Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant - November 2013 Waste Treatment and Immobilation Plant Pretreatment Facility

  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{sub 2}O{sub 3} species to MnO during melter preprocessing. At the lower redox limit of Fe{sup +2}/{summation}Fe {approx} 0.09 about 99% of the Mn{sup +4}/Mn{sup +3} is converted to Mn{sup +2}. Therefore, the lower REDOX limits eliminates melter foaming from deoxygenation.

  13. 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.

  14. 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.

  15. 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...

  16. 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.

  17. 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)

  18. 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.

  19. 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 glass melting rate. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of {approx}1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150 C and by increasing the waste loading in the glass product Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage. The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP contract requirements. The WTP's overall mission will require the immobilization oftank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in waste-loading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.

  20. 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...

  1. 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.

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

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

    ... ProdMod tracks three key waste constituents: 1) sodium, ... HLW System will operate as fast as available funding and ... Restoration and Reactors are also treated.After ...

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

    Office of Environmental Management (EM)

    on Separate Disposal of Defense High- Level 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) resulting from atomic energy defense activities ("Defense HLW Repository") is "required" within the meaning of Section 8(b)(2) of the Nuclear Waste Policy Act of 1982 (NWPA). In 1985, the U.S. Department of Energy (DOE) and President Reagan considered this

  4. 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)

  5. Hight-Level Waste & Facilities Disposition

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

    High-Level Waste (HLW) and Facilities Disposition Final Environmental Impact Statement You are here: DOE-ID Home > Environmental Management > Idaho High-Level Waste (HLW) Table of Contents Documents are in the Adobe� PDF format and require the Adobe� Reader to access them. If you do not currently have the Acrobat Reader, you can download the Free Adobe Reader at http://get.adobe.com/reader/ Icon link to Free Adobe Acrobat Reader software * Large chapters broken down into sections

  6. 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...

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

    Office of Environmental Management (EM)

    Liquid Waste Tank Integrity Workshop - 2008 Karthik Subramanian Bruce Wiersma November 2008 High Level Waste Corporate Board Meeting karthik.subramanian@srnl.doe.gov bruce.wiersma@srnl.doe.gov 2 Acknowledgements * Bruce Wiersma (SRNL) * Kayle Boomer (Hanford) * Michael T. Terry (Facilitator) * SRS - Liquid Waste Organization * Hanford Tank Farms * DOE-EM 3 Background * High level radioactive waste (HLW) tanks provide critical interim confinement for waste prior to processing and permanent

  8. 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 will be selected for the disposal container inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and natural barrier, will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel and the outer cylinder and outer cylinder lids will be a barrier made of high-nickel alloy. The defense HLW disposal container interfaces with the emplacement drift environment and the internal waste by transferring heat from the canisters to the external environment and by protecting the canisters and their contents from damage/degradation by the external environment. The disposal container also interfaces with the canisters by limiting access of moderator and oxidizing agents to the waste. A loaded and sealed disposal container (waste package) interfaces with the Emplacement Drift System's emplacement drift waste package supports upon which the waste packages are placed. The disposal container interfaces with the Canister Transfer System, Waste Emplacement /Retrieval System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement, and retrieval for the disposal container/waste package.

  9. 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.

  10. 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.

  11. Enterprise Assessments Operational Awareness Record, Waste Treatment...

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

    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))...

  12. 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.

  13. 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.

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

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

    ... SZ QQ System PlanFY96 FYP 38.9 43.2 47.1 63.4 62.2 66.1 WRP Baseline 5a .6.Q..2 ... The WSRC proposal has been developed and accepted by DOE-SA. A matrixed organization has ...

  15. 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

  16. 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.

  17. High-Level Waste Corporate Board, Mark Gilbertson

    Office of Environmental Management (EM)

    High-Level Waste Corporate Board April 1, 2008 safety v performance v cleanup v closure M E Environmental Management Environmental Management Path Forward * Initial meeting lays groundwork for future meetings * There are enough issues to manage that meetings will be more frequent than semiannual * HLW Issues Board will address include: - Within Site - Between Sites - Affecting Entire Complex

  18. 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...

  19. 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.

  20. 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 formulationsmore »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.« less

  1. 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.

  2. High-Level Waste Corporate Board Performance Assessment Subcommittee

    Office of Environmental Management (EM)

    Level Waste Corporate Board Performance Assessment Subcommittee John E. Marra, Ph.D. Associate Laboratory Director November 6, 2008 Richland, WA DOE-EM HLW Corporate Board Meeting Background - Performance Assessment Process Performance assessments are the fundamental risk assessment tool used by the DOE to evaluate and communicate the effectiveness and long-term impact of waste management and cleanup decisions. This includes demonstrations of compliance, NEPA analyses, and decisions about

  3. 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.

  4. 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)

  5. 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.

  6. 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 numbers and interim storage facilities available until the repository is closed. (authors)

  7. 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.

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

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

    for existing subsurface contamination The Oregon proposal Regulatory compliance Climate change Secondary-waste-form performance High-level radioactive waste (HLW)...

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

    Office of Environmental Management (EM)

    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 Nuclear Fuel (SNF) & 275 High-Level Radioactive Waste (HLW) Canisters. PDF icon Slide 1 More Documents & Publications Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel FY 2007 Total System Life Cycle Cost, Pub 2008 Microsoft Word - TSLCC 2007_5_05_08 rev

  10. 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.

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

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

    in Salt | Department of Energy Coupled Model for Heat and Water Transport in a High Level Waste Repository in Salt Coupled Model for Heat and Water Transport in a High Level Waste Repository in Salt This report summarizes efforts to simulate coupled thermal-hydrological-chemical (THC) processes occurring within a generic hypothetical high-level waste (HLW) repository in bedded salt; chemical processes of the system allow precipitation and dissolution of salt with elevated temperatures that

  12. 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 the achievements of this program with emphasis on the recent enhancements in Al{sub 2}O{sub 3} loadings in HLW glass and its processing characteristics. Glass formulation development included crucible-scale preparation and characterization of glass samples to assess compliance with all melt processing and product quality requirements, followed by small-scale screening tests to estimate processing rates. These results were used to down-select formulations for subsequent engineering-scale melter testing. Finally, further testing was performed on the DM1200 vitrification system installed at VSL, which is a one-third scale (1.20 m{sup 2}) pilot melter for the WTP HLW melters and which is fitted with a fully prototypical off-gas treatment system. These tests employed glass formulations with high waste loadings and Al{sub 2}O{sub 3} contents of {approx}25 wt%, which represents a near-doubling of the present WTP baseline maximum Al{sub 2}O{sub 3} loading. In addition, these formulations were processed successfully at glass production rates that exceeded the present requirements for WTP HLW vitrification by up to 88%. The higher aluminum loading in the HLW glass has an added benefit in that the aluminum leaching requirements in pretreatment are reduced, thus allowing less sodium addition in pretreatment, which in turn reduces the amount of LAW glass to be produced at the WTP. The impact of the results from this ORP program in reducing the overall cost and schedule for the Hanford waste treatment mission will be discussed.

  13. 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.

  14. 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 during cooling and crystals may be prone to dissolution. By designing a glass-ceramics, the risks of deleterious effects from devitrification are removed. Furthermore, glass-ceramics have higher mechanical strength and impact strengths and possess greater chemical durability as noted above. Glass-ceramics should provide a waste form with the advantages of glass - ease of manufacture - with improved mechanical properties, thermal stability, and chemical durability. This report will cover aspects relevant for the validation of the CCIM use in the production of glass-ceramic waste forms.

  15. 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 the as poured state and after being slowly cooled according to the canister centerline cooling (CCC) profile. Glass formulation development was also completed on other Hanford tank wastes that were identified to further challenge waste loading due to the presence of appreciable quantities (>750 g) of plutonium in the waste tanks. In addition to containing appreciable Pu quantities, the C-102 waste tank and the 244-TX waste tank contain high concentrations of aluminum and iron, respectively that will further challenge vitrification processing. Glass formulation testing also demonstrated that high waste loadings could be achieved with these tank compositions using the attributes afforded by the CCIM technology.

  16. 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.

  17. 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.

  18. 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.

  19. West Valley Demonstration Project Prepares to Relocate High-Level Waste

    Broader source: Energy.gov [DOE]

    WEST VALLEY, N.Y. – With completion of the recently constructed interim storage pad at the West Valley Demonstration Project (WVDP), EM crews have begun a multi-year effort to relocate high-level waste (HLW) canisters stored inside the Main Plant Process Building to prepare the facility for demolition.

  20. 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/NaAlSiO?) 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)?O?), 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.

  1. 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.

  2. Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste

    Energy Savers [EERE]

    and Spent Nuclear Fuel | Department of Energy Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel 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

  3. Locations of Spent Nuclear Fuel and High-Level Radioactive Waste |

    Office of Environmental Management (EM)

    Department of Energy Locations of Spent Nuclear Fuel and High-Level Radioactive Waste Locations of Spent Nuclear Fuel and High-Level Radioactive Waste Map of the United States of America showing the locations of spent nuclear fuel and high-level radioactive waste. PDF icon Slide 1 More Documents & Publications Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel End of Year 2010 SNF & HLW Inventories Evaluation of Options for Permanent

  4. 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.

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

    SciTech Connect (OSTI)

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

    2013-07-01

    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)

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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 combined into three batches for a total of about 3.2 million gallons of liquid waste. The chemical and radiological composition of these batches was estimated from the SpaceMan Plus{trademark} model using the same data set and assumptions as the baseline plans.

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

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

    ... evolution of Site Decontamination & Decommissioning initiatives, and other factors ... The Committee meets periodically to formally review the status and operational plan for ...

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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 support Site Recommendation reports and to assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the Development Plan ''Design Analysis for the Defense High-Level Waste Disposal Container'' (CRWMS M&O 2000c) with no deviations from the plan.

  19. 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.

  20. 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.

  1. 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.

  2. 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 representative WTP HLW and LAW glasses over a wide range of temperatures, from the melter operating temperature to the glass transition.

  3. 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.

  4. 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.

  5. 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 * ...

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Enterprise Assessments Operational Awareness Record, Waste Treatment...

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

    Assessments Operational Awareness Record, Waste Treatment and Immobilization Plant - December 2014 Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility

  13. 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.

  14. 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.

  15. 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 selected glass, HAL-17 with 26 wt% Al{sub 2}O{sub 3}, had viscosity and electrical conductivity within the boundaries for adequate processing in the Joule heated melters operated at 1150 deg. C. This HAL-17 glass was successfully processed using small-scale (SMK) and larger scale (EP-5) melters. There was no indication of spinel settling during processing. The product glass samples from these melter tests contained 1 to 4 vol% spinel crystals that are likely formed during cooling. The PCT tests on the product glasses are underway. The present study demonstrated that it is possible to formulate the glasses with up to 26 wt% Al{sub 2}O{sub 3} that satisfy the property requirements and is processable with Joule-heated melters operated at 1150 deg. C. The 'nepheline discriminator' for HAL-17 glass is 0.45, which supports that claim that the current rule ('nepheline discriminator' < 0.62) is too restrictive. Considering that the cost of HLW treatment is highly dependent on loading of waste in glass, this result provides a potential for significant cost saving for Hanford. The maximum Al{sub 2}O{sub 3} loading that can be achieved will also depend on concentrations of other components in wastes. For example, the loading of waste used in this study was also limited by the spinel crystallization after 950 deg. C 24 hr heat treatment, which suggests that the concentrations of spinel-forming components such as Fe{sub 2}O{sub 3}, Cr{sub 2}O{sub 3}, NiO, ZnO, and MnO would be critical in addition to Al{sub 2}O{sub 3} for the maximum Al{sub 2}O{sub 3} loading achievable. The observed glass production rate per unit melter surface area of 0.75 MT/(d.m{sup 2}) for SMK test is comparable to the design capacity of WTP HLW melters at 0.8 MT/(d.m{sup 2}). However, the test with EP-5 melter achieved 0.38 MT/(d.m{sup 2}), which is roughly a half of the WTP design capacity. This result may imply that the glass with 26 wt% Al{sub 2}O{sub 3} may not achieve the WTP design production rate. However, this hypothesis is not conclusive because of unknown effects of melter size and operation

  16. 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 PDF icon West Valley Demonstration Project Low-Level Waste Shipment More Documents &...

  17. Low-Level Waste Disposal Facility Federal Review Group (LFRG...

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

    Program Management Compliance Low-Level Waste Disposal Facility Federal Review Group (LFRG) Low-Level Waste Disposal Facility Federal Review Group (LFRG) The Low-Level Waste...

  18. 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)

  19. 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.

  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 technology, and its limitations to aid in the benefit-cost analysis for management decision whether to deploy the technology or to abandon the pipeline as has been done in the past. In conclusion: The ultimate objective of this study is to qualify NuVision's unplugging technology for use at Hanford. Experimental testing has been conducted using three pipeline lengths and three types of blockages. Erosion rates have been obtained and pressure data is being analyzed. An amplification of the inlet pressure has been observed along the pipeline and is the key to determining up to what pipe lengths the technology can be used without surpassing the site pressure limit. In addition, we will attempt to establish what the expected unplugging rates will be at the longer pipe lengths for each of the three blockages tested. Detailed information resulting from the testing will provide the DOE end-user with sufficient data and understanding of the technology, and its limitations so that management decisions can be made whether the technology has a reasonable chance to successfully unplug a pipeline, such as a cross site transfer line or process transfer pipeline at the Waste Treatment Plant. (authors)

  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. 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 ...

  3. 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.

  4. 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.%.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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''.

  13. 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.

  14. 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.

  15. 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)

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

    Energy Savers [EERE]

    High Level Waste Corporate Board Charter High Level Waste Corporate Board Charter PDF icon High Level Waste Corporate Board Charter More Documents & Publications Corporate Board By-Laws NNMCAB Bylaws ORSSAB Bylaws

  17. 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)

  18. 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 ...

  19. 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.

  20. 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.

  1. 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. [PaR Systems, Inc., Shoreview, MN (United States); Darwen, N.J. [Bechtel National, Inc., Richland, WA (United States)

    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 functional and timing studies throughout the design process. Since no humans can go in or out of the cell, there are several recovery options that have been designed into the system including jack-down wheels for the bridge and trolley, recovery drums for the manipulator hoist, and a wire rope cable cutter for the slewer jib hoist. If the entire crane fails in cell, the large diameter cable reel that provides power, signal, and control to the crane can be used to retrieve the crane from the cell into the crane maintenance area. (authors)

  2. 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.

  3. 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.

  4. 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 and sludge processing life cycle, thereby reducing the Defense Waste Processing Facility (DWPF) mission by 7 years. Additionally at the Hanford site, problematic waste streams, such as high boehmite and phosphate wastes, could be treated prior to receipt by WTP and thus dramatically improve the capacity of the facility to process HLW. Treatment of boehmite by continuous sludge leaching (CSL) before receipt by WTP will dramatically reduce the process cycle time for the WTP pretreatment facility, while treatment of phosphate will significantly reduce the number of HLW borosilicate glass canisters produced at the WTP. These and other promising technologies will be discussed.

  5. 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.

  6. 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.

  7. 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 cracking of Alloy 22 and titanium alloys. The effects of microbial activity and radiation on degradation of Alloy 22 and titanium alloys are also discussed. Further, for titanium alloys, the effects of fluorides, bromides, calcium ions, and galvanic coupling to less noble metals are further considered. It is concluded that, as far as materials degradation is concerned, the materials and design adopted in the U.S. Yucca Mountain Project will provide sufficient safety margins within the 10,000-years regulatory period.

  8. 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.

  9. 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.

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

    Office of Environmental Management (EM)

  11. 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.

  12. 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.

  13. 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 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. Department of Energy's Office of Enterprise Assessments, reviewed the hazard analyses for the radioactive liquid waste disposal system of the Waste Treatment and Immobilization Plant

  14. 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...

  15. 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 composition and temperature, will evolve as additional data on crystal accumulation are gathered. Model validation steps will be included to guide the development process and ensure the value of the effort (i.e., increased waste loading and waste throughput). A summary of the stages of the road map for developing the crystal-tolerant glass approach, their estimated durations, and deliverables is provided.

  16. High-Level Waste Corporate Board Presentation Archive | Department of

    Energy Savers [EERE]

    Energy High-Level Waste Corporate Board Presentation Archive High-Level Waste Corporate Board Presentation Archive Archived Documents PDF icon High-Level Waste Corporate Board, Dr. Inés Triay PDF icon High-Level Waste Corporate Board, Mark Gilbertson PDF icon EM Engineering & Technology Roadmap and Major Technology Demonstrations PDF icon Office of River Protection PDF icon Idaho National Laboratory Description, Chellenges, Technology, Issues, and Needs PDF icon West Valley

  17. Northeast High-Level Radioactive Waste Transportation Task Force Agenda |

    Office of Environmental Management (EM)

    Department of Energy Northeast High-Level Radioactive Waste Transportation Task Force Agenda Northeast High-Level Radioactive Waste Transportation Task Force Agenda PDF icon Northeast High-Level Radioactive Waste Transportation Task Force Agenda More Documents & Publications NTSF Spring 2013 Save The Date NTSF 2013 Agenda NTSF Spring 2010 Final Agenda

  18. 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.

  19. 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.

  20. 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)

  1. 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.

  2. West Valley Demonstration Project High-Level Waste Management

    Office of Environmental Management (EM)

    DRAFT_19507_1 High-Level Waste Management Bryan Bower, DOE Director - WVDP DOE High-Level Waste Corporate Board Meeting Savannah River Site April 1, 2008 West Valley Demonstration Project West Valley Demonstration Project DRAFT_19507_2 West Valley High-Level Waste To solidify the radioactive material from approximately 600,000 gallons of high-level radioactive waste into a durable, high-quality glass, both a pretreatment system to remove salts and sulfates from the waste and a vitrification

  3. 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 (Matyáš 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.

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

    Broader source: Energy.gov [DOE]

    This Revision 3 of the Low-Level Waste Disposal  Facility Federal Review Group (LFRG) Manual was prepared primarily to include review criteria for the review of transuranic (TRU) waste disposal...

  5. 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.

  6. 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.

  7. 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.

  8. Summary - Flowsheet for the Hanford Waste Treatment Plant

    Office of Environmental Management (EM)

    Waste Treatment Plant ETR Report Date: March 2006 ETR-1 United States Department of Energy Office of Environmental Management (DOE-EM) External Technical Review of the Flowsheet for the Hanford Waste Treatment Plant (WTP) Why DOE-EM Did This Review The Hanford Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the 53 million gallons of radioactive waste, separate it into high- and low-activity fractions, and produce canisters of high-level (HLW) glass (left) and

  9. 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.

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

    SciTech Connect (OSTI)

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

    2012-12-13

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

  11. 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.

  12. 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 divided into Part 1 that defines time-dependent releases from each regional site, Part 2 that defines transport conditions through the groundwater, and Part 3 that defines transport through surface water and populations using the surface waters for drinking.

  13. 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.

  14. 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)...

  15. 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.

  16. 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.

  17. Summary - WTP HLW Waste Vitrification Facility

    Office of Environmental Management (EM)

    ( adequacy (W processing H corrosion res stainless ste Testing activ contacted wi halogens), s Testing the p blending liqu requirements established. maturity of certain key...

  18. 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.

  19. 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.

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

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

    . H L W System Plan Revision 1 (U) Westinghouse Savannah River Company Aiken, South Carolina . August 4, 1993 1.0 2.0 3.0 4.0 5.0 6.0 HLW System Plan - Revision 1 (U) Table of Contents Mission Statement Purpose Executive Summary 3.1 Reference Date 3.2 Key Milestones 3.3 Operational Plan Summary 3.4 Key Issues and Assumptions 3.5 HLW System Plan Management .HLW System Description Principles and Guidelines 5.1 Safety Documentation 5.2 Regulatory Permits 5.3 Long Range Planning 5.4 Roadmaps 5.5 DOE

  1. 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.

  2. 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 prepared in response to an ORP-supplied statement of work. The present glass formulation and melter testing work was aimed at one of the four waste streams previously specified by the Office of River Protection (ORP). Such testing supports the ORP basis for projection of the amount of Immobilized High Level Waste (IHLW) to be produced at Hanford and evaluation of the likely potential for future enhancements of the WTP over and above the present well-developed baseline. It should be noted that the compositions of the four ORP-specified waste streams differ significantly from those of the feed tanks (AZ-101, AZ-102, C-16/AY-102, and C-104/AY-101) that have been the focus of the extensive technology development and design work performed for the WTP baseline. In this regard, the work on the ORP-specified compositions is complementary to and necessarily of a more exploratory nature than the work in support of the current WTP baseline.

  3. Process for solidifying high-level nuclear waste

    DOE Patents [OSTI]

    Ross, Wayne A. (Richland, WA)

    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.

  4. 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 relatively high concentrations of zirconium and aluminum, representative of the cladding material of the reprocessed fuel that generated the calcine. A separate study to define the CCIM testing needs of these other calcine classifications in currently being prepared under a separate work package (WP-0) and will be provided as a milestone report at the end of this fiscal year.

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

    DOE Patents [OSTI]

    Cowan, Richard G. (Kennewick, WA); Blasewitz, Albert G. (Richland, WA)

    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.

  6. 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.

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

    Office of Environmental Management (EM)

    Low- LEVEL WASTE DISPOSAL FACILITY FEDERAL REVIEW GROUP MANUAL REVISION 3 JUNE 2008 (This page intentionally left blank) Low-Level JVllsfe Disposal Fllcili~l' Federal Review Group il1allUlli Revision 3, June 200S Concurrence The Low-Level Waste Disposal Facility Federal Review Group Manual, Revision 3, is approved for use as of the most recent date below. Date Chair, Low-Level Waste Disposal Federal Review Group Andrew WalJo, 1II Deputy Director, Otlice of Nuclear Safety, Quality Assurance, and

  8. 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)

  9. 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-resistive tomography. This current work utilized the same instruments to monitor simulated waste transfers. This paper will discuss some of the scaling compromises when it came to the scaled sampling system design, handling of large quantities of material for sampling, and present data for the discuss of likely behavior of the full scale DST based on scaling correlations using a scale ratio exponent (SRE) from 0.25 to 0.45 and the behavior observed in the SSMD platform. This does not establish a scaling factor for DST mixing using paired jet mixers but is an attempt to envelope the likely performance ranges in terms of certification sampling bias, certification sample root-mean-square-deviation, and bath to batch relative standard deviation. (authors)

  10. 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 simulant was used for all tests and was spiked with the required amount of noble metals immediately prior to performing the test. Acid addition was kept effectively constant except to compensate for variations in the starting mercury concentration. SME cycles were also performed during six of the tests.

  11. 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.

  12. 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.

  13. 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.

  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. 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)

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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)

  6. 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.

  7. 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 ...

  8. 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. DOE’s 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.

  9. 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.

  10. Maintenance Guide for DOE Low-Level Waste Disposal Facility | Department of

    Energy Savers [EERE]

    Energy Maintenance Guide for DOE Low-Level Waste Disposal Facility Maintenance Guide for DOE Low-Level Waste Disposal Facility Maintenance Guide for U.S. Department of Energy Low-Level Waste Disposal Facility Performance Assessments and Composite Analyses PDF icon Maintenance Guide for DOE Low-Level Waste Disposal Facility More Documents & Publications Format and Content Guide for DOE Low-Level Waste Disposal Facility Format and Content Guide for DOE Low-Level Waste Disposal Facility

  11. 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.

  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. 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.

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

    Office of Environmental Management (EM)

    3 June 2008 UPCOMING EVENTS: Next High-Level Waste Corporate Board meeting will be held at DOE-ID on 24 July 2008. Meeting details will be presented here and e-mailed to those persons with an interest to participate. Topics for discussion include: * Strategic Planning Initiative * Technology Development / Needs Collection / Prioritization * Waste Acceptance Product Specification This meeting will include a members-only executive session OTHER NEWS DOE SELECTS WASHINGTON RIVER PROTECTION

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

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

    SciTech Connect (OSTI)

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

    2012-09-30

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

  4. 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.

  5. A preliminary evaluation of alternatives for treatment of INEL Low-Level Waste and low-level mixed waste

    SciTech Connect (OSTI)

    Smith, T.H.; Roesener, W.S.; Jorgensen-Waters, M.J.; Edinborough, C.R.

    1992-06-01

    The Mixed and Low-Level Waste Treatment Facility (MLLWTF) project was established in 1991 by the US Department of Energy Idaho Field Office to provide treatment capabilities for Idaho National Engineering Laboratory (INEL) low-level mixed waste and low-level waste. This report identifies and evaluates the alternatives for treating that waste. Twelve treatment alternatives, ranging from ``no-action`` to constructing and operating the MLLWTF, are identified and evaluated. Evaluations include facility performance, environmental, safety, institutional, schedule, and rough order-of-magnitude 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 decision making. Analysis of results indicated further study is necessary to obtain the best estimate of future waste volumes and characteristics from the expanded INEL Decontamination and Decommissioning Program. It is also recommended that conceptual design begin as scheduled on the MLLWTF, maximum treatment alternative while re-evaluating the waste volume projections.

  6. Low-level radioactive waste form qualification testing

    SciTech Connect (OSTI)

    Sohal, M.S.; Akers, D.W.

    1998-06-01

    This report summarizes activities that have already been completed as well as yet to be performed by the Idaho National Engineering and Environmental Laboratory (INEEL) to develop a plan to quantify the behavior of radioactive low-level waste forms. It briefly describes the status of various tasks, including DOE approval of the proposed work, several regulatory and environmental related documents, tests to qualify the waste form, preliminary schedule, and approximate cost. It is anticipated that INEEL and Brookhaven National Laboratory will perform the majority of the tests. For some tests, services of other testing organizations may be used. It should take approximately nine months to provide the final report on the results of tests on a waste form prepared for qualification. It is anticipated that the overall cost of the waste quantifying service is approximately $150,000. The following tests are planned: compression, thermal cycling, irradiation, biodegradation, leaching, immersion, free-standing liquid tests, and full-scale testing.

  7. Mixed and Low-Level Treatment Facility Project. Appendix B, Waste stream engineering files, Part 1, Mixed waste streams

    SciTech Connect (OSTI)

    Not Available

    1992-04-01

    This appendix contains the mixed and low-level waste engineering design files (EDFS) documenting each low-level and mixed waste stream investigated during preengineering studies for Mixed and Low-Level Waste Treatment Facility Project. The EDFs provide background information on mixed and low-level waste generated at the Idaho National Engineering Laboratory. They identify, characterize, and provide treatment strategies for the waste streams. Mixed waste is waste containing both radioactive and hazardous components as defined by the Atomic Energy Act and the Resource Conservation and Recovery Act, respectively. Low-level waste is waste that contains radioactivity and is not classified as high-level waste, transuranic waste, spent nuclear fuel, or 11e(2) byproduct material as defined by DOE 5820.2A. Test specimens of fissionable material irradiated for research and development only, and not for the production of power or plutonium, may be classified as low-level waste, provided the concentration of transuranic is less than 100 nCi/g. This appendix is a tool that clarifies presentation format for the EDFS. The EDFs contain waste stream characterization data and potential treatment strategies that will facilitate system tradeoff studies and conceptual design development. A total of 43 mixed waste and 55 low-level waste EDFs are provided.

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

    Office of Environmental Management (EM)

    UPCOMING EVENTS: The Low-Level 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 Board meeting will be held at DOE- RL on 6 November 2008. Meeting details will be presented here and e- mailed to those persons with an interest to participate. Topics for discussion include but are not limited to: ï‚· Results of the Tank Integrity Workshop ï‚· Strategic Initiative Briefing ï‚·

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

    Office of Environmental Management (EM)

    4 G Approved: XX-XX-XX IMPLEMENTATION GUIDE for use with DOE M 435.1-1 Maintenance Guide for U.S. Department of Energy Low-Level Waste Disposal Facility Performance Assessments and Composite Analyses U.S. DEPARTMENT OF ENERGY DOE G 435.1-4 i (and ii) DRAFT XX-XX-XX LLW Maintenance Guide Revision 0, XX-XX-XX Maintenance Guide for U.S. Department of Energy Low-Level Waste Disposal Facility Performance Assessments and Composite Analyses CONTENTS 1. INTRODUCTION . . . . . . . . . . . . . . . . . . .

  10. Steam Reforming of Low-Level Mixed Waste

    SciTech Connect (OSTI)

    1998-01-01

    Under DOE Contract No. DE-AR21-95MC32091, Steam Reforming of Low-Level Mixed Waste, ThermoChem has successfully designed, fabricated and operated a nominal 90 pound per hour Process Development Unit (PDU) on various low-level mixed waste surrogates. The design construction, and testing of the PDU as well as performance and economic projections for a 500- lb/hr demonstration and commercial system are described. The overall system offers an environmentally safe, non-incinerating, cost-effective, and publicly acceptable method of processing LLMW. The steam-reforming technology was ranked the No. 1 non-incineration technology for destruction of hazardous organic wastes in a study commissioned by the Mixed Waste Focus Area published April 1997.1 The ThermoChem steam-reforming system has been developed over the last 13 years culminating in this successful test campaign on LLMW surrogates. Six surrogates were successfidly tested including a 750-hour test on material simulating a PCB- and Uranium- contaminated solid waste found at the Portsmouth Gaseous Diffusion Plant. The test results indicated essentially total (>99.9999oA) destruction of RCRA and TSCA hazardous halogenated organics, significant levels of volume reduction (> 400 to 1), and retention of radlonuclides in the volume-reduced solids. Cost studies have shown the steam-reforming system to be very cost competitive with more conventional and other emerging technologies.

  11. Mixed and low-level waste treatment facility project

    SciTech Connect (OSTI)

    Not Available

    1992-04-01

    The technology information provided in this report is only the first step toward the identification and selection of process systems that may be recommended for a proposed mixed and low-level waste treatment facility. More specific information on each technology will be required to conduct the system and equipment tradeoff studies that will follow these preengineering studies. For example, capacity, maintainability, reliability, cost, applicability to specific waste streams, and technology availability must be further defined. This report does not currently contain all needed information; however, all major technologies considered to be potentially applicable to the treatment of mixed and low-level waste are identified and described herein. Future reports will seek to improve the depth of information on technologies.

  12. Flammable gas tank waste level reconcilliation for 241-SX-102

    SciTech Connect (OSTI)

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

    1997-06-23

    Fluoro Dynel Northwest (FDNW) was authorized to address flammable gas issues by reconciling the unexplained surface level increases in Tank 24 1-S-1 1 1 (S-I 1 1, typical). The trapped gas evaluation document (ref 1) states that Tank SX-102 exceeds the 25% of the lower flammable limit (FL) criterion (ref 2), based on a surface level rise evaluation. The Waste Storage Tank Status and Leak Detection Criteria document, commonly referred to as the ``Wallet Report`` is the basis for this letter report (ref 3). The Wallet Report is also a part of the trapped gas evaluation document criteria. The Wallet Report contains various tank information, including: physical information, status, levels, and dry wells, see Appendix A. The unexplained waste level rises were attributed to the production and retention of gas in the column of waste corresponding to the unacquainted for surface level rise. From 1973 through 1980, the Wallet Report tracked Tank S- 102 transfers and reported a net cumulative change of 19.95 in. This surface level increase is from an unknown source or is unacquainted for. Duke Engineering and Services Hanford (DASH) and Leached Martin Hanford Corporation (LMHC) are interested in determining the validity of the unexplained surface level changes reported in the 0611e Wallet 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 unacquainted for surface level changes as shown in the Wallet Report from 1973 through 1980.

  13. Solvent Extraction of Tc and Cs from Alkaline Nitrate Wastes

    SciTech Connect (OSTI)

    Bonnesen, P.V.; Conner, C.; Delmau, L.H.; Haverlock, T.J.; Leonard, R.A.; Lumetta, G.J.; Moyer, B.A.; Sachleben, R.A.

    1999-07-11

    This paper summarizes progress at three collaborating US national laboratories on the extraction of the fission products {sup 99}Tc and {sup 137}Cs from alkaline high-level wastes (HLW). Efficient, economical processes for Tc and Cs extraction (SRTALK and alkaline-side CSEX, respectively) have been developed, and testing has progressed through batch tests on actual wastes and continuous countercurrent centrifugal-contactor tests on simulants.

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

    SciTech Connect (OSTI)

    WILLIS, W.L.

    2000-06-15

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

  15. Management of radioactive mixed wastes in commercial low-level wastes

    SciTech Connect (OSTI)

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

    1986-01-01

    Potential mixed wastes in commercial low-level wastes have been identified and management options applicable to these wastes have been evaluated. Both the identification and management evaluation have necessarily been based on review of NRC and EPA regulations and recommendations. The underlying intent of both agencies is protection of man and/or environment, but differences may occur in the means by which intent is achieved. Apparent discrepancies, data gaps and unresolved issues that have surfaced during the course of this work are discussed.

  16. Spent Fuel and High-Level Radioactive Waste Transportation Report

    SciTech Connect (OSTI)

    Not Available

    1992-03-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by SSEB in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste Issues. In addition. this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  17. Spent fuel and high-level radioactive waste transportation report

    SciTech Connect (OSTI)

    Not Available

    1989-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages sew be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  18. Spent fuel and high-level radioactive waste transportation report

    SciTech Connect (OSTI)

    Not Available

    1990-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

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

    SciTech Connect (OSTI)

    Kelly, Steven E.

    2013-11-11

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

  20. 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 carried out in the frame of the 6. Framework Program of the European Commission) aims at completing and optimizing the direct disposal concept for spent fuel by a full-scale demonstration of the technology of emplacement in vertical boreholes. The third project is devoted to the development of a reference concept to dispose of HLW in deep geological repository in clay in Germany. In the following a brief overview is given on the achievements, the projects, and ideas about the consequences for HLW disposal in Germany. (author)

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

    SciTech Connect (OSTI)

    Deffenbaugh, M.L.

    1997-09-04

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

  2. Impact Of Particle Agglomeration On Accumulation Rates In The Glass Discharge Riser Of HLW Melter

    SciTech Connect (OSTI)

    Kruger, A. A.; Rodriguez, C. A.; Matyas, J.; Owen, A. T.; Jansik, D. P.; Lang, J. B.

    2012-11-12

    The major factor limiting waste loading in continuous high-level radioactive waste (HLW) melters is an accumulation of particles in the glass discharge riser during a frequent and periodic idling of more than 20 days. An excessive accumulation can produce robust layers a few centimeters thick, which may clog the riser, preventing molten glass from being poured into canisters. Since the accumulation rate is driven by the size of particles we investigated with x-ray microtomography, scanning electron microscopy, and image analysis the impact of spinel forming components, noble metals, and alumina on the size, concentration, and spatial distribution of particles, and on the accumulation rate. Increased concentrations of Fe and Ni in the baseline glass resulted in the formation of large agglomerates that grew over the time to an average size of ~185+-155 {mu}m, and produced >3 mm thick layer after 120 h at 850 deg C. The noble metals decreased the particle size, and therefore significantly slowed down the accumulation rate. Addition of alumina resulted in the formation of a network of spinel dendrites which prevented accumulation of particles into compact layers.

  3. Technical considerations for evaluating substantially complete containment of high-level waste within the waste package

    SciTech Connect (OSTI)

    Manaktala, H.K. (Southwest Research Inst., San Antonio, TX (USA). Center for Nuclear Waste Regulatory Analyses); Interrante, C.G. (Nuclear Regulatory Commission, Washington, DC (USA). Div. of High-Level Waste Management)

    1990-12-01

    This report deals with technical information that is considered essential for demonstrating the ability of the high-level radioactive waste package to provide substantially complete containment'' of its contents (vitrified waste form or spent light-water reactor fuel) for a period of 300 to 1000 years in a geological repository environment. The discussion is centered around technical considerations of the repository environment, materials and fabrication processes for the waste package components, various degradation modes of the materials of construction of the waste packages, and inspection and monitoring of the waste package during the preclosure and retrievability period, which could begin up to 50 years after initiation of waste emplacement. The emphasis in this report is on metallic materials. However, brief references have been made to other materials such as ceramics, graphite, bonded ceramic-metal systems, and other types of composites. The content of this report was presented to an external peer review panel of nine members at a workshop held at the Center for Nuclear Waste Regulatory Analyses (CNWRA), Southwest Research Institute, San Antonio, Texas, April 2--4, 1990. The recommendations of the peer review panel have been incorporated in this report. There are two companion reports; the second report in the series provides state-of-the-art techniques for uncertainty evaluations. 97 refs., 1 fig.

  4. Waste Processing Annual Technology Development Report 2007 | Department of

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

    Energy PDF icon Waste Processing Annual Technology Development Report 2007 More Documents & Publications System Planning for Low-Activity Waste at Hanford Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility Caustic Recovery Technology

  5. Statement of work for conceptual design of solidified high-level waste interim storage system project (phase I)

    SciTech Connect (OSTI)

    Calmus, R.B., Westinghouse Hanford

    1996-12-17

    The U.S. Department of Energy (DOE) has embarked upon a course to acquire Hanford Site tank waste treatment 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 (IS) and disposal of various products from privatized facilities are to be DOE furnished. The path forward adopted for Phase I solidification HLW IS entails use of Vaults 2 and 3 in the Spent Nuclear Fuel Canister Storage Building, to be located in the Hanford Site 200 East Area. This Statement of Work describes the work scope to be performed by the Architect-Engineer to prepare a conceptual design for the solidified HLW IS System.

  6. Control of high level radioactive waste-glass melters

    SciTech Connect (OSTI)

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

    1991-01-01

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

  7. Evaluation of Flygt Propeller Xixers for Double Shell Tank (DST) High Level Waste Auxiliary Solids Mobilization

    SciTech Connect (OSTI)

    PACQUET, E.A.

    2000-07-20

    The River Protection Project (RPP) is planning to retrieve radioactive waste from the single-shell tanks (SST) and double-shell tanks (DST) underground at the Hanford Site. This waste will then be transferred to a waste treatment plant to be immobilized (vitrified) in a stable glass form. Over the years, the waste solids in many of the tanks have settled to form a layer of sludge at the bottom. The thickness of the sludge layer varies from tank to tank, from no sludge or a few inches of sludge to about 15 ft of sludge. The purpose of this technology and engineering case study is to evaluate the Flygt{trademark} submersible propeller mixer as a potential technology for auxiliary mobilization of DST HLW solids. Considering the usage and development to date by other sites in the development of this technology, this study also has the objective of expanding the knowledge base of the Flygt{trademark} mixer concept with the broader perspective of Hanford Site tank waste retrieval. More specifically, the objectives of this study delineated from the work plan are described.

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

    Energy Savers [EERE]

    Department of Energy Report on Separate Disposal of Defense High-Level Radioactive Waste Report on Separate Disposal of Defense High-Level Radioactive Waste This report considers whether a separate repository for high-level radioactive waste resulting from atomic energy defense activities is "required" within the meaning of Section 8(b)(2) of the Nuclear Waste Policy Act of 1982. PDF icon Report on Separate Disposal of Defense High-Level Radioactive Waste More Documents &

  9. Steam reforming of low-level mixed waste. Final report

    SciTech Connect (OSTI)

    1998-06-01

    ThermoChem has successfully designed, fabricated and operated a nominal 90 pound per hour Process Development Unit (PDU) on various low-level mixed waste surrogates. The design, construction, and testing of the PDU as well as performance and economic projections for a 300-lb/hr demonstration and commercial system are described. The overall system offers an environmentally safe, non-incinerating, cost-effective, and publicly acceptable method of processing LLMW. The steam-reforming technology was ranked the No. 1 non-incineration technology for destruction of hazardous organic wastes in a study commissioned by the Mixed Waste Focus Area and published in April 1997. The ThermoChem steam-reforming system has been developed over the last 13 years culminating in this successful test campaign on LLMW surrogates. Six surrogates were successfully tested including a 750-hour test on material simulating a PCB- and Uranium-contaminated solid waste found at the Portsmouth Gaseous Diffusion Plant. The test results indicated essentially total (> 99.9999%) destruction of RCRA and TSCA hazardous halogenated organics, significant levels of volume reduction (> 400 to 1), and retention of radionuclides in the volume-reduced solids. Economic evaluations have shown the steam-reforming system to be very cost competitive with more conventional and other emerging technologies.

  10. Final Report - Melt Rate Enhancement for High Aluminum HLW Glass Formulation, VSL-08R1360-1, Rev. 0, dated 12/19/08

    SciTech Connect (OSTI)

    Kruger, Albert A.; Pegg, I. L.; Chaudhuri, M.; Gong, W.; Gan, H.; Matlack, K. S.; Bardakci, T.; Kot, W.

    2013-11-13

    The principal objective of the work reported here was to develop and identify HLW glass compositions that maximize waste processing rates for the aluminum limted waste composition specified by ORP while maintaining high waste loadings and acceptable glass properties. This was accomplished through a combination of crucible-scale tests, confirmation tests on the DM100 melter system, and demonstration at pilot scale (DM1200). The DM100-BL unit was selected for these tests since it was used previously with the HLW waste streams evaluated in this study, was used for tests on HLW glass compositions to support subsequent tests on the HLW Pilot Melter, conduct tests to determine the effect of various glass properties (viscosity and conductivity) and oxide concentrations on glass production rates with HLW feed streams, and to assess the volatility of cesium and technetium during the vitrification of an HLW AZ-102 composition. The same melter was selected for the present tests in order to maintain comparisons between the previously collected data. These tests provide information on melter processing characteristics and off-gas data, including formation of secondary phases and partitioning. Once DM100 tests were completed, one of the compositions was selected for further testing on the DM1200; the DM1200 system has been used for processing a variety of simulated Hanford waste streams. Tests on the larger melter provide processing data at one third of the scale of the actual WTP HLW melter and, therefore, provide a more accurate and reliable assessment of production rates and potential processing issues. The work focused on maximizing waste processing rates for high aluminum HLW compositions. In view of the diversity of forms of aluminum in the Hanford tanks, tests were also conducted on the DM100 to determine the effect of changes in the form of aluminum on feed properties and production rate. In addition, the work evaluated the effect on production rate of modest increases in melter operating temperature. Glass composition development was based on one of the HLW waste compositions specified by ORP that has a high concentration of aluminum. Small-scale tests were used to provide an initial screening of various glass formulations with respect to melt rates; more definitive screening was provided by the subsequent DM100 tests. Glass properties evaluated included: viscosity, electrical conductivity, crystallinity, gross glass phase separation and the 7- day Product Consistency Test (ASTM-1285). Glass property limits were based upon the reference properties for the WTP HLW melter. However, the WTP crystallinity limit (< 1 vol% at 950oC) was relaxed slightly as a waste loading constraint for the crucible melts.

  11. Mixed and low-level waste treatment facility project. Volume 3, Waste treatment technologies (Draft)

    SciTech Connect (OSTI)

    Not Available

    1992-04-01

    The technology information provided in this report is only the first step toward the identification and selection of process systems that may be recommended for a proposed mixed and low-level waste treatment facility. More specific information on each technology will be required to conduct the system and equipment tradeoff studies that will follow these preengineering studies. For example, capacity, maintainability, reliability, cost, applicability to specific waste streams, and technology availability must be further defined. This report does not currently contain all needed information; however, all major technologies considered to be potentially applicable to the treatment of mixed and low-level waste are identified and described herein. Future reports will seek to improve the depth of information on technologies.

  12. International low level waste disposal practices and facilities

    SciTech Connect (OSTI)

    Nutt, W.M.

    2011-12-19

    The safe management of nuclear waste arising from nuclear activities is an issue of great importance for the protection of human health and the environment now and in the future. The primary goal of this report is to identify the current situation and practices being utilized across the globe to manage and store low and intermediate level radioactive waste. The countries included in this report were selected based on their nuclear power capabilities and involvement in the nuclear fuel cycle. This report highlights the nuclear waste management laws and regulations, current disposal practices, and future plans for facilities of the selected international nuclear countries. For each country presented, background information and the history of nuclear facilities are also summarized to frame the country's nuclear activities and set stage for the management practices employed. The production of nuclear energy, including all the steps in the nuclear fuel cycle, results in the generation of radioactive waste. However, radioactive waste may also be generated by other activities such as medical, laboratory, research institution, or industrial use of radioisotopes and sealed radiation sources, defense and weapons programs, and processing (mostly large scale) of mineral ores or other materials containing naturally occurring radionuclides. Radioactive waste also arises from intervention activities, which are necessary after accidents or to remediate areas affected by past practices. The radioactive waste generated arises in a wide range of physical, chemical, and radiological forms. It may be solid, liquid, or gaseous. Levels of activity concentration can vary from extremely high, such as levels associated with spent fuel and residues from fuel reprocessing, to very low, for instance those associated with radioisotope applications. Equally broad is the spectrum of half-lives of the radionuclides contained in the waste. These differences result in an equally wide variety of options for the management of radioactive waste. There is a variety of alternatives for processing waste and for short term or long term storage prior to disposal. Likewise, there are various alternatives currently in use across the globe for the safe disposal of waste, ranging from near surface to geological disposal, depending on the specific classification of the waste. At present, there appears to be a clear and unequivocal understanding that each country is ethically and legally responsible for its own wastes, in accordance with the provisions of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. Therefore the default position is that all nuclear wastes will be disposed of in each of the 40 or so countries concerned with nuclear power generation or part of the fuel cycle. To illustrate the global distribution of radioactive waste now and in the near future, Table 1 provides the regional breakdown, based on the UN classification of the world in regions illustrated in Figure 1, of nuclear power reactors in operation and under construction worldwide. In summary, 31 countries operate 433 plants, with a total capacity of more than 365 gigawatts of electrical energy (GW[e]). A further 65 units, totaling nearly 63 GW(e), are under construction across 15 of these nations. In addition, 65 countries are expressing new interest in, considering, or actively planning for nuclear power to help address growing energy demands to fuel economic growth and development, climate change concerns, and volatile fossil fuel prices. Of these 65 new countries, 21 are in Asia and the Pacific region, 21 are from the Africa region, 12 are in Europe (mostly Eastern Europe), and 11 in Central and South America. However, 31 of these 65 are not currently planning to build reactors, and 17 of those 31 have grids of less than 5 GW, which is said to be too small to accommodate most of the reactor designs available. For the remaining 34 countries actively planning reactors, as of September 2010: 14 indicate a strong intention to precede with introduction of nuclear power; 7 are preparing but haven't made a final decision, 10 have made a decision and are preparing infrastructure, 2 have ordered a new nuclear power plant, and 1 has a plant under construction. In all countries interested in pursuing nuclear power, it is necessary for the governments to create an environment for investment and advancement of nuclear power, including development of a professional and independent regulatory framework and regime, knowledge and refinement of skills in nuclear safety and control, definition of policies on nuclear waste management and decommissioning, and participation in international non-proliferation measures. Specifically related to radioactive waste management, nuclear facilities and industries that utilize radioactive material work to well-established safety standards for the management of their waste.

  13. Microbial degradation of low-level radioactive waste. Final report

    SciTech Connect (OSTI)

    Rogers, R.D.; Hamilton, M.A.; Veeh, R.H.; McConnell, J.W. Jr

    1996-06-01

    The Nuclear Regulatory Commission stipulates in 10 CFR 61 that disposed low-level radioactive waste (LLW) be stabilized. To provide guidance to disposal vendors and nuclear station waste generators for implementing those requirements, the NRC developed the Technical Position on Waste Form, Revision 1. That document details a specified set of recommended testing procedures and criteria, including several tests for determining the biodegradation properties of waste forms. Information has been presented by a number of researchers, which indicated that those tests may be inappropriate for examining microbial degradation of cement-solidified LLW. Cement has been widely used to solidify LLW; however, the resulting waste forms are sometimes susceptible to failure due to the actions of waste constituents, stress, and environment. The purpose of this research program was to develop modified microbial degradation test procedures that would be more appropriate than the existing procedures for evaluation of the effects of microbiologically influenced chemical attack on cement-solidified LLW. The procedures that have been developed in this work are presented and discussed. Groups of microorganisms indigenous to LLW disposal sites were employed that can metabolically convert organic and inorganic substrates into organic and mineral acids. Such acids aggressively react with cement and can ultimately lead to structural failure. Results on the application of mechanisms inherent in microbially influenced degradation of cement-based material are the focus of this final report. Data-validated evidence of the potential for microbially influenced deterioration of cement-solidified LLW and subsequent release of radionuclides developed during this study are presented.

  14. Disposing of High-Level Radioactive Waste in Germany - A Note from the Licensing Authority - 12530

    SciTech Connect (OSTI)

    Pick, Thomas Stefan; Bluth, Joachim; Lauenstein, Christof; Markhoefer, Joerg

    2012-07-01

    Following the national German consensus on the termination of utilisation of nuclear energy in the summer of 2011, the Federal and Laender Governments have declared their intention to work together on a national consensus on the disposal of radioactive waste as well. Projected in the early 1970's the Federal Government had started exploring the possibility to establish a repository for HLW at the Gorleben site in 1977. However, there is still no repository available in Germany today. The delay results mainly from the national conflict over the suitability of the designated Gorleben site, considerably disrupting German society along the crevice that runs between supporters and opponents of nuclear energy. The Gorleben salt dome is situated in Lower Saxony, the German state that also hosts the infamous Asse mine repository for LLW and ILW and the Konrad repository project designated to receive LLW and ILW as well. With the fourth German project, the Morsleben L/ILW repository only 20 km away across the state border, the state of Lower Saxony carries the main load for the disposal of radioactive waste in Germany. After more than 25 years of exploration and a 10 year moratorium the Gorleben project has now reached a cross-road. Current plans for setting up a new site selection procedure in Germany call for the selection and exploration of up to four alternative sites, depending only on suitable geology. In the meantime the discussion is still open on whether the Gorleben project should be terminated in order to pacify the societal conflict or being kept in the selection process on account of its promising geology. The Lower Saxony Ministry for Environment and Climate Protection proposes to follow a twelve-step-program for finding the appropriate site, including the Gorleben site in the process. With its long history of exploration the site is the benchmark that alternative sites will have to compare with. Following the national consensus of 2011 on the termination of nuclear energy utilisation, it is now the time to reach a national consensus on the disposal of radioactive waste as well. This is a task that the country and society, federal and state governments, political parties and the citizens will have to jointly master within the current generation and within German territory. The basis for the consensus will be a reset to the beginning of this process. It has to start with a new site selection procedure that will take into account and compare up to four alternative sites. This procedure will have to follow the principle of highest possible security. It should be based on a stepwise approach, strictly following scientific criteria. Public confidence in the process and trust can only be achieved by a transparent procedure allowing for the participation of the public and the stakeholders. It is therefore mandatory to consult, both on a national and regional level, all involved parties (public authority, scientist and citizen). The national consensus must also include a decision on the future of the Gorleben exploratory site. The site selection procedure must therefore take this site into account as well. Furthermore, the final decision on safe disposal of German radioactive wastes must be made by sovereign rule by Federal Parliament and Federal Council. (authors)

  15. Secondary Waste Cast Stone Waste Form Qualification Testing Plan

    SciTech Connect (OSTI)

    Westsik, Joseph H.; Serne, R. Jeffrey

    2012-09-26

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

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

    SciTech Connect (OSTI)

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

    2011-10-06

    The purpose of the high-level waste (HLW) glass melt rate study is two-fold: (1) to gain a better understanding of the impact of feed chemistry on melt rate through bench-scale testing, and (2) to develop a predictive tool for melt rate in support of the on-going frit development efforts for the Defense Waste Processing Facility (DWPF). In particular, the focus is on predicting relative melt rates, not the absolute melt rates, of various HLW glass formulations solely based on feed chemistry, i.e., the chemistry of both waste and glass-forming frit for DWPF. Critical to the successful melt rate modeling is the accurate determination of the melting rates of various HLW glass formulations. The baseline procedure being used at the Savannah River National Laboratory (SRNL) is to; (1) heat a 4 inch-diameter stainless steel beaker containing a mixture of dried sludge and frit in a furnace for a preset period of time, (2) section the cooled beaker along its diameter, and (3) measure the average glass height across the sectioned face using a ruler. As illustrated in Figure 1-1, the glass height is measured for each of the 16 horizontal segments up to the red lines where relatively large-sized bubbles begin to appear. The linear melt rate (LMR) is determined as the average of all 16 glass height readings divided by the time during which the sample was kept in the furnace. This 'visual' method has proved useful in identifying melting accelerants such as alkalis and sulfate and further ranking the relative melt rates of candidate frits for a given sludge batch. However, one of the inherent technical difficulties of this method is to determine the glass height in the presence of numerous gas bubbles of varying sizes, which is prevalent especially for the higher-waste-loading glasses. That is, how the red lines are drawn in Figure 1-1 can be subjective and, therefore, may influence the resulting melt rates significantly. For example, if the red lines are drawn too low, a significant amount of glassy material interspersed among the gas bubbles will be excluded, thus underestimating the melt rate. Likewise, if they are drawn too high, many large voids will be counted as glass, thus overestimating the melt rate. As will be shown later in this report, there is also no guarantee that a given distribution of glass and gas bubbles along a particular sectioned plane will always be representative of the entire sample volume. Poor reproducibility seen in some LMR data may be related to these difficulties of the visual method. In addition, further improvement of the existing melt rate model requires that the overall impact of feed chemistry on melt rate be reflected on measured data at a greater quantitative resolution on a more consistent basis than the visual method can provide. An alternate method being pursued is X-ray computed tomography (CT). It involves X-ray scanning of glass samples, performing CT on the 2-D X-ray images to build 3-D volumetric data, and adaptive segmentation analysis of CT results to not only identify but quantify the distinct regions within each sample based on material density and morphologies. The main advantage of this new method is that it can determine the relative local density of the material remaining in the beaker after the heat treatment regardless of its morphological conditions by selectively excluding all the voids greater than a given volumetric pixel (voxel) size, thus eliminating much of the subjectivity involved in the visual method. As a result, the melt rate data obtained from CT scan will give quantitative descriptions not only on the fully-melted glass, but partially-melted and unmelted feed materials. Therefore, the CT data are presumed to be more reflective of the actual melt rate trends in continuously-fed melters than the visual data. In order to test the applicability of X-ray CT scan to the HLW glass melt rate study, several new series of HLW simulant/frit mixtures were melted in the Melt Rate Furnace (MRF) and the contents of each cooled but un-sectioned beaker were CT scanned and analyzed.

  17. The siting record: An account of the programs of federal agencies and events that have led to the selection of a potential site for a geologic respository for high-level radioactive waste

    SciTech Connect (OSTI)

    Lomenick, T.F.

    1996-03-01

    This record of siting a geologic repository for high-level radioactive wastes (HLW) and spent fuel describes the many investigations that culminated on December 22, 1987 in the designation of Yucca Mountain (YM), as the site to undergo detailed geologic characterization. It recounts the important issues and events that have been instrumental in shaping the course of siting over the last three and one half decades. In this long task, which was initiated in 1954, more than 60 regions, areas, or sites involving nine different rock types have been investigated. This effort became sharply focused in 1983 with the identification of nine potentially suitable sites for the first repository. From these nine sites, five were subsequently nominated by the U.S. Department of Energy (DOE) as suitable for characterization and then, in 1986, as required by the Nuclear Waste Policy Act of 1982 (NWPA), three of these five were recommended to the President as candidates for site characterization. President Reagan approved the recommendation on May 28, 1986. DOE was preparing site characterization plans for the three candidate sites, namely Deaf Smith County, Texas; Hanford Site, Washington; and YM. As a consequence of the 1987 Amendment to the NWPA, only the latter was authorized to undergo detailed characterization. A final Site Characterization Plan for Yucca Mountain was published in 1988. Prior to 1954, there was no program for the siting of disposal facilities for high-level waste (HLW). In the 1940s and 1950s, the volume of waste, which was small and which resulted entirely from military weapons and research programs, was stored as a liquid in large steel tanks buried at geographically remote government installations principally in Washington and Tennessee.

  18. Radioactive Waste Management Complex low-level waste radiological performance assessment

    SciTech Connect (OSTI)

    Maheras, S.J.; Rood, A.S.; Magnuson, S.O.; Sussman, M.E.; Bhatt, R.N.

    1994-04-01

    This report documents the projected radiological dose impacts associated with the disposal of radioactive low-level waste at the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. This radiological performance assessment was conducted to evaluate compliance with applicable radiological criteria of the US Department of Energy and the US Environmental Protection Agency for protection of the public and the environment. The calculations involved modeling the transport of radionuclides from buried waste, to surface soil and subsurface media, and eventually to members of the public via air, groundwater, and food chain pathways. Projections of doses were made for both offsite receptors and individuals inadvertently intruding onto the site after closure. In addition, uncertainty and sensitivity analyses were performed. The results of the analyses indicate compliance with established radiological criteria and provide reasonable assurance that public health and safety will be protected.

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

    SciTech Connect (OSTI)

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

    2009-07-01

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

  20. Final Report - Testing of Optimized Bubbler Configuration for HLW Melter VSL-13R2950-1, Rev. 0, dated 6/12/2013

    SciTech Connect (OSTI)

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

    2013-11-13

    The principal objective of this work was to determine the glass production rate increase and ancillary effects of adding more bubbler outlets to the current WTP HLW melter baseline. This was accomplished through testing on the HLW Pilot Melter (DM1200) at VSL. The DM1200 unit was selected for these tests since it was used previously with several HLW waste streams including the four tank wastes proposed for initial processing at Hanford. This melter system was also used for the development and optimization of the present baseline WTP HLW bubbler configuration for the WTP HLW melter, as well as for MACT testing for both HLW and LAW. Specific objectives of these tests were to: Conduct DM1200 melter testing with the baseline WTP bubbling configuration and as augmented with additional bubblers. Conduct DM1200 melter testing to differentiate the effects of total bubbler air flow and bubbler distribution on glass production rate and cold cap formation. Collect melter operating data including processing rate, temperatures at a variety of locations within the melter plenum space, melt pool temperature, glass melt density, and melter pressure with the baseline WTP bubbling configuration and as augmented with additional bubblers. Collect melter exhaust samples to compare particulate carryover for different bubbler configurations. Analyze all collected data to determine the effects of adding more bubblers to the WTP HLW melter to inform decisions regarding future lid re-designs. The work used a high aluminum HLW stream composition defined by ORP, for which an appropriate simulant and high waste loading glass formulation were developed and have been previously processed on the DM1200.

  1. 12/2000 Low-Level Waste Disposal Capacity Report Version 2 | Department of

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

    Energy Waste Disposition » 12/2000 Low-Level Waste Disposal Capacity Report Version 2 12/2000 Low-Level Waste Disposal Capacity Report Version 2 The purpose of this Report is to assess whether U.S. Department of Energy (DOE or the Department) disposal facilities have sufficient volumetric and radiological capacity to accommodate the low-level waste (LLW) and mixed low-level waste (MLLW) that the Department expects to dispose at these facilities. PDF icon 12/2000 Low-Level Waste Disposal

  2. Clean option: An alternative strategy for Hanford Tank Waste Remediation. Volume 2, Detailed description of first example flowsheet

    SciTech Connect (OSTI)

    Swanson, J.L.

    1993-09-01

    Disposal of high-level tank wastes at the Hanford Site is currently envisioned to divide the waste between two principal waste forms: glass for the high-level waste (HLW) and grout for the low-level waste (LLW). The draft flow diagram shown in Figure 1.1 was developed as part of the current planning process for the Tank Waste Remediation System (TWRS), which is evaluating options for tank cleanup. The TWRS has been established by the US Department of Energy (DOE) to safely manage the Hanford tank wastes. It includes tank safety and waste disposal issues, as well as the waste pretreatment and waste minimization issues that are involved in the ``clean option`` discussed in this report. This report describes the results of a study led by Pacific Northwest Laboratory to determine if a more aggressive separations scheme could be devised which could mitigate concerns over the quantity of the HLW and the toxicity of the LLW produced by the reference system. This aggressive scheme, which would meet NRC Class A restrictions (10 CFR 61), would fit within the overall concept depicted in Figure 1.1; it would perform additional and/or modified operations in the areas identified as interim storage, pretreatment, and LLW concentration. Additional benefits of this scheme might result from using HLW and LLW disposal forms other than glass and grout, but such departures from the reference case are not included at this time. The evaluation of this aggressive separations scheme addressed institutional issues such as: radioactivity remaining in the Hanford Site LLW grout, volume of HLW glass that must be shipped offsite, and disposition of appropriate waste constituents to nonwaste forms.

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

    SciTech Connect (OSTI)

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

    2011-12-29

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

  4. Greater-than-Class C Low-Level Radioactive Waste (GTCC LLW) | Department of

    Energy Savers [EERE]

    Energy Greater-than-Class C Low-Level Radioactive Waste (GTCC LLW) Greater-than-Class C Low-Level Radioactive Waste (GTCC LLW) A transuranic (TRU) waste shipment makes its way to the Waste Isolation Pilot Plant in Carlsbad, N.M. A transuranic (TRU) waste shipment makes its way to the Waste Isolation Pilot Plant in Carlsbad, N.M. On February 17, 2011, DOE issued the Draft Environmental Impact Statement (EIS) for the Disposal of Greater-Than-Class C (GTCC) Low-Level Radioactive Waste (LLRW)

  5. Application of decontamination and melting of low-level waste

    SciTech Connect (OSTI)

    Clements, D.W.; Hall, M.

    1996-12-31

    This paper describes the range of plant, equipment, and techniques developed by British Nuclear Fuels plc at their Capenhurst site to minimize land burial, environmental impact, and recycling of metals. This large nuclear processing facility in the United Kingdom yielded more than 160000 t of suspect surface contaminated material. By the time the project is finally completed at the end of 1996, {approx}99.5% of the contaminated material will have been safely and cost-effectively treated so that it can be recycled for use in a nonnuclear environment. The remaining material as well as minimal quantities of secondary wastes arising from decontamination activities will have been size reduced and/or encapsulated to maximize the cost-effective use of the U.K. low-level-waste burial facility.

  6. High-level waste tank farm set point document

    SciTech Connect (OSTI)

    Anthony, J.A. III

    1995-01-15

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

  7. On-site waste storage assuring the success of on-site, low-level nuclear waste storage

    SciTech Connect (OSTI)

    Preston, E.L.

    1986-09-21

    Waste management has reached paramount importance in recent years. The successful management of radioactive waste is a key ingredient in the successful operation of any nuclear facility. This paper discusses the options available for on-site storage of low-level radioactive waste and those options that have been selected by the Department of Energy facilities operated by Martin Marietta Energy Systems, Inc. in Oak Ridge, Tennessee. The focus of the paper is on quality assurance (QA) features of waste management activities such as accountability and retrievability of waste materials and waste packages, retrievability of data, waste containment, safety and environmental monitoring. Technical performance and careful documentation of that performance are goals which can be achieved only through the cooperation of numerous individuals from waste generating and waste managing organizations, engineering, QA, and environmental management.

  8. The demonstration of continuous stirred tank reactor operations with high level waste

    SciTech Connect (OSTI)

    Peterson, R.A.

    2000-07-19

    This report contains the results of testing performed at the request of High Level Waste Engineering. These tests involved the operation of two continuous stirred tank reactors with high level waste.

  9. Format and Content Guide for DOE Low-Level Waste Disposal Facility

    Office of Environmental Management (EM)

    Format and Content 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. Format and Content Guide for DOE Low-Level Waste Disposal Facility...

    Office of Environmental Management (EM)

    Format and Content Guide for U.S. Department of Energy Low-Level Waste Disposal Facility ... for U.S. Department of Energy Low-Level Waste Disposal Facility Closure Plans CONTENTS ...

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

    Office of Scientific and Technical Information (OSTI)

    Road Map for Development of Crystal-Tolerant High Level Waste Glasses Citation Details In-Document Search Title: Road Map for Development of Crystal-Tolerant High Level Waste...

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

    SciTech Connect (OSTI)

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

    2012-01-10

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

  13. 1989 Annual report on low-level radioactive waste management progress

    SciTech Connect (OSTI)

    Not Available

    1990-10-01

    This report summarizes the progress during 1989 of states and compacts in establishing new low-level radioactive waste disposal facilities. It also provides summary information on the volume of low-level waste received for disposal in 1989 by commercially operated low-level waste disposal facilities. This report is in response to Section 7(b) of Title I of Public Law 99--240, the Low-Level Radioactive Waste Policy Amendments Act of 1985. 2 figs., 5 tabs.

  14. Commercial low-level radioactive waste transportation liability and radiological risk

    SciTech Connect (OSTI)

    Quinn, G.J.; Brown, O.F. II; Garcia, R.S.

    1992-08-01

    This report was prepared for States, compact regions, and other interested parties to address two subjects related to transporting low-level radioactive waste to disposal facilities. One is the potential liabilities associated with low-level radioactive waste transportation from the perspective of States as hosts to low-level radioactive waste disposal facilities. The other is the radiological risks of low-level radioactive waste transportation for drivers, the public, and disposal facility workers.

  15. Superconducting open-gradient magnetic separation for the pretreatment of radioactive or mixed waste vitrification feeds. 1997 annual progress report

    SciTech Connect (OSTI)

    Doctor, R.; Nunez, L. [Argonne National Lab., IL (US); Cicero-Herman, C.A. [Westinghouse Savannah River Co., Aiken, SC (US). Savannah River Technology Center; Ritter, J.A. [Univ. of South Carolina, Columbia, SC (US). Chemical Engineering Dept.; Landsberger, S. [Univ. of Texas, Austin, TX (US). Nuclear Engineering Dept.

    1997-01-01

    'Vitrification has been selected as a final waste form technology in the US for long-term storage of high-level radioactive wastes (HLW). However, a foreseeable problem during vitrification in some waste feed streams lies in the presence of elements (e.g., transition metals) in the HLW that may cause instabilities in the final glass product. The formation of spinel compounds, such as Fe{sub 3}O{sub 4} and FeCrO{sub 4}, results in glass phase separation and reduces vitrifier lifetime, and durability of the final waste form. A superconducting open gradient magnetic separation (OGMS) system maybe suitable for the removal of the deleterious transition elements (e.g. Fe, Co, and Ni) and other elements (lanthanides) from vitrification feed streams due to their ferromagnetic or paramagnetic nature. The OGMS systems are designed to deflect and collect paramagnetic minerals as they interact with a magnetic field gradient. This system has the potential to reduce the volume of HLW for vitrification and ensure a stable product. In order to design efficient OGMS and High gradient magnetic separation (HGMS) processes, a fundamental understanding of the physical and chemical properties of the waste feed streams is required. Using HLW simulant and radioactive fly ash and sludge samples from the Savannah River Technology Center, Rocky Flats site, and the Hanford reservation, several techniques were used to characterize and predict the separation capability for a superconducting OGMS system.'

  16. Soil characterization methods for unsaturated low-level waste sites

    SciTech Connect (OSTI)

    Wierenga, P.J.; Young, M.H. . Dept. of Soil and Water Science); Gee, G.W.; Kincaid, C.T. ); Hills, R.G. . Dept. of Mechanical Engineering); Nicholson, T.J.; Cady, R.E. )

    1993-01-01

    To support a license application for the disposal of low-level radioactive waste (LLW), applicants must characterize the unsaturated zone and demonstrate that waste will not migrate from the facility boundary. This document provides a strategy for developing this characterization plan. It describes principles of contaminant flow and transport, site characterization and monitoring strategies, and data management. It also discusses methods and practices that are currently used to monitor properties and conditions in the soil profile, how these properties influence water and waste migration, and why they are important to the license application. The methods part of the document is divided into sections on laboratory and field-based properties, then further subdivided into the description of methods for determining 18 physical, flow, and transport properties. Because of the availability of detailed procedures in many texts and journal articles, the reader is often directed for details to the available literature. References are made to experiments performed at the Las Cruces Trench site, New Mexico, that support LLW site characterization activities. A major contribution from the Las Cruces study is the experience gained in handling data sets for site characterization and the subsequent use of these data sets in modeling studies.

  17. Conversion of transuranic waste to low level waste by decontamination: a site specific update

    SciTech Connect (OSTI)

    Allen, R.P.; Hazelton, R.F.

    1985-09-01

    As a followup to an FY-1984 cost/benefit study, a program was conducted in FY-1985 to transfer to the relevant DOE sites the information and technology for the direct conversion of transuranic (TRU) waste to low-level waste (LLW) by decontamination. As part of this work, the economic evaluation of the various TRUW volume reduction and conversion options was updated and expanded to include site-specific factors. The results show, for the assumptions used, that size reduction, size reduction followed by decontamination, or in situ decontamination are cost effective compared with the no-processing option. The technology transfer activities included site presentations and discussions with operations and waste management personnel to identify application opportunities and site-specific considerations and constraints that could affect the implementation of TRU waste conversion principles. These discussions disclosed definite potential for the beneficial application of these principles at most of the sites, but also confirmed the existence of site-specific factors ranging from space limitations to LLW disposal restrictions that could preclude particular applications or diminish expected benefits. 8 refs., 2 figs., 4 tabs.

  18. Optimising the Performance of the Low Level Waste Repository - 12144

    SciTech Connect (OSTI)

    Huntington, Amy; Baker, Andrew; Cummings, Richard; Shevelan, John; Sumerling, Trevor

    2012-07-01

    The Low Level Waste Repository (LLWR) is the United Kingdom's principal facility for the disposal of low-level waste (LLW). The LLWR made a major submission to its environmental regulator (the Environment Agency) on 1 May 2011, the LLWR's 2011 Environmental Safety Case (ESC). One of the key regulatory requirements is that all aspects of the construction, operation and closure of the disposal facility should be optimised. An optimised Site Development Plan for the repository was developed and produced as part of the ESC. The Site Development Plan covers all aspects of the construction, operation and closure of the disposal facility. This includes the management of past and future disposals, emplacement strategies, design of the disposal vaults, and the closure engineering for the site. The Site Development Plan also covers the period of active institutional control, when disposals at the site have ceased, but it is still under active management, and plans for the long-term sustainable use of the site. We have a practical approach to optimisation based on recorded judgements and realistic assessments of practicable options framed within the demands of UK policy for LLW management and the characteristics the LLWR site and existing elements of the facility. The final performance assessments undertaken for the ESC were based on the Site Development Plan. The ESC will be used as a tool to inform future decision-making concerning the repository design, operation and the acceptance of wastes, as set out in the evolving Site Development Plan. Maintaining the ESC is thus essential to ensure that the Site Development Plan takes account of an up-to-date understanding and analysis of environmental performance, and that the Plan continues to be optimised. (authors)

  19. Properties of slag concrete for low-level waste containment

    SciTech Connect (OSTI)

    Langton, C.A. (Westinghouse Savannah River Co., Aiken, SC (United States)); Wong, P.B. (Bechtel National, Inc., Aiken, SC (United States))

    1991-01-01

    Ground granulated blast furnace slag was incorporated in the concrete mix used for construction of low-level radioactive waste disposal vaults. The vaults were constructed as six 100 {times} 100 {times} 25 ft cells with each cell sharing internal walls with the two adjacent cells. The vaults were designed to contain a low-level radioactive wasteform called saltstone and to isolate the saltstone from the environment until the landfill is closed. Closure involves backfilling with native soil, installation of clay cap, and run-off control. The design criteria for the slag-substituted concrete included compressive strength, 4000 psi after 28 days; slump, 6 inch; permeability, less than 10{sup {minus}7} cm/sec; and effective nitrate, chromium and technetium diffusivities of 10{sup {minus}8}, 10{sup {minus}12} and 10{sup {minus}12} cm{sup 2}/sec, respectively. The reducing capacity of the slag resulted in chemically reducing Cr{sup +6} to Cr{sup +3} and Tc{sup +7} to Tc{sup +4} and subsequent precipitation of the respective hydroxides in the alkaline pore solution. Consequently, the concrete vault enhances containment of otherwise mobile waste ions and contributes to the overall protection of the groundwater at the disposal site.

  20. Properties of slag concrete for low-level waste containment

    SciTech Connect (OSTI)

    Langton, C.A. [Westinghouse Savannah River Co., Aiken, SC (United States); Wong, P.B. [Bechtel National, Inc., Aiken, SC (United States)

    1991-12-31

    Ground granulated blast furnace slag was incorporated in the concrete mix used for construction of low-level radioactive waste disposal vaults. The vaults were constructed as six 100 {times} 100 {times} 25 ft cells with each cell sharing internal walls with the two adjacent cells. The vaults were designed to contain a low-level radioactive wasteform called saltstone and to isolate the saltstone from the environment until the landfill is closed. Closure involves backfilling with native soil, installation of clay cap, and run-off control. The design criteria for the slag-substituted concrete included compressive strength, 4000 psi after 28 days; slump, 6 inch; permeability, less than 10{sup {minus}7} cm/sec; and effective nitrate, chromium and technetium diffusivities of 10{sup {minus}8}, 10{sup {minus}12} and 10{sup {minus}12} cm{sup 2}/sec, respectively. The reducing capacity of the slag resulted in chemically reducing Cr{sup +6} to Cr{sup +3} and Tc{sup +7} to Tc{sup +4} and subsequent precipitation of the respective hydroxides in the alkaline pore solution. Consequently, the concrete vault enhances containment of otherwise mobile waste ions and contributes to the overall protection of the groundwater at the disposal site.

  1. SRS Workers Moved Millions of Gallons of High-Level Waste Safely in 2014 |

    Energy Savers [EERE]

    Department of Energy Workers Moved Millions of Gallons of High-Level Waste Safely in 2014 SRS Workers Moved Millions of Gallons of High-Level Waste Safely in 2014 February 26, 2015 - 12:00pm Addthis A view of the Savannah River Site, which includes underground waste tanks and facilities. A view of the Savannah River Site, which includes underground waste tanks and facilities. AIKEN, S.C. - EM and its liquid waste contractor safely transferred more than 20 million gallons of high-level waste

  2. Format and Content Guide for DOE Low-Level Waste Disposal Facility Closure

    Energy Savers [EERE]

    Plans | Department of Energy Closure Plans Format and Content Guide for DOE Low-Level Waste Disposal Facility Closure Plans Format and Content Guide for U.S. Department of Energy Low-Level Waste Disposal Facility Closure Plans PDF icon Format and Content Guide for DOE Low-Level Waste Disposal Facility Closure Plans More Documents & Publications Maintenance Guide for DOE Low-Level Waste Disposal Facility Format and Content Guide for DOE Low-Level Waste Disposal Facility U.S. Department of

  3. Format and Content Guide for DOE Low-Level Waste Disposal Facility |

    Energy Savers [EERE]

    Department of Energy Format and Content Guide for DOE Low-Level Waste Disposal Facility Format and Content Guide for U.S. Department of Energy Low-Level Waste Disposal Facility Performance Assessments and Composite Analyses PDF icon Format and Content Guide for DOE Low-Level Waste Disposal Facility More Documents & Publications Maintenance Guide for DOE Low-Level Waste Disposal Facility Format and Content Guide for DOE Low-Level Waste Disposal Facility Closure Plans LFRG Execution Plan

  4. Greater-than-Class C low-level radioactive waste characterization. Appendix E-4: Packaging factors for greater-than-Class C low-level radioactive waste

    SciTech Connect (OSTI)

    Quinn, G.; Grant, P.; Winberg, M.; Williams, K.

    1994-09-01

    This report estimates packaging factors for several waste types that are potential greater-than-Class C (GTCC) low-level radioactive waste (LLW). The packaging factor is defined as the volume of a GTCC LLW disposal container divided by the as-generated or ``unpackaged`` volume of the waste loaded into the disposal container. Packaging factors reflect any processes that reduce or increase an original unpackaged volume of GTCC LLW, the volume inside a waste container not occupied by the waste, and the volume of the waste container itself. Three values are developed that represent (a) the base case or most likely value for a packaging factor, (b) a high case packaging factor that corresponds to the largest anticipated disposal volume of waste, and (c) a low case packaging factor for the smallest volume expected. GTCC LLW is placed in three categories for evaluation in this report: activated metals, sealed sources, and all other waste.

  5. Direct cementitious waste option study report

    SciTech Connect (OSTI)

    Dafoe, R.E.; Losinski, S.J.

    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) will be treated so that it is ready to be moved out of Idaho for disposal by a target data of 2035. This study investigates the direct grouting of all ICPP calcine (including the HLW dry calcine and those resulting from calcining sodium-bearing liquid waste currently residing in the ICPP storage tanks) as the treatment method to comply with the settlement agreement. This method involves grouting the calcined waste and casting the resulting hydroceramic grout into stainless steel canisters. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory (INEEL) until they are sent to a national geologic repository. The operating period for grouting treatment will be from 2013 through 2032, and all the HLW will be treated and in interim storage by the end of 2032.

  6. THERMAL ANALYSIS OF WASTE GLASS MELTER FEEDS

    SciTech Connect (OSTI)

    KRUGER AA; HRMA PR; POKORNY R; PIERCE DA

    2011-10-21

    Melter feeds for high-level nuclear waste (HLW) typically contain a large number of constituents that evolve gas on heating, Multiple gas-evolving reactions are both successive and simultaneous, and include the release of chemically bonded water, reactions of nitrates with organics, and reactions of molten salts with solid silica. Consequently, when a sample of a HLW feed is subjected to thermogravimetric analysis (TGA), the rate of change of the sample mass reveals multiple overlapping peaks. In this study, a melter feed, formulated for a simulated high-alumina HLW to be vitrified in the Waste Treatment and Immobilization Plant, currently under construction at the Hanford Site in Washington State, USA, was subjected to TGA. In addition, a modified melter feed was prepared as an all-nitrate version of the baseline feed to test the effect of sucrose addition on the gas-evolving reactions. Activation energies for major reactions were determined using the Kissinger method. The ultimate aim of TGA studies is to obtain a kinetic model of the gas-evolving reactions for use in mathematical modeling of the cold cap as an element of the overall model of the waste-glass melter. In this study, we focused on computing the kinetic parameters of individual reactions without identifying their actual chemistry, The rough provisional model presented is based on the first-order kinetics.

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

    SciTech Connect (OSTI)

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

    2013-09-30

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

  8. WRAP low level waste restricted waste management (LLW RWM) glovebox acceptance test report

    SciTech Connect (OSTI)

    Leist, K.J.

    1997-11-24

    On April 22, 1997, the Low Level Waste Restricted Waste Management (LLW RWM) glovebox was tested using acceptance test procedure 13027A-87. Mr. Robert L. Warmenhoven served as test director, Mr. Kendrick Leist acted as test operator and test witness, and Michael Lane provided miscellaneous software support. The primary focus of the glovebox acceptance test was to examine glovebox control system interlocks, operator Interface Unit (OIU) menus, alarms, and messages. Basic drum port and lift table control sequences were demonstrated. OIU menus, messages, and alarm sequences were examined, with few exceptions noted. Barcode testing was bypassed, due to the lack of installed equipment as well as the switch from basic reliance on fixed bar code readers to the enhanced use of portable bar code readers. Bar code testing was completed during performance of the LLW RWM OTP. Mechanical and control deficiencies were documented as Test Exceptions during performance of this Acceptance Test. These items are attached as Appendix A to this report.

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

    SciTech Connect (OSTI)

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

    2011-12-29

    This document provides the final report on data and results obtained from a series of nine tests performed on the one-third scale DuraMelter{trademark} 1200 (DM1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part B1 [1]. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plans. The nine tests reported here were preceded by an initial series of short-duration tests conducted to support the start-up and commissioning of this system. This report is a followup to the previously issued Preliminary Data Summary Reports. The DM1200 system was deployed for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. These tests include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The primary objective of the present series of tests was to determine the effects of a variety of parameters on the glass production rate in comparison to the RPP-WTP HL W design basis of 400 kg/m{sup 2}/d. Previous testing on the DMIOOO system [1] concluded that achievement of that rate with simulants of projected WTP melter feeds (AZ-101 and C-106/AY-102) was unlikely without the use of bubblers. As part of those tests, the same feed that was used during the cold-commissioning of the West Valley Demonstration Project (WVDP) HLW vitrification system was run on the DM1000 system. The DM1000 tests reproduced the rates that were obtained at the larger WVDP facility, lending confidence to the tests results [1]. Since the inclusion or exclusion of a bubbler has significant design implications, the Project commissioned further tests to address this issue. In an effort to identify factors that might increase the glass production rate for projected WTP melter feeds, a subsequent series of tests was performed on the DM100 system. Several tests variables led to glass production rate increases to values significantly above the 400 kg/m2/d requirement. However, while small-scale melter tests are useful for screening relative effects, they tend to overestimate absolute glass production rates, particularly for un-bubbled tests. Consequently, when scale-up effects were taken into account, it was not clear that any of the variables investigated would conclusively meet the 400 kg/m{sup 2}/d requirement without bubbling. The present series of tests was therefore performed on the DM1200 one-third scale HLW pilot melter system to provide the required basis for a final decision on whether bubblers would be included in the HLW melter. The present tests employed the same AZ-101 waste simulant and glass composition that was used for previous testing for consistency and comparability with the results from the earlier tests.

  10. Low-level waste certification plan for the WSCF Laboratory Complex

    SciTech Connect (OSTI)

    Morrison, J.A.

    1994-09-19

    The solid, low-level waste certification plan for the Waste Sampling and Characterization Facility (WSCF) describes the organization and methodology for the certification of the solid low-level waste (LLW) that is transferred to the Hanford Site 200 Areas Storage and Disposal Facilities. This plan incorporates the applicable elements of waste reduction, including up-front minimization, and end product treatment to reduce the volume or toxicity of the waste. The plan also includes segregation of different waste types. This low-level waste certification plan applies only to waste generated in, or is the responsibility of the WSCF Laboratory Complex. The WSCF Laboratory Complex supports technical activities performed at the Hanford Site. Wet Chemical and radiochemical analyses are performed to support site operations, including environmental and effluent monitoring, chemical processing, RCRA and CERCLA analysis, and waste management activities. Environmental and effluent samples include liquid effluents, ground and surface waters, soils, animals, vegetation, and air filters.

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

    SciTech Connect (OSTI)

    D.C. Richardson

    2003-03-19

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

  12. The incorporation of P, S, Cr, F, Cl, I, Mn, Ti, U, and Bi into simulated nuclear waste glasses: Literature study

    SciTech Connect (OSTI)

    Langowski, M.H.

    1996-02-01

    Waste currently stored on the Hanford Reservation in underground tanks will be into High Level Waste (HLW) and Low Level Waste (LLW). The HLW melter will high-level and transuranic wastes to a vitrified form for disposal in a geological repository. The LLW melter will vitrify the low-level waste which is mainly a sodium solution. Characterization of the tank wastes is still in progress, and the pretreatment processes are still under development Apart from tank-to-tank variations, the feed delivered to the HLW melter will be subject to process control variability which consists of blending and pretreating the waste. The challenge is then to develop glass formulation models which can produce durable and processable glass compositions for all potential vitrification feed compositions and processing conditions. The work under HLW glass formulation is to study and model glass and melt pro functions of glass composition and temperature. The properties of interest include viscosity, electrical conductivity, liquidus temperature, crystallization, immiscibility durability. It is these properties that determine the glass processability and ac waste glass. Apart from composition, some properties, such as viscosity are affected by temperature. The processing temperature may vary from 1050{degrees}C to 1550{degrees}C dependent upon the melter type. The glass will also experience a temperature profile upon cooling. The purpose of this letter report is to assess the expected vitrification feed compositions for critical components with the greatest potential impact on waste loading for double shell tank (DST) and single shell tank (SST) wastes. The basis for critical component selection is identified along with the planned approach for evaluation. The proposed experimental work is a crucial part of model development and verification.

  13. Advances in the Glass Formulations for the Hanford Tank Waste Treatment and Immobilization Plant

    SciTech Connect (OSTI)

    Kruger, Albert A.; Vienna, John D.; Kim, Dong Sang

    2015-01-14

    The Department of Energy-Office of River Protection (DOE-ORP) is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to treat radioactive waste currently stored in underground tanks at the Hanford site in Washington. The WTP that is being designed and constructed by a team led by Bechtel National, Inc. (BNI) will separate the tank waste into High Level Waste (HLW) and Low Activity Waste (LAW) fractions with the majority of the mass (~90%) directed to LAW and most of the activity (>95%) directed to HLW. The pretreatment process, envisioned in the baseline, involves the dissolution of aluminum-bearing solids so as to allow the aluminum salts to be processed through the cesium ion exchange and report to the LAW Facility. There is an oxidative leaching process to affect a similar outcome for chromium-bearing wastes. Both of these unit operations were advanced to accommodate shortcomings in glass formulation for HLW inventories. A by-product of this are a series of technical challenges placed upon materials selected for the processing vessels. The advances in glass formulation play a role in revisiting the flow sheet for the WTP and hence, the unit operations that were being imposed by minimal waste loading requirements set forth in the contract for the design and construction of the plant. Another significant consideration to the most recent revision of the glass models are the impacts on resolution of technical questions associated with current efforts for design completion.

  14. Selected radionuclides important to low-level radioactive waste management

    SciTech Connect (OSTI)

    1996-11-01

    The purpose of this document is to provide information to state representatives and developers of low level radioactive waste (LLW) management facilities about the radiological, chemical, and physical characteristics of selected radionuclides and their behavior in the environment. Extensive surveys of available literature provided information for this report. Certain radionuclides may contribute significantly to the dose estimated during a radiological performance assessment analysis of an LLW disposal facility. Among these are the radionuclides listed in Title 10 of the Code of Federal Regulations Part 61.55, Tables 1 and 2 (including alpha emitting transuranics with half-lives greater than 5 years). This report discusses these radionuclides and other radionuclides that may be significant during a radiological performance assessment analysis of an LLW disposal facility. This report not only includes essential information on each radionuclide, but also incorporates waste and disposal information on the radionuclide, and behavior of the radionuclide in the environment and in the human body. Radionuclides addressed in this document include technetium-99, carbon-14, iodine-129, tritium, cesium-137, strontium-90, nickel-59, plutonium-241, nickel-63, niobium-94, cobalt-60, curium -42, americium-241, uranium-238, and neptunium-237.

  15. H:\cindy_pratt\hlw rod.tif

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

    RECORD OF DECISION For The Idaho High- Level Waste and Facilities Disposition Final Environmental Impact Statement December 2005 United States Department of Energy 1 U.S. DEPARTMENT OF ENERGY Office of Environmental Management Record of Decision for the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement SUMMARY: DOE is making decisions pursuant to the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (Final EIS)

  16. Greater-than-Class C Low-Level Radioactive Waste (GTCC LLW) ...

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

    Disposal of Greater-Than-Class C (GTCC) Low-Level Radioactive Waste (LLRW) and GTCC-Like Waste (Draft EIS, DOEEIS-0375D) for public review and comment. DOE is inviting public...

  17. Evaluation of prospective hazardous waste treatment technologies for use in processing low-level mixed wastes at Rocky Flats

    SciTech Connect (OSTI)

    McGlochlin, S.C.; Harder, R.V.; Jensen, R.T.; Pettis, S.A.; Roggenthen, D.K.

    1990-09-18

    Several technologies for destroying or decontaminating hazardous wastes were evaluated (during early 1988) as potential processes for treating low-level mixed wastes destined for destruction in the Fluidized Bed Incinerator. The processes that showed promise were retained for further consideration and placed into one (or more) of three categories based on projected availability: short, intermediate, and long-term. Three potential short-term options were identified for managing low-level mixed wastes generated or stored at the Rocky Flats Plant (operated by Rockwell International in 1988). These options are: (1) Continue storing at Rocky Flats, (2) Ship to Nevada Test Site for landfill disposal, or (3) Ship to the Idaho National Engineering Laboratory for incineration in the Waste Experimental Reduction Facility. The third option is preferable because the wastes will be destroyed. Idaho National Engineering Laboratory has received interim status for processing solid and liquid low-level mixed wastes. However, low-level mixed wastes will continue to be stored at Rocky Flats until the Department of Energy approval is received to ship to the Nevada Test Site or Idaho National Engineering Laboratory. Potential intermediate and long-term processes were identified; however, these processes should be combined into complete waste treatment systems'' that may serve as alternatives to the Fluidized Bed Incinerator. Waste treatment systems will be the subject of later work. 59 refs., 2 figs.

  18. Letter report: Minor component study for low-level radioactive waste glasses

    SciTech Connect (OSTI)

    Li, H.

    1996-03-01

    During the waste vitrification process, troublesome minor components in low-level radioactive waste streams could adversely affect either waste vitrification rate or melter life-time. Knowing the solubility limits for these minor components is important to determine pretreatment options for waste streams and glass formulation to prevent or to minimize these problems during the waste vitrification. A joint study between Pacific Northwest Laboratory and Rensselaer Polytechnic Institute has been conducted to determine minor component impacts in low-level nuclear waste glass.

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

    SciTech Connect (OSTI)

    KETUSKY, EDWARD

    2006-04-20

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

  20. Calculates Neutron Production in Canisters of High-level Waste

    Energy Science and Technology Software Center (OSTI)

    1993-01-15

    ALPHN calculates the (alpha,n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the (alpha,n) neutron production of each actinide in neutrons per second and the total for the canister. The (alpha,n) neutron production rates are source terms only; that is, they are production rates within the glass andmore » do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister.« less

  1. WRAP low level waste (LLW) glovebox acceptance test report

    SciTech Connect (OSTI)

    Leist, K.J.

    1998-02-17

    In June 28, 1997, the Low Level Waste (LLW) glovebox was tested using glovebox acceptance test procedure 13031A-85. The primary focus of the glovebox acceptance test was to examine control system interlocks, display menus, alarms, and operator messages. Limited mechanical testing involving the drum ports, hoists, drum lifter, compacted drum lifter, drum tipper, transfer car, conveyors, lidder/delidder device and the supercompactor were also conducted. As of November 24, 1997, 2 of the 131 test exceptions that affect the LLW glovebox remain open. These items will be tracked and closed via the WRAP Master Test Exception Database. As part of Test Exception resolution/closure the responsible individual closing the Test Exception performs a retest of the affected item(s) to ensure the identified deficiency is corrected, and, or to test items not previously available to support testing. Test Exceptions are provided as appendices to this report.

  2. Evaluation Of The Integrated Solubility Model, A Graded Approach For Predicting Phase Distribution In Hanford Tank Waste

    SciTech Connect (OSTI)

    Pierson, Kayla L.; Belsher, Jeremy D.; Seniow, Kendra R.

    2012-10-19

    The mission of the DOE River Protection Project (RPP) is to store, retrieve, treat and dispose of Hanford's tank waste. Waste is retrieved from the underground tanks and delivered to the Waste Treatment and Immobilization Plant (WTP). Waste is processed through a pretreatment facility where it is separated into low activity waste (LAW), which is primarily liquid, and high level waste (HLW), which is primarily solid. The LAW and HLW are sent to two different vitrification facilities and glass canisters are then disposed of onsite (for LAW) or shipped off-site (for HLW). The RPP mission is modeled by the Hanford Tank Waste Operations Simulator (HTWOS), a dynamic flowsheet simulator and mass balance model that is used for mission analysis and strategic planning. The integrated solubility model (ISM) was developed to improve the chemistry basis in HTWOS and better predict the outcome of the RPP mission. The ISM uses a graded approach to focus on the components that have the greatest impact to the mission while building the infrastructure for continued future improvement and expansion. Components in the ISM are grouped depending upon their relative solubility and impact to the RPP mission. The solubility of each group of components is characterized by sub-models of varying levels of complexity, ranging from simplified correlations to a set of Pitzer equations used for the minimization of Gibbs Energy.

  3. Operational Strategies for Low-Level Radioactive Waste Disposal Site in Egypt - 13513

    SciTech Connect (OSTI)

    Mohamed, Yasser T.

    2013-07-01

    The ultimate aims of treatment and conditioning is to prepare waste for disposal by ensuring that the waste will meet the waste acceptance criteria of a disposal facility. Hence the purpose of low-level waste disposal is to isolate the waste from both people and the environment. The radioactive particles in low-level waste emit the same types of radiation that everyone receives from nature. Most low-level waste fades away to natural background levels of radioactivity in months or years. Virtually all of it diminishes to natural levels in less than 300 years. In Egypt, The Hot Laboratories and Waste Management Center has been established since 1983, as a waste management facility for LLW and ILW and the disposal site licensed for preoperational in 2005. The site accepts the low level waste generated on site and off site and unwanted radioactive sealed sources with half-life less than 30 years for disposal and all types of sources for interim storage prior to the final disposal. Operational requirements at the low-level (LLRW) disposal site are listed in the National Center for Nuclear Safety and Radiation Control NCNSRC guidelines. Additional procedures are listed in the Low-Level Radioactive Waste Disposal Facility Standards Manual. The following describes the current operations at the LLRW disposal site. (authors)

  4. National low-level waste management program radionuclide report series, Volume 15: Uranium-238

    SciTech Connect (OSTI)

    Adams, J.P.

    1995-09-01

    This report, Volume 15 of the National Low-Level Waste Management Program Radionuclide Report Series, discusses the radiological and chemical characteristics of uranium-238 ({sup 238}U). The purpose of the National Low-Level Waste Management Program Radionuclide Report Series is to provide information to state representatives and developers of low-level radioactive waste disposal facilities about the radiological, chemical, and physical characteristics of selected radionuclides and their behavior in the waste disposal facility environment. This report also includes discussions about waste types and forms in which {sup 238}U can be found, and {sup 238}U behavior in the environment and in the human body.

  5. Process system evaluation-consolidated letters. Volume 1. Alternatives for the off-gas treatment system for the low-level waste vitrification process

    SciTech Connect (OSTI)

    Peurrung, L.M.; Deforest, T.J; Richards, J.R.

    1996-03-01

    This report provides an evaluation of alternatives for treating off-gas from the low-level waste (LLW) melter. The study used expertise obtained from the commercial nonradioactive off-gas treatment industry. It was assumed that contact maintenance is possible, although the subsequent risk to maintenance personnel was qualitatively considered in selecting equipment. Some adaptations to the alternatives described may be required, depending on the extent of contact maintenance that can be achieved. This evaluation identified key issues for the off-gas system design. To provide background information, technology reviews were assembled for various classifications of off-gas treatment equipment, including off-gas cooling, particulate control, acid gas control, mist elimination, NO{sub x} reduction, and SO{sub 2} removal. An order-of-magnitude cost estimate for one of the off-gas systems considered is provided using both the off-gas characteristics associated with the Joule-heated and combustion-fired melters. The key issues identified and a description of the preferred off-gas system options are provided below. Five candidate treatment systems were evaluated. All of the systems are appropriate for the different melting/feed preparations currently being considered. The lowest technical risk is achieved using option 1, which is similar to designs for high-level waste (HLW) vitrification in the Hanford Waste Vitrification Project (HWVP) and the West Valley. Demonstration Project. Option 1 uses a film cooler, submerged bed scrubber (SBS), and high-efficiency mist eliminator (HEME) prior to NO{sub x} reduction and high-efficiency particulate air (HEPA) filtration. However, several advantages were identified for option 2, which uses high-temperature filtration. Based on the evaluation, option 2 was identified as the preferred alternative. The characteristics of this option are described below.

  6. Bis-diglycol-amides (Bis-DGA) as new extractants for An(III) and Ln(III) from aqueous high-level wastes issued from the Purex process

    SciTech Connect (OSTI)

    Espartero, A.G.; Murillo, M.T.; Almaraz, M.; Sanchez-Quesada, J.; Iglesias-Sanchez, J.C.; Prados, P.; Segura, M.; Mendoza, J. de

    2008-07-01

    A new family of compounds with two diglycolamide moieties in their molecule have been synthesized and studied as suitable extractants for trivalent actinides (An(III)) and trivalent lanthanides (Ln(III)) present in high-level wastes (HLW) issued from the PUREX process. Although the obtained distribution ratios are comparable with those from TODGA under similar experimental conditions, the bis-DGA compounds showed higher selectivity towards Ln(III). The number of bis-DGA molecules involved in the formation of the dominant complex is two, and it is possible to recover more than 99% of the extracted An and Ln with 0.01 M nitric acid in order to recycle the solvent in subsequent extraction cycles. (authors)

  7. Tank Waste Corporate Board Meeting 11/06/08 | Department of Energy

    Energy Savers [EERE]

    06/08 Tank Waste Corporate Board Meeting 11/06/08 The following documents are associated with the Tank Waste Corporate Board Meeting held on November 6th, 2008. Note: (Please contact Steven Ross at steven.ross@em.doe.gov for a HLW Glass Waste Loadings version with animations on slide 6). PDF icon Slurry Retrieval, Pipeline Transport & Plugging and Mixing Workshop PDF icon The Way Ahead - West Valley Demonstration Project PDF icon High-Level Liquid Waste Tank Integrity Workshop - 2008 PDF

  8. Final Report - Management of High Sulfur HLW, VSL-13R2920-1, Rev. 0, dated 10/31/2013

    SciTech Connect (OSTI)

    Kruger, Albert A.; Gan, H.; Pegg, I. L.; Feng, Z.; Gan, H,; Joseph, I.; Matlack, K. S.

    2013-11-13

    The present report describes results from a series of small-scale crucible tests to determine the extent of corrosion associated with sulfur containing HLW glasses and to develop a glass composition for a sulfur-rich HLW waste stream, which was then subjected to small-scale melter testing to determine the maximum acceptable sulfate loadings. In the present work, a new glass formulation was developed and tested for a projected Hanford HLW composition with sulfate concentrations high enough to limit waste loading. Testing was then performed on the DM10 melter system at successively higher waste loadings to determine the maximum waste loading without the formation of a separate sulfate salt phase. Small scale corrosion testing was also conducted using the glass developed in the present work, the glass developed in the initial phase of this work [26], and a high iron composition, all at maximum sulfur concentrations determined from melter testing, in order to assess the extent of Inconel 690 and MA758 corrosion at elevated sulfate contents.

  9. Low-level Waste Forum meeting report. Fall meeting, October 20--22, 1993

    SciTech Connect (OSTI)

    1993-12-31

    The Low-Level Radioactive Waste Forum is an association of representatives of states and compacts established to facilitate state and compact commission implementation of the Low-Level Radioactive Waste Policy Act of 1980 and the Low-Level Radioactive Waste Policy Amendments Act of 1985 and to promote the objectives of low-level radioactive waste regional compacts. The Forum provides an opportunity for states and compacts to share information with one another and to exchange views with officials of federal agencies. The Forum participants include representatives from regional compacts, designated host states, unaffiliated states, and states with currently-operating low-level radioactive waste facilities. This report contains information synthesizing the accomplishments of the Forum, as well as any new advances that have been made in the management of low-level radioactive wastes.

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

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

    Department of Energy of Low-Level Waste at the Nevada National Security Site Disposal of Low-Level Waste at the Nevada National Security Site PDF icon Disposal of Low-Level Waste at the Nevada National Security Site More Documents & Publications Keynote Address: Update on Environmental Management Communication Is Key to Packaging and Transportation Safety and Compliance Status Updates on the Performance and Risk Assessment Community of Practice

  11. The effect of high-level waste glass composition on spinel liquidus

    Office of Scientific and Technical Information (OSTI)

    temperature (Journal Article) | SciTech Connect Journal Article: The effect of high-level waste glass composition on spinel liquidus temperature Citation Details In-Document Search Title: The effect of high-level waste glass composition on spinel liquidus temperature Spinel crystals precipitate in high-level waste glasses containing Fe, Cr, Ni , Mn, Zn, and Ru. The liquidus temperature (TL) of spinel as the primary crystallization phase is a function of glass composition and the spinel

  12. Time of Compliance for Disposal of Low-Level Radioactive Waste | Department

    Energy Savers [EERE]

    of Energy Time of Compliance for Disposal of Low-Level Radioactive Waste Time of Compliance for Disposal of Low-Level Radioactive Waste Roger Seitz*, Savannah River National Laboratory ; Andrew Wallo, U.S. Department of Energy Abstract: The United States Department of Energy (DOE) has more than 25 years of experience conducting and overseeing performance assessments (PAs) for low-level waste (LLW) and mixed LLW from on-going operations, decommissioning and environmental restoration

  13. Experimental data and analysis to support the design of an ion-exchange process for the treatment of Hanford tank waste supernatant liquids

    SciTech Connect (OSTI)

    Kurath, D.E.; Bray, L.A.; Brooks, K.P.; Brown, G.N.; Bryan, S.A.; Carlson, C.D.; Carson, K.J.; DesChane, J.R.; Elovich, R.J.; Kim, A.Y.

    1994-12-01

    Hanford`s 177 underground storage tanks contain a mixture of sludge, salt cake, and alkaline supernatant liquids. Disposal options for these wastes are high-level waste (HLW) glass for disposal in a repository or low-level waste (LLW) glass for onsite disposal. Systems-engineering studies show that economic and environmental considerations preclude disposal of these wastes without further treatment. Difficulties inherent in transportation and disposal of relatively large volumes of HLW make it impossible to vitrify all of the tank waste as HLW. Potential environmental impacts make direct disposal of all of the tank waste as LLW glass unacceptable. Although the pretreatment and disposal requirements are still being defined, most pretreatment scenarios include retrieval of the aqueous liquids, dissolution of the salt cakes, and washing of the sludges to remove soluble components. Most of the cesium is expected to be in the aqueous liquids, which are the focus of this report on cesium removal by ion exchange. The main objectives of the ion-exchange process are removing cesium from the bulk of the tank waste (i.e., decontamination) and concentrating the separated cesium for vitrification. Because exact requirements for removal of {sup 137}Cs have not yet been defined, a range of removal requirements will be considered. This study addresses requirements to achieve {sup 137}Cs levels in LLW glass between (1) the Nuclear Regulatory Commission (NRC) Class C (10 CFR 61) limit of 4600 Ci/m{sup 3} and (2) 1/10th of the NRC Class A limit of 1 Ci/m{sup 3} i.e., 0.1/m{sup 3}. The required degrees of separation of cesium from other waste components is a complex function involving interactions between the design of the vitrification process, waste form considerations, and other HLW stream components that are to be vitrified.

  14. Evaluation and compilation of DOE waste package test data: Biannual report, August 1987--January 1988

    SciTech Connect (OSTI)

    Interrante, C.; Escalante, E.; Fraker, A.; Ondik, H.; Plante, E.; Ricker, R.; Ruspi, J.

    1988-08-01

    This report summarizes results of the National Bureau of Standards (NBS) evaluations on waste packages designed for containment of radioactive high-level nuclear waste (HLW). The waste package is a proposed engineered barrier that is part of a permanent repository for HLW. Metal alloys are the principal barriers within the engineered system. Since enactment of the Budget Reconciliation Act for Fiscal Year 1988, the Yucca Mountain, Nevada, site (in which tuff is the geologic medium) is the only site that will be characterized for use as high-level nuclear waste repository. During the reporting period of August 1987 to January 1988, five reviews were completed for tuff, and these were grouped into the categories: ferrous alloys, copper, groundwater chemistry, and glass. Two issues are identified for the Yucca Mountain site: the approach used to calculate corrosion rates for ferrous alloys, and crevice corrosion was observed in a copper-nickel alloy. Plutonium can form pseudo-colloids that may facilitate transport. NBS work related to the vitrification of HLW borosilicate glass at the West Valley Demonstration Project (WVDP) and the Defense Waste Processing Facility (DWPF) and activities of the DOE Materials Characterization Center (MCC) for the 6-month reporting period are also included. 27 refs., 3 figs.

  15. Northeast High-Level Radioactive Waste Transportation Task Force...

    Office of Environmental Management (EM)

    members & alternates appointment status Legislative Liaisons Staff ... (by speaker phone) 1:45 p.m. Update: Decommissioning Plant Coalition Nuclear Waste ...

  16. High-Level Waste Corporate Board Meeting Agenda

    Office of Environmental Management (EM)

    talks using flowsheet figure Gary Smith 9:15 AM Waste Treatment & Immobilization ... Presentations Recap & Conclusions Gary Smith 1:30 PM EM TEG Panel Discussion with ...

  17. 1996 annual report on low-level radioactive waste management progress. Report to Congress

    SciTech Connect (OSTI)

    1997-11-01

    This report is prepared in response to the Low-Level Radioactive Waste Policy Act (the Act), Public Law 96-573, 1980, as amended by the Low-Level Radioactive Waste Policy Amendments Act of 1985, Public Law 99-240. The report summarizes the activities during calendar year 1996 related to the establishment of new disposal facilities for commercially-generated low-level radioactive waste. The report emphasizes significant issues and events that have affected progress in developing new disposal facilities, and also includes an introduction that provides background information and perspective on US policy for low-level radioactive waste disposal.

  18. Report to Congress: 1995 Annual report on low-level radioactive waste management progress

    SciTech Connect (OSTI)

    1996-06-01

    This report is prepared in response to the Low-Level Radioactive Waste Policy Act, Public Law 96-573, 1980, as amended by the Low-Level Radioactive Waste Policy Amendments Act of 1985, Public Law 99-240. The report summarizes the progress of states and compact regions during calendar year 1995 in establishing new disposal facilities for commercially-generated low-level radioactive waste. The report emphasizes significant issues and events that have affected progress, and also includes an introduction that provides background information and perspective on United States policy for low-level radioactive waste disposal.

  19. The effect of high-level waste glass composition on spinel liquidus...

    Office of Scientific and Technical Information (OSTI)

    on spinel liquidus temperature Citation Details In-Document Search Title: The effect of high-level waste glass composition on spinel liquidus temperature Spinel crystals ...

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

    Broader source: Energy.gov [DOE]

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

  1. Canister storage building evaluation of nuclear safety for solidified high-level waste transfer and storage

    SciTech Connect (OSTI)

    Kidder, R.J., Westinghouse Hanford

    1996-09-17

    This document is issued to evaluate the safety impacts to the Canister Storage Building from transfer and storage of solidified high-level waste.

  2. EIS-0375: Disposal of Greater-than-Class-C Low-Level Radioactive Waste and

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

    Department of Energy GTCC-like Waste | Department of Energy 5: Disposal of Greater-than-Class-C Low-Level Radioactive Waste and Department of Energy GTCC-like Waste EIS-0375: Disposal of Greater-than-Class-C Low-Level Radioactive Waste and Department of Energy GTCC-like Waste Summary This EIS evaluates the reasonably foreseeable environmental impacts associated with the proposed development, operation, and long-term management of a disposal facility or facilities for Greater-Than-Class C

  3. EIS-0375: Disposal of Greater-than-Class-C Low-Level Radioactive Waste and

    Office of Environmental Management (EM)

    Department of Energy GTCC-like Waste | Department of Energy 5: Disposal of Greater-than-Class-C Low-Level Radioactive Waste and Department of Energy GTCC-like Waste EIS-0375: Disposal of Greater-than-Class-C Low-Level Radioactive Waste and Department of Energy GTCC-like Waste Summary This EIS evaluates the reasonably foreseeable environmental impacts associated with the proposed development, operation, and long-term management of a disposal facility or facilities for Greater-Than-Class C

  4. Low-Level Radioactive Waste Management at the Nevada Test Site - Current Status

    SciTech Connect (OSTI)

    Bruce D. Becker, Bechtel Nevada; Bruce M. Crowe, Los Alamos National Laboratory; Carl P. Gertz, DOE Nevada Operations Office; Wendy A. Clayton, DOE Nevada Operations Office

    1999-02-01

    The performance objective of the Department of Energy's Low-Level Radioactive Waste disposal facility at the Nevada Test Site transcends those of any other radioactive waste disposal site in the United States. This paper describes the technical attributes of the facility, present and future capacities and capabilities, and provides a description of the process from waste approval to final disposition. The paper also summarizes the current status of the waste disposal operations.

  5. Low-Level Radioactive Waste Management at the Nevada Test Site - Year 2000 Current Status

    SciTech Connect (OSTI)

    Bruce D. Becker, Bechtel Nevada; Bruce M. Crowe, Los Alamos National Laboratory; Carl P. Gertz, DOE Nevada; Wendy A. Clayton, DOE Nevada

    1999-08-06

    The performance objectives of the Department of Energy's Low-level radioactive waste disposal facilities at the Nevada Test Site transcend those of any other radioactive waste disposal site in the United States. The expanded paper will describe the technical attributes of the facilities, the present and the future disposal capacities and capabilities, and includes a description of the process from waste approval to final disposition. The paper also summarizes the current status of the waste disposal operations.

  6. Proposed research and development plan for mixed low-level waste forms

    SciTech Connect (OSTI)

    O`Holleran, T.O.; Feng, X.; Kalb, P.

    1996-12-01

    The objective of this report is to recommend a waste form program plan that addresses waste form issues for mixed low-level waste (MLLW). The report compares the suitability of proposed waste forms for immobilizing MLLW in preparation for permanent near-surface disposal and relates them to their impact on the U.S. Department of Energy`s mixed waste mission. Waste forms are classified into four categories: high-temperature waste forms, hydraulic cements, encapsulants, and specialty waste forms. Waste forms are evaluated concerning their ability to immobilize MLLW under certain test conditions established by regulatory agencies and research institutions. The tests focused mainly on leach rate and compressive strength. Results indicate that all of the waste forms considered can be tailored to give satisfactory performance immobilizing large fractions of the Department`s MLLW inventory. Final waste form selection will ultimately be determined by the interaction of other, often nontechnical factors, such as economics and politics. As a result of this report, three top-level programmatic needs have been identified: (1) a basic set of requirements for waste package performance and disposal; (2) standardized tests for determining waste form performance and suitability for disposal; and (3) engineering experience operating production-scale treatment and disposal systems for MLLW.

  7. Advances in Glass Formulations for Hanford High-Alumimum, High-Iron and Enhanced Sulphate Management in HLW Streams - 13000

    SciTech Connect (OSTI)

    Kruger, Albert A.

    2013-01-16

    The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet Hanford Tank Waste Treatment and Immobilization Plant (WTP) Contract terms. The WTP?s overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulphur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings and higher throughput efficiencies. Results of this work have demonstrated the feasibility of increases in waste loading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. In view of the importance of aluminum limited waste streams at Hanford (and also Savannah River), the ability to achieve high waste loadings without adversely impacting melt rates has the potential for enormous cost savings from reductions in canister count and the potential for schedule acceleration. Consequently, the potential return on the investment made in the development of these enhancements is extremely favorable. Glass composition development for one of the latest Hanford HLW projected compositions with sulphate concentrations high enough to limit waste loading have been successfully tested and show tolerance for previously unreported tolerance for sulphate. Though a significant increase in waste loading for high-iron wastes has been achieved, the magnitude of the increase is not as substantial as those achieved for high-aluminum, high-chromium, high-bismuth or sulphur. Waste processing rate increases for high-iron streams as a combined effect of higher waste loadings and higher melt rates resulting from new formulations have been achieved.

  8. Hanford low-level waste process chemistry testing data package

    SciTech Connect (OSTI)

    Smith, H.D.; Tracey, E.M.; Darab, J.G.; Smith, P.A.

    1996-03-01

    Recently, the Tri-Party Agreement (TPA) among the State of Washington Department of Ecology, U.S. Department of Energy (DOE) and the US Environmental Protection Agency (EPA) for the cleanup of the Hanford Site was renegotiated. The revised agreement specifies vitrification as the encapsulation technology for low level waste (LLW). A demonstration, testing, and evaluation program underway at Westinghouse Hanford Company to identify the best overall melter-system technology available for vitrification of Hanford Site LLW to meet the TPA milestones. Phase I is a {open_quotes}proof of principle{close_quotes} test to demonstrate that a melter system can process a simulated highly alkaline, high nitrate/nitrite content aqueous LLW feed into a glass product of consistent quality. Seven melter vendors were selected for the Phase I evaluation: joule-heated melters from GTS Duratek, Incorporated (GDI); Envitco, Incorporated (EVI); Penberthy Electomelt, Incorporated (PEI); and Vectra Technologies, Incorporated (VTI); a gas-fired cyclone burner from Babcock & Wilcox (BCW); a plasma torch-fired, cupola furnace from Westinghouse Science and Technology Center (WSTC); and an electric arc furnace with top-entering vertical carbon electrodes from the U.S. Bureau of Mines (USBM).

  9. Interim data quality objectives for waste pretreatment and vitrification. Revision 1

    SciTech Connect (OSTI)

    Kupfer, M.J.; Conner, J.M.; Kirkbride, R.A.; Mobley, J.R.

    1994-09-15

    The Tank Waste Remediation System (TWRS) is responsible for storing, processing, and immobilizing the Hanford Site tank wastes. Characterization information on the tank wastes is needed so that safety concerns can be addressed, and retrieval, pretreatment, and immobilization processes can be designed, permitted, and implemented. This document describes the near-term tank waste sampling and characterization needs of the Pretreatment, High-Level Waste (HLW) Disposal, and Low-Level Waste (LLW) Disposal Programs to support the TWRS disposal mission. The final DQO (Data Quality Objective) will define specific waste tanks to be sampled, sample timing requirements, an appropriate analytical scheme, and a list of required analytes. This interim DQO, however, focuses primarily on the required analytes since the tanks to be sampled in FY 1994 and early FY 1995 are being driven most heavily by other considerations, particularly safety. The major objective of this Interim DQO is to provide guidance for tank waste characterization requirements for samples taken before completion of the final DQO. The characterization data needs defined herein will support the final DQO to help perform the following: Support the TWRS technical strategy by identification of the chemical and physical composition of the waste in the tanks and Guide development efforts to define waste pretreatment processes, which will in turn define HLW and LLW feed to vitrification processes.

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

    SciTech Connect (OSTI)

    McDonell, W R

    1982-04-01

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

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

    SciTech Connect (OSTI)

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

    2011-01-04

    The solubility of plutonium in a Sludge Batch 6 (SB6) reference glass and the effect of incorporation of Pu in the glass on specific glass properties were evaluated. A Pu loading of 1 wt % in glass was studied. Prior to actual plutonium glass testing, surrogate testing (using Hf as a surrogate for Pu) was conducted to evaluate the homogeneity of significant quantities of Hf (Pu) in the glass, determine the most appropriate methods to evaluate homogeneity for Pu glass testing, and to evaluate the impact of Hf loading in the glass on select glass properties. Surrogate testing was conducted using Hf to represent between 0 and 1 wt % Pu in glass on an equivalent molar basis. A Pu loading of 1 wt % in glass translated to {approx}18 kg Pu per Defense Waste Processing Facility (DWPF) canister, or about 10X the current allowed limit per the Waste Acceptance Product Specifications (2500 g/m{sup 3} of glass or about 1700 g/canister) and about 30X the current allowable concentration based on the fissile material concentration limit referenced in the Yucca Mountain Project License Application (897 g/m{sup 3}3 of glass or about 600 g Pu/canister). Based on historical process throughput data, this level was considered to represent a reasonable upper bound for Pu loading based on the ability to provide Pu containing feed to the DWPF. The task elements included evaluating the distribution of Pu in the glass (e.g. homogeneity), evaluating crystallization within the glass, evaluating select glass properties (with surrogates), and evaluating durability using the Product Consistency Test -- Method A (PCT-A). The behavior of Pu in the melter was evaluated using paper studies and corresponding analyses of DWPF melter pour samples.The results of the testing indicated that at 1 wt % Pu in the glass, the Pu was homogeneously distributed and did not result in any formation of plutonium-containing crystalline phases as long as the glass was prepared under 'well-mixed' conditions. The incorporation of 1 wt % Pu in the glass did not adversely impact glass viscosity (as assessed using Hf surrogate) or glass durability. Finally, evaluation of DWPF glass pour samples that had Pu concentrations below the 897 g/m{sup 3} limit showed that Pu concentrations in the glass pour stream were close to targeted compositions in the melter feed indicating that Pu neither volatilized from the melt nor stratified in the melter when processed in the DWPF melter.

  12. Environmental assessment for the treatment of Class A low-level radioactive waste and mixed low-level waste generated by the West Valley Demonstration Project

    SciTech Connect (OSTI)

    NONE

    1995-11-01

    The U.S. Department of Energy (DOE) is currently evaluating low-level radioactive waste management alternatives at the West Valley Demonstration Project (WVDP) located on the Western New York Nuclear Service Center (WNYNSC) near West Valley, New York. The WVDP`s mission is to vitrify high-level radioactive waste resulting from commercial fuel reprocessing operations that took place at the WNYNSC from 1966 to 1972. During the process of high-level waste vitrification, low-level radioactive waste (LLW) and mixed low-level waste (MILLW) will result and must be properly managed. It is estimated that the WVDP`s LLW storage facilities will be filled to capacity in 1996. In order to provide sufficient safe storage of LLW until disposal options become available and partially fulfill requirements under the Federal Facilities Compliance Act (FFCA), the DOE is proposing to use U.S. Nuclear Regulatory Commission-licensed and permitted commercial facilities in Oak Ridge, Tennessee; Clive, Utah; and Houston, Texas to treat (volume-reduce) a limited amount of Class A LLW and MLLW generated from the WVDP. Alternatives for ultimate disposal of the West Valley LLW are currently being evaluated in an environmental impact statement. This proposed action is for a limited quantity of waste, over a limited period of time, and for treatment only; this proposal does not include disposal. The proposed action consists of sorting, repacking, and loading waste at the WVDP; transporting the waste for commercial treatment; and returning the residual waste to the WVDP for interim storage. For the purposes of this assessment, environmental impacts were quantified for a five-year operating period (1996 - 2001). Alternatives to the proposed action include no action, construction of additional on-site storage facilities, construction of a treatment facility at the WVDP comparable to commercial treatment, and off-site disposal at a commercial or DOE facility.

  13. Radiolytic gas generation from cement-based waste hosts for DOE low-level radioactive wastes

    SciTech Connect (OSTI)

    Dole, L.R.; Friedman, H.A.

    1986-01-01

    Using cement-based immobilization binders with simulated radioactive waste containing sulfate, nitrate, nitrite, phosphate, and fluoride anions, the gamma- and alpha-radiolytic gas generation factors (G/sub t/, molecules/100 eV) and gas compositions were measured on specimens of cured grouts. These tests studied the effects of; (1) waste composition; (2) the sample surface-to-volume ratio; (3) the waste slurry particle size; and (4) the water content of the waste host formula. The radiolysis test vessels were designed to minimize the ''dead'' volume and to simulate the configuration of waste packages.

  14. Environmental Assessment Idaho National Engineering Laboratory, low-level and mixed waste processing

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    The Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA-0843, for the Idaho National Engineering Laboratory (INEL) low-level and mixed waste processing. The original proposed action, as reviewed in this EA, was (1) to incinerate INEL`s mixed low-level waste (MLLW) at the Waste Experimental Reduction Facility (WERF); (2) reduce the volume of INEL generated low-level waste (LLW) through sizing, compaction, and stabilization at the WERF; and (3) to ship INEL LLW to a commercial incinerator for supplemental LLW volume reduction.

  15. Synergistic Inhibitors for Dilute High-Level Radioactive Waste

    SciTech Connect (OSTI)

    Wiersma, B.J.; Zapp, P.E.

    1995-11-01

    Cyclic potentiodynamic polarization scans were conducted to determine the effectiveness of various combinations of anodic inhibitors in the prevention of pitting in carbon steel exposed to dilute radioactive waste. Chromate, molybdate, and phosphate were investigated as replacements for nitrite, whose effective concentrations are incompatible with the waste vitrification process. The polarization scans were performed in non-radioactive waste simulants. Their results showed that acceptable combinations of phosphate with chromate and phosphate with molybdate effectively prevented pitting corrosion. Chromate with molybdate could not replace nitrite.

  16. B Plant complex hazardous, mixed and low level waste certification plan

    SciTech Connect (OSTI)

    Beam, T.G.

    1994-11-01

    This plan describes the administrative steps and handling methodology for certification of hazardous waste, mixed waste, and low level waste generated at B Plant Complex. The plan also provides the applicable elements of waste reduction and pollution prevention, including up front minimization and end product reduction of volume and/or toxicity. The plan is written to satisfy requirements for Hanford Site waste generators to have a waste certification program in place at their facility. This plan, as described, applies only to waste which is generated at, or is the responsibility of, B Plant Complex. The scope of this plan is derived from the requirements found in WHC-EP-0063, Hanford Site Solid Waste Acceptance Criteria.

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

    SciTech Connect (OSTI)

    Kitchen, B.G.

    1989-08-23

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

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

    SciTech Connect (OSTI)

    Kitchen, B.G.

    1989-08-23

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

  19. Preliminary Safety Design Report for Remote Handled Low-Level Waste Disposal Facility

    SciTech Connect (OSTI)

    Timothy Solack; Carol Mason

    2012-03-01

    A new onsite, remote-handled low-level waste disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled low-level waste disposal for remote-handled low-level waste from the Idaho National Laboratory and for nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled low-level waste in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This preliminary safety design report supports the design of a proposed onsite remote-handled low-level waste disposal facility by providing an initial nuclear facility hazard categorization, by discussing site characteristics that impact accident analysis, by providing the facility and process information necessary to support the hazard analysis, by identifying and evaluating potential hazards for processes associated with onsite handling and disposal of remote-handled low-level waste, and by discussing the need for safety features that will become part of the facility design.

  20. Low-level tank waste simulant data base

    SciTech Connect (OSTI)

    Lokken, R.O.

    1996-04-01

    The majority of defense wastes generated from reprocessing spent N- Reactor fuel at Hanford are stored in underground Double-shell Tanks (DST) and in older Single-Shell Tanks (SST) in the form of liquids, slurries, sludges, and salt cakes. The tank waste remediation System (TWRS) Program has the responsibility of safely managing and immobilizing these tank wastes for disposal. This report discusses three principle topics: the need for and basis for selecting target or reference LLW simulants, tanks waste analyses and simulants that have been defined, developed, and used for the GDP and activities in support of preparing and characterizing simulants for the current LLW vitrification project. The procedures and the data that were generated to characterized the LLW vitrification simulants were reported and are presented in this report. The final section of this report addresses the applicability of the data to the current program and presents recommendations for additional data needs including characterization and simulant compositional variability studies.

  1. 18th U.S. Department of Energy Low-Level Radioactive Waste Management Conference. Program

    SciTech Connect (OSTI)

    1997-05-20

    This conference explored the latest developments in low-level radioactive waste management through presentations from professionals in both the public and the private sectors and special guests. The conference included two continuing education seminars, a workshop, exhibits, and a tour of Envirocare of Utah, Inc., one of America's three commercial low-level radioactive waste depositories.

  2. Project report for the commercial disposal of mixed low-level waste debris

    SciTech Connect (OSTI)

    Andrews, G.; Balls, V.; Shea, T.; Thiesen, T.

    1994-05-01

    This report summarizes the basis for the commercial disposal of Idaho National Engineering Laboratory (INEL) mixed low-level waste (MLLW) debris and the associated activities. Mixed waste is radioactive waste plus hazardous waste as defined by the Resource Conservation and Recovery Act (RCRA). The critical factors for this project were DOE 5820.2A exemption, contracting mechanism, NEPA documentation, sampling and analysis, time limitation and transportation of waste. This report also will provide a guide or a starting place for future use of Envirocare of Utah or other private sector disposal/treatment facilities, and the lessons learned during this project.

  3. Microwave energy for post-calcination treatment of high-level nuclear wastes

    SciTech Connect (OSTI)

    Gombert, D.; Priebe, S.J.; Berreth, J.R.

    1980-01-01

    High-level radioactive wastes generated from nuclear fuel reprocessing require treatment for effective long-term storage. Heating by microwave energy is explored in processing of two possible waste forms: (1) drying of a pelleted form of calcined waste; and (2) vitrification of calcined waste. It is shown that residence times for these processes can be greatly reduced when using microwave energy rather than conventional heating sources, without affecting product properties. Compounds in the waste and in the glass frit additives couple very well with the 2.45 GHz microwave field so that no special microwave absorbers are necessary.

  4. Disposal of low-level and mixed low-level radioactive waste during 1990

    SciTech Connect (OSTI)

    Not Available

    1993-08-01

    Isotopic inventories and other data are presented for low-level radioactive waste (LLW) and mixed LLW disposed (and occasionally stored) during calendar year 1990 at commercial disposal facilities and Department of Energy (DOE) sites. Detailed isotopic information is presented for the three commercial disposal facilities located near Barnwell, SC, Richland, WA, and Beatty, NV. Less information is presented for the Envirocare disposal facility located near Clive, UT, and for LLW stored during 1990 at the West Valley site. DOE disposal information is included for the Savannah River Site (including the saltstone facility), Nevada Test Site, Los Alamos National Laboratory, Idaho National Engineering Laboratory, Hanford Site, Y-12 Site, and Oak Ridge National Laboratory. Summary information is presented about stored DOE LLW. Suggestions are made about improving LLW disposal data.

  5. Underground storage tank integrated demonstration: Evaluation of pretreatment options for Hanford tank wastes

    SciTech Connect (OSTI)

    Lumetta, G.J.; Wagner, M.J.; Colton, N.G.; Jones, E.O.

    1993-06-01

    Separation science plays a central role inn the pretreatment and disposal of nuclear wastes. The potential benefits of applying chemical separations in the pretreatment of the radioactive wastes stored at the various US Department of Energy sites cover both economic and environmental incentives. This is especially true at the Hanford Site, where the huge volume (>60 Mgal) of radioactive wastes stored in underground tanks could be partitioned into a very small volume of high-level waste (HLW) and a relatively large volume of low-level waste (LLW). The cost associated with vitrifying and disposing of just the HLW fraction in a geologic repository would be much less than those associated with vitrifying and disposing of all the wastes directly. Futhermore, the quality of the LLW form (e.g., grout) would be improved due to the lower inventory of radionuclides present in the LLW stream. In this report, we present the results of an evaluation of the pretreatment options for sludge taken from two different single-shell tanks at the Hanford Site-Tanks 241-B-110 and 241-U-110 (referred to as B-110 and U-110, respectively). The pretreatment options examined for these wastes included (1) leaching of transuranic (TRU) elements from the sludge, and (2) dissolution of the sludge followed by extraction of TRUs and {sup 90}Sr. In addition, the TRU leaching approach was examined for a third tank waste type, neutralized cladding removal waste.

  6. Preliminary assessment of candidate immobilization technologies for retrieved single-shell tank wastes

    SciTech Connect (OSTI)

    Wiemers, K.D.; Mendel, J.E.; Kruger, A.A.; Bunnell, L.R.; Mellinger, G.B.

    1992-01-01

    This report describes the initial work that has been performed to select technologies for immobilization of wastes that may be retrieved from Hanford single-shell tanks (SSTs). Two classes of waste will require immobilization. One is the combined high-level waste/transuranic (HLW/TRU) fraction, the other the low-level waste (LLW) fraction. A number of potential immobilization technologies are identified for each class of waste. Immobilization technologies were initially selected based on a number of considerations, including (1) the waste loading that could likely be achieved within the constraint of producing acceptable waste forms, (2) process flexibility (primarily compatibility with anticipated waste variability), (3) process complexity, and (4) state of development. Immobilization technologies selected for further development include the following: for HLW/TRU waste -- borosilicate glass, lead-iron phosphate glass, glass-calcine composites, glass-ceramics, and cement based forms; for non-denitrated LLW -- grout, laxtex-modified concrete, and polyethylene; and for denitrated LLW -- silicate glass, phosphate glass, and clay calcination or tailored ceramic in various matrices.

  7. Process description and plant design for preparing ceramic high-level waste forms

    SciTech Connect (OSTI)

    Grantham, L.F.; McKisson, R.L.; Guon, J.; Flintoff, J.F.; McKenzie, D.E.

    1983-02-25

    The ceramics process flow diagram has been simplified and upgraded to utilize only two major processing steps - fluid-bed calcination and hot isostatic press consolidating. Full-scale fluid-bed calcination has been used at INEL to calcine high-level waste for 18 y; and a second-generation calciner, a fully remotely operated and maintained calciner that meets ALARA guidelines, started calcining high-level waste in 1982. Full-scale hot isostatic consolidation has been used by DOE and commercial enterprises to consolidate radioactive components and to encapsulate spent fuel elements for several years. With further development aimed at process integration and parametric optimization, the operating knowledge of full-scale demonstration of the key process steps should be rapidly adaptable to scale-up of the ceramic process to full plant size. Process flowsheets used to prepare ceramic and glass waste forms from defense and commercial high-level liquid waste are described. Preliminary layouts of process flow diagrams in a high-level processing canyon were prepared and used to estimate the preliminary cost of the plant to fabricate both waste forms. The estimated costs for using both options were compared for total waste management costs of SRP high-level liquid waste. Using our design, for both the ceramic and glass plant, capital and operating costs are essentially the same for both defense and commercial wastes, but total waste management costs are calculated to be significantly less for defense wastes using the ceramic option. It is concluded from this and other studies that the ceramic form may offer important advantages over glass in leach resistance, waste loading, density, and process flexibility. Preliminary economic calculations indicate that ceramics must be considered a leading candidate for the form to immobilize high-level wastes.

  8. Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study. National Low-Level Waste Management Program

    SciTech Connect (OSTI)

    Tyacke, M.

    1993-08-01

    This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood.

  9. Ohio State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-04-01

    The Ohio State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Ohio. The profile is the result of a survey of NRC licensees in Ohio. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Ohio.

  10. Massachusetts State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-03-12

    The Massachusetts State Briefing Book is one of a series of State briefing books on low-level radioactive waste management practices. It has been prepared to assist State and Federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Massachusetts. The profile is the result of a survey of NRC licensees in Massachusetts. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Massachusetts.

  11. Vermont State Briefing Book on low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-07-01

    The Vermont State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Vermont. The profile is the result of a survey of Nuclear Regulatory Commission licensees in Vermont. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may affect waste management practices in Vermont.

  12. New Jersey State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-04-01

    The New Jersey state Briefing Book is one of a series of State briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in New Jersey. The profile is the result of a survey of NRC licensees in New Jersey. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in New Jersey.

  13. Texas State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-08-01

    The Texas State Briefing Book is one of a series of state briefing books on low-level radioactivee waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Texas. The profile is the result of a survey of NRC licensees in Texas. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Texas.

  14. Kentucky State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-08-01

    The Kentucky State Briefing Book is one of a series of State briefing books on low-level radioactive waste management practices. It has been prepared to assist State and Federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Kentucky. The profile is the result of a survey of NRC licensees in Kentucky. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Kentucky.

  15. Radiation Effects in Nuclear Waste Materials

    SciTech Connect (OSTI)

    Weber, William J.

    2005-09-30

    The objective of this project is to develop a fundamental understanding of radiation effects in glasses and ceramics, as well as the influence of solid-state radiation effects on aqueous dissolution kinetics, which may impact the performance of nuclear waste forms and stabilized nuclear materials. This work provides the underpinning science to develop improved glass and ceramic waste forms for the immobilization and disposition of high-level tank waste, excess plutonium, plutonium residues and scrap, other actinides, and other nuclear waste streams. Furthermore, this work is developing develop predictive models for the performance of nuclear waste forms and stabilized nuclear materials. Thus, the research performed under this project has significant implications for the immobilization of High-Level Waste (HLW) and Nuclear Materials, two mission areas within the Office of Environmental Management (EM). With regard to the HLW mission, this research will lead to improved understanding of radiation-induced degradation mechanisms and their effects on dissolution kinetics, as well as development of predictive models for waste form performance. In the Nuclear Materials mission, this research will lead to improvements in the understanding of radiation effects on the chemical and structural properties of materials for the stabilization and long-term storage of plutonium, highly-enriched uranium, and other actinides. The research uses plutonium incorporation, ion-beam irradiation, and electron-beam irradiation to simulate the effects of alpha decay and beta decay on relevant glasses and ceramics. The research under this project has the potential to result in improved glass and ceramic materials for the stabilization and immobilization of high-level tank waste, plutonium residues and scraps, surplus weapons plutonium, highly-enriched uranium, other actinides, and other radioactive materials.

  16. Radiation Effects in Nuclear Waste Materials

    SciTech Connect (OSTI)

    Weber, William J.

    2005-06-01

    The objective of this project is to develop a fundamental understanding of radiation effects in glasses and ceramics, as well as the influence of solid-state radiation effects on aqueous dissolution kinetics, which may impact the performance of nuclear waste forms and stabilized nuclear materials. This work provides the underpinning science to develop improved glass and ceramic waste forms for the immobilization and disposition of high-level tank waste, excess plutonium, plutonium residues and scrap, other actinides, and other nuclear waste streams. Furthermore, this work is developing develop predictive models for the performance of nuclear waste forms and stabilized nuclear materials. Thus, the research performed under this project has significant implications for the immobilization of High-Level Waste (HLW) and Nuclear Materials, two mission areas within the Office of Environmental Management (EM). With regard to the HLW mission, this research will lead to improved understanding of radiation-induced degradation mechanisms and their effects on dissolution kinetics, as well as development of predictive models for waste form performance. In the Nuclear Materials mission, this research will lead to improvements in the understanding of radiation effects on the chemical and structural properties of materials for the stabilization and long-term storage of plutonium, highly-enriched uranium, and other actinides. The research uses plutonium incorporation, ion-beam irradiation, and electron-beam irradiation to simulate the effects of alpha decay and beta decay on relevant glasses and ceramics. The research under this project has the potential to result in improved glass and ceramic materials for the stabilization and immobilization of high-level tank waste, plutonium residues and scraps, surplus weapons plutonium, highly-enriched uranium, other actinides, and other radioactive materials.

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

    SciTech Connect (OSTI)

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

    1981-10-01

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

  18. Project plan for resolution of the organic waste tank safety issues at the Hanford Site

    SciTech Connect (OSTI)

    Meacham, J.E.

    1996-10-03

    A multi-year project plan for the Organic Safety Project has been developed with the objective of resolving the organic safety issues associated with the High Level Waste (HLW) in Hanford`s single-shell tanks (SSTS) and double-shell tanks (DSTs). The objective of the Organic Safety Project is to ensure safe interim storage until retrieval for pretreatment and disposal operations begins, and to resolve the organic safety issues by September 2001. Since the initial identification of organics as a tank waste safety issue, progress has been made in understanding the specific aspects of organic waste combustibility, and in developing and implementing activities to resolve the organic safety issues.

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

    SciTech Connect (OSTI)

    Pavel Hrma; Pert Schill; Lubomir Nemec

    2002-01-07

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

  20. Development of low-level radioactive waste disposal capacity in the United States - progress or stalemate?

    SciTech Connect (OSTI)

    Devgun, J.S. [Argonne National Lab., IL (United States); Larson, G.S. [Midwest Low-Level Radioactive Waste Commission, St. Paul, MN (United States)

    1995-12-31

    It has been fifteen years since responsibility for the disposal of commercially generated low-level radioactive waste (LLW) was shifted to the states by the United States Congress through the Low-Level Radioactive Waste Policy Act of 1980 (LLRWPA). In December 1985, Congress revisited the issue and enacted the Low-Level Radioactive Waste Policy Amendments Act of 1985 (LLRWPAA). No new disposal sites have opened yet, however, and it is now evident that disposal facility development is more complex, time-consuming, and controversial than originally anticipated. For a nation with a large nuclear power industry, the lack of availability of LLW disposal capacity coupled with a similar lack of high-level radioactive waste disposal capacity could adversely affect the future viability of the nuclear energy option. The U.S. nuclear power industry, with 109 operating reactors, generates about half of the LLW shipped to commercial disposal sites and faces dwindling access to waste disposal sites and escalating waste management costs. The other producers of LLW - industries, government (except the defense related research and production waste), academic institutions, and medical institutions that account for the remaining half of the commercial LLW - face the same storage and cost uncertainties. This paper will summarize the current status of U.S. low-level radioactive waste generation and the status of new disposal facility development efforts by the states. The paper will also examine the factors that have contributed to delays, the most frequently suggested alternatives, and the likelihood of change.

  1. Final Report - Effects of High Spinel and Chromium Oxide Crystal Contents on Simulated HLW Vitrification in DM100 Melter Tests, VSL-09R1520-1, Rev. 0, dated 6/22/09

    SciTech Connect (OSTI)

    Kruger, Albert A.; Matlack, K. S.; Kot, W.; Pegg, I. L.; Chaudhuri, M.; Lutze, W.

    2013-11-13

    The principal objective of the work was to evaluate the effects of spinel and chromium oxide particles on WTP HLW melter operations and potential impacts on melter life. This was accomplished through a combination of crucible-scale tests, settling and rheological tests, and tests on the DM100 melter system. Crucible testing was designed to develop and identify HLW glass compositions with high waste loadings that exhibit formation of crystalline spinel and/or chromium oxide phases up to relatively high crystal contents (i.e., > 1 vol%). Characterization of crystal settling and the effects on melt rheology was performed on the HLW glass formulations. Appropriate candidate HLW glass formulations were selected, based on characterization results, to support subsequent melter tests. In the present work, crucible melts were formulated that exhibit up to about 4.4 vol% crystallization.

  2. FINAL REPORT START-UP AND COMMISSIONING TESTS ON THE DURAMELTER 1200 HLW PILOT MELTER SYSTEM USING AZ-101 HLW SIMULANTS VSL-01R0100-2 REV 0 1/20/03

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; KOT WK; BRANDYS M; WILSON CN; SCHATZ TR; GONG W; PEGG IL

    2011-12-29

    This document provides the final report on data and results obtained from commissioning tests performed on the one-third scale DuraMelter{trademark} 1200 (DM 1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part BI [1]. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plan. This report is a followup to the previously issued Preliminary Data Summary Report. The DM1200 system will be used for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. This will include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The results presented in this report are from the initial series of short-duration tests that were conducted to support the start-up and commissioning of this system prior to conducting the main body of development tests that have been planned for this system. These tests were directed primarily at system 'debugging,' operator training, and procedure refinement. The AZ-101 waste simulant and glass composition that was used for previous testing was selected for these tests.

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

    SciTech Connect (OSTI)

    Ramsey, William Gene

    2013-08-15

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

  4. DOE to Weigh Alternatives for Greater Than Class C Low-Level Waste Disposal

    Energy Savers [EERE]

    | Department of Energy to Weigh Alternatives for Greater Than Class C Low-Level Waste Disposal DOE to Weigh Alternatives for Greater Than Class C Low-Level Waste Disposal July 20, 2007 - 2:55pm Addthis WASHINGTON, DC - The U.S. Department of Energy (DOE) today announced that it will evaluate disposal options for Greater Than Class C (GTCC) low-level radioactive waste (LLW) generated from the decommissioning of nuclear power plants, medical activities and nuclear research. DOE delivered to

  5. DOE - NNSA/NFO -- EM Low-Level Waste Grant Assistance Program

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

    Grant Assistance Program NNSA/NFO Language Options U.S. DOE/NNSA - Nevada Field Office Click to subscribe to NNSS News Low-Level Waste Grant Assistance Program An Emergency Management grant program was instituted in fiscal year 2000 to provide funding for enhancing county emergency response capabilities in communities near the Nevada National Security Site. To fund this grant, approved waste generators are charged an additional $.50 per cubic foot of low-level and mixed-level waste disposed. The

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

    SciTech Connect (OSTI)

    Not Available

    1983-06-01

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

  7. High Level Waste Corporate Board Newsletter - 06/03/09

    Office of Environmental Management (EM)

    UPCOMING EVENTS: Tank Waste Corporate Board Oak Ridge National Laboratory Oak Ridge, Tennessee 28 - 29 July 2009 The Board meeting will be preceded by a tour of the Radiochemical Engineering and Development Center on the afternoon of Tuesday, 28 July, and the meeting is planned for a full day on Wednesday, 29 July. Agenda Items include: ï‚· Future Directions for DOE Office of Nuclear Energy ï‚· Robotic Arm for Tank Cleaning ï‚· AREVA Mobile Hot Cell ï‚· Integrated Project Team Report ï‚· DOE

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

    SciTech Connect (OSTI)

    R.A. Levich; J.S. Stuckless

    2006-09-25

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

  9. Secondary Waste Form Down-Selection Data Package—Fluidized Bed Steam Reforming Waste Form

    SciTech Connect (OSTI)

    Qafoku, Nikolla; Westsik, Joseph H.; Strachan, Denis M.; Valenta, Michelle M.; Pires, Richard P.

    2011-09-12

    The Hanford Site in southeast Washington State has 56 million gallons of radioactive and chemically hazardous wastes stored in 177 underground tanks (ORP 2010). The U.S. Department of Energy (DOE), Office of River Protection (ORP), through its contractors, is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to convert the radioactive and hazardous wastes into stable glass waste forms for disposal. Within the WTP, the pretreatment facility will receive the retrieved waste from the tank farms and separate it into two treated process streams. These waste streams will be vitrified, and the resulting waste canisters will be sent to offsite (high-level waste [HLW]) and onsite (immobilized low-activity waste [ILAW]) repositories. As part of the pretreatment and ILAW processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility (ETF) on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed of in the Integrated Disposal Facility (IDF). To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is developing data packages to support that down-selection. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilizing and solidifying the liquid secondary wastes. At the Hanford Site, the FBSR process is being evaluated as a supplemental technology for treating and immobilizing Hanford LAW radioactive tank waste and for treating secondary wastes from the WTP pretreatment and LAW vitrification processes.

  10. 3-D MAPPING TECHNOLOGIES FOR HIGH LEVEL WASTE TANKS

    SciTech Connect (OSTI)

    Marzolf, A.; Folsom, M.

    2010-08-31

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

  11. Stabilization of liquid low-level and mixed wastes: a treatability study

    SciTech Connect (OSTI)

    Carson, S.; Cheng, Yu-Cheng; Yellowhorse, L.; Peterson, P.

    1996-02-01

    A treatability study has been conducted on liquid low-level and mixed wastes using the stabilization agents Aquaset, Aquaset II, Aquaset II-H, Petroset, Petroset-H, and Petroset and Petroset II. A total of 40 different waste types with activities ranging from 10{sup {minus}14} to 10{sup {minus}4} curies/ml have been stabilized. Reported data for each waste include its chemical and radiological composition and the optimum composition or range of compositions (weight of agent/volume of waste) for each stabilization agent used. All wastes were successfully stabilized with one or more of the stabilization agents and all final waste forms passed the Paint Filter Liquids Test (EPA Method 9095).

  12. Methods for verifying compliance with low-level radioactive waste acceptance criteria

    SciTech Connect (OSTI)

    NONE

    1993-09-01

    This report summarizes the methods that are currently employed and those that can be used to verify compliance with low-level radioactive waste (LLW) disposal facility waste acceptance criteria (WAC). This report presents the applicable regulations representing the Federal, State, and site-specific criteria for accepting LLW. Typical LLW generators are summarized, along with descriptions of their waste streams and final waste forms. General procedures and methods used by the LLW generators to verify compliance with the disposal facility WAC are presented. The report was written to provide an understanding of how a regulator could verify compliance with a LLW disposal facility`s WAC. A comprehensive study of the methodology used to verify waste generator compliance with the disposal facility WAC is presented in this report. The study involved compiling the relevant regulations to define the WAC, reviewing regulatory agency inspection programs, and summarizing waste verification technology and equipment. The results of the study indicate that waste generators conduct verification programs that include packaging, classification, characterization, and stabilization elements. The current LLW disposal facilities perform waste verification steps on incoming shipments. A model inspection and verification program, which includes an emphasis on the generator`s waste application documentation of their waste verification program, is recommended. The disposal facility verification procedures primarily involve the use of portable radiological survey instrumentation. The actual verification of generator compliance to the LLW disposal facility WAC is performed through a combination of incoming shipment checks and generator site audits.

  13. Closed Fuel Cycle Waste Treatment Strategy

    SciTech Connect (OSTI)

    Vienna, J. D.; Collins, E. D.; Crum, J. V.; Ebert, W. L.; Frank, S. M.; Garn, T. G.; Gombert, D.; Jones, R.; Jubin, R. T.; Maio, V. C.; Marra, J. C.; Matyas, J.; Nenoff, T. M.; Riley, B. J.; Sevigny, G. J.; Soelberg, N. R.; Strachan, D. M.; Thallapally, P. K.; Westsik, J. H.

    2015-02-01

    This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

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

    SciTech Connect (OSTI)

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

    2011-12-29

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

  15. Extended storage of low-level radioactive waste: potential problem areas

    SciTech Connect (OSTI)

    Siskind, B.; Dougherty, D.R.; MacKenzie, D.R.

    1985-01-01

    If a state or state compact does not have adequate disposal capacity for low-level radioactive waste (LLRW) by 1986 as required by the Low-Level Waste Policy Act, then extended storage of certain LLRW may be necessary. The issue of extended storage of LLRW is addressed in order to determine for the Nuclear Regulatory Commission the areas of concern and the actions recommended to resolve these concerns. The focus is on the properties and behavior of the waste form and waste container. Storage alternatives are considered in order to characterize the likely storage environments for these wastes. The areas of concern about extended storage of LLRW are grouped into two categories: 1. Behavior of the waste form and/or container during storage, e.g., radiolytic gas generation, radiation-enhanced degradation of polymeric materials, and corrosion. 2. Effects of extended storage on the properties of the waste form and/or container that are important after storage (e.g., radiation-induced oxidative embrittlement of high-density polyethylene and the weakening of steel containers resulting from corrosion by the waste). The additional information and actions required to address these concerns are discussed and, in particular, it is concluded that further information is needed on the rates of corrosion of container material by Class A wastes and on the apparent dose-rate dependence of radiolytic processes in Class B and C waste packages. Modifications to the guidance for solidified wastes and high-integrity containers in NRC's Technical Position on Waste Form are recommended. 27 references.

  16. 1994 annual report on low-level radioactive waste management progress

    SciTech Connect (OSTI)

    1995-04-01

    This report for calendar year 1994 summarizes the progress that states and compact regions made during the year in establishing new low-level radioactive waste disposal facilities. Although events that have occurred in 1995 greatly alter the perspective in terms of storage versus disposal, the purpose of this report is to convey the concerns as evidenced during calendar year 1994. Significant developments occurring in 1995 are briefly outlined in the transmittal letter and will be detailed in the report for calendar year 1995. The report also provides summary information on the volume of low-level radioactive waste received for disposal in 1994 by commercially operated low-level radioactive waste disposal facilities, and is prepared is in response to Section 7(b) of Title I of Public Law 99-240, the Low-Level Radioactive Waste Policy Amendments Act of 1985.

  17. EA-1135: Offsite Thermal Treatment of Low-level Mixed Waste, Richland, Washington

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of the proposal to treat contact-handled low-level mixed waste, containing polychlorinated biphenyls and other organics, to meet existing regulatory...

  18. EA-1189: Non-thermal Treatment of Hanford Site Low-level Mixed Waste, Richland, Washington

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts for the proposal to demonstrate the feasibility of commercial treatment of contact-handled low-level mixed waste to meet existing Federal and State...

  19. EA-1292: On-site Treatment of Low Level Mixed Waste, Golden, Colorado

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts for the proposal to evaluate the proposed treatment of low level mixed waste at the U.S. Department of Energy's Rocky Flats Environmental Technology Site.

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

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

    Introduction to DOE Order 435.1 Low Level Radioactive Waste Disposal Requirements Christine Gelles*, U.S. Department of Energy ; Edward Regnier, U.S. Department of Energy; Andrew ...

  1. EA-0843: Idaho National Engineering Laboratory Low-Level and Mixed Waste Processing, Idaho Falls, Idaho

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of a proposal to (1) reduce the volume of the U.S. Department of Energy's Idaho National Engineering Laboratory's (INEL) generated low-level waste (LLW)...

  2. Low-level radioactive waste disposal technologies used outside the United States

    SciTech Connect (OSTI)

    Templeton, K.J.; Mitchell, S.J.; Molton, P.M.; Leigh, I.W.

    1994-01-01

    Low-level radioactive waste (LLW) disposal technologies are an integral part of the waste management process. In the United States, commercial LLW disposal is the responsibility of the State or groups of States (compact regions). The United States defines LLW as all radioactive waste that is not classified as spent nuclear fuel, high- level radioactive waste, transuranic waste, or by-product material as defined in Section II(e)(2) of the Atomic Energy Act. LLW may contain some long-lived components in very low concentrations. Countries outside the United States, however, may define LLW differently and may use different disposal technologies. This paper outlines the LLW disposal technologies that are planned or being used in Canada, China, Finland, France, Germany, Japan, Sweden, Taiwan, and the United Kingdom (UK).

  3. HIGH LEVEL WASTE MECHANCIAL SLUDGE REMOVAL AT THE SAVANNAH RIVER SITE F TANK FARM CLOSURE PROJECT

    SciTech Connect (OSTI)

    Jolly, R; Bruce Martin, B

    2008-01-15

    The Savannah River Site F-Tank Farm Closure project has successfully performed Mechanical Sludge Removal (MSR) using the Waste on Wheels (WOW) system for the first time within one of its storage tanks. The WOW system is designed to be relatively mobile with the ability for many components to be redeployed to multiple waste tanks. It is primarily comprised of Submersible Mixer Pumps (SMPs), Submersible Transfer Pumps (STPs), and a mobile control room with a control panel and variable speed drives. In addition, the project is currently preparing another waste tank for MSR utilizing lessons learned from this previous operational activity. These tanks, designated as Tank 6 and Tank 5 respectively, are Type I waste tanks located in F-Tank Farm (FTF) with a capacity of 2,840 cubic meters (750,000 gallons) each. The construction of these tanks was completed in 1953, and they were placed into waste storage service in 1959. The tank's primary shell is 23 meters (75 feet) in diameter, and 7.5 meters (24.5 feet) in height. Type I tanks have 34 vertically oriented cooling coils and two horizontal cooling coil circuits along the tank floor. Both Tank 5 and Tank 6 received and stored F-PUREX waste during their operating service time before sludge removal was performed. DOE intends to remove from service and operationally close (fill with grout) Tank 5 and Tank 6 and other HLW tanks that do not meet current containment standards. Mechanical Sludge Removal, the first step in the tank closure process, will be followed by chemical cleaning. After obtaining regulatory approval, the tanks will be isolated and filled with grout for long-term stabilization. Mechanical Sludge Removal operations within Tank 6 removed approximately 75% of the original 95,000 liters (25,000 gallons). This sludge material was transferred in batches to an interim storage tank to prepare for vitrification. This operation consisted of eleven (11) Submersible Mixer Pump(s) mixing campaigns and multiple intraarea transfers utilizing STPs from July 2006 to August 2007. This operation and successful removal of sludge material meets requirement of approximately 19,000 to 28,000 liters (5,000 to 7,500 gallons) remaining prior to the Chemical Cleaning process. Removal of the last 35% of sludge was exponentially more difficult, as less and less sludge was available to mobilize and the lighter sludge particles were likely removed during the early mixing campaigns. The removal of the 72,000 liters (19,000 gallons) of sludge was challenging due to a number factors. One primary factor was the complex internal cooling coil array within Tank 6 that obstructed mixer discharge jets and impacted the Effective Cleaning Radius (ECR) of the Submersible Mixer Pumps. Minimal access locations into the tank through tank openings (risers) presented a challenge because the available options for equipment locations were very limited. Mechanical Sludge Removal activities using SMPs caused the sludge to migrate to areas of the tank that were outside of the SMP ECR. Various SMP operational strategies were used to address the challenge of moving sludge from remote areas of the tank to the transfer pump. This paper describes in detail the Mechanical Sludge Removal activities and mitigative solutions to cooling coil obstructions and other challenges. The performance of the WOW system and SMP operational strategies were evaluated and the resulting lessons learned are described for application to future Mechanical Sludge Removal operations.

  4. 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 Nuclear Fuel October 2014 [This page left blank.] FOREWORD In January 2012, the final report of the Blue Ribbon Commission on America's Nuclear Future urged the Administration to conduct a review of the current policy to dispose of defense and commercial high level radioactive waste and spent nuclear fuel in a single repository or repositories. Based on that recommendation, in January 2013, the Administration's

  5. Technical area status report for low-level mixed waste final waste forms. Volume 1

    SciTech Connect (OSTI)

    Mayberry, J.L.; DeWitt, L.M.; Darnell, R.

    1993-08-01

    The Final Waste Forms (FWF) Technical Area Status Report (TASR) Working Group, the Vitrification Working Group (WG), and the Performance Standards Working Group were established as subgroups to the FWF Technical Support Group (TSG). The FWF TASR WG is comprised of technical representatives from most of the major DOE sites, the Nuclear Regulatory Commission (NRC), the EPA Office of Solid Waste, and the EPA`s Risk Reduction Engineering Laboratory (RREL). The primary activity of the FWF TASR Working Group was to investigate and report on the current status of FWFs for LLNM in this TASR. The FWF TASR Working Group determined the current status of the development of various waste forms described above by reviewing selected articles and technical reports, summarizing data, and establishing an initial set of FWF characteristics to be used in evaluating candidate FWFS; these characteristics are summarized in Section 2. After an initial review of available information, the FWF TASR Working Group chose to study the following groups of final waste forms: hydraulic cement, sulfur polymer cement, glass, ceramic, and organic binders. The organic binders included polyethylene, bitumen, vinyl ester styrene, epoxy, and urea formaldehyde. Section 3 provides a description of each final waste form. Based on the literature review, the gaps and deficiencies in information were summarized, and conclusions and recommendations were established. The information and data presented in this TASR are intended to assist the FWF Production and Assessment TSG in evaluating the Technical Task Plans (TTPs) submitted to DOE EM-50, and thus provide DOE with the necessary information for their FWF decision-making process. This FWF TASR will also assist the DOE and the MWIP in establishing the most acceptable final waste forms for the various LLMW streams stored at DOE facilities.

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

    Office of Scientific and Technical Information (OSTI)

    (Technical Report) | SciTech Connect Road Map for Development of Crystal-Tolerant High Level Waste Glasses Citation Details In-Document Search Title: Road Map for Development of Crystal-Tolerant High Level Waste Glasses This road map guides the research and development for formulation and processing of crystal-tolerant glasses, identifying near- and long-term activities that need to be completed over the period from 2014 to 2019. The primary objective is to maximize waste loading for Hanford

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

    Energy Savers [EERE]

    Requirements | Department of Energy 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 Gelles*, U.S. Department of Energy ; Edward Regnier, U.S. Department of Energy; Andrew Wallo, U.S. Department of Energy Abstract: The Atomic Energy Act gives the U.S. Department of Energy (US DOE), the authority to regulate the management of radioactive waste generated by US DOE. This

  8. Transuranic and Low-Level Boxed Waste Form Nondestructive Assay Technology Overview and Assessment

    SciTech Connect (OSTI)

    G. Becker; M. Connolly; M. McIlwain

    1999-02-01

    The Mixed Waste Focus Area (MWFA) identified the need to perform an assessment of the functionality and performance of existing nondestructive assay (NDA) techniques relative to the low-level and transuranic waste inventory packaged in large-volume box-type containers. The primary objectives of this assessment were to: (1) determine the capability of existing boxed waste form NDA technology to comply with applicable waste radiological characterization requirements, (2) determine deficiencies associated with existing boxed waste assay technology implementation strategies, and (3) recommend a path forward for future technology development activities, if required. Based on this assessment, it is recommended that a boxed waste NDA development and demonstration project that expands the existing boxed waste NDA capability to accommodate the indicated deficiency set be implemented. To ensure that technology will be commercially available in a timely fashion, it is recommended this development and demonstration project be directed to the private sector. It is further recommended that the box NDA technology be of an innovative design incorporating sufficient NDA modalities, e.g., passive neutron, gamma, etc., to address the majority of the boxed waste inventory. The overall design should be modular such that subsets of the overall NDA system can be combined in optimal configurations tailored to differing waste types.

  9. National profile on commercially generated low-level radioactive mixed waste

    SciTech Connect (OSTI)

    Klein, J.A.; Mrochek, J.E.; Jolley, R.L.; Osborne-Lee, I.W.; Francis, A.A.; Wright, T.

    1992-12-01

    This report details the findings and conclusions drawn from a survey undertaken as part of a joint US Nuclear Regulatory Commission and US Environmental Protection Agency-sponsored project entitled ``National Profile on Commercially Generated Low-Level Radioactive Mixed Waste.`` The overall objective of the work was to compile a national profile on the volumes, characteristics, and treatability of commercially generated low-level mixed waste for 1990 by five major facility categories-academic, industrial, medical, and NRC-/Agreement State-licensed goverment facilities and nuclear utilities. Included in this report are descriptions of the methodology used to collect and collate the data, the procedures used to estimate the mixed waste generation rate for commercial facilities in the United States in 1990, and the identification of available treatment technologies to meet applicable EPA treatment standards (40 CFR Part 268) and, if possible, to render the hazardous component of specific mixed waste streams nonhazardous. The report also contains information on existing and potential commercial waste treatment facilities that may provide treatment for specific waste streams identified in the national survey. The report does not include any aspect of the Department of Energy`s (DOES) management of mixed waste and generally does not address wastes from remedial action activities.

  10. Tank Waste Remediation System tank waste pretreatment and vitrification process development testing requirements assessment

    SciTech Connect (OSTI)

    Howden, G.F.

    1994-10-24

    A multi-faceted study was initiated in November 1993 to provide assurance that needed testing capabilities, facilities, and support infrastructure (sampling systems, casks, transportation systems, permits, etc.) would be available when needed for process and equipment development to support pretreatment and vitrification facility design and construction schedules. This first major report provides a snapshot of the known testing needs for pretreatment, low-level waste (LLW) and high-level waste (HLW) vitrification, and documents the results of a series of preliminary studies and workshops to define the issues needing resolution by cold or hot testing. Identified in this report are more than 140 Hanford Site tank waste pretreatment and LLW/HLW vitrification technology issues that can only be resolved by testing. The report also broadly characterizes the level of testing needed to resolve each issue. A second report will provide a strategy(ies) for ensuring timely test capability. Later reports will assess the capabilities of existing facilities to support needed testing and will recommend siting of the tests together with needed facility and infrastructure upgrades or additions.

  11. Final Tank Closure and Waste Management Environmental Impact Statement for the Hanford Site, Richland, Washington--Frequent Asked Questions

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

    Final Tank Closure and Waste Management Environmental Impact Statement for the Hanford Site, Richland, Washington (Final TC & WM EIS) (DOE/EIS-0391) Frequently Asked Questions What are the U.S. Department of Energy (DOE) proposed actions in the Final TC & WM EIS? The Final TC & WM EIS (DOE/EIS-0391) evaluates three sets of proposed actions, as follows: Retrieve and treat the waste remaining in 177 underground storage tanks; store the high-level radioactive waste (HLW); dispose of

  12. Proceedings of the Third Annual Information Meeting DOE Low-Level Waste-Management Program

    SciTech Connect (OSTI)

    Large, D.E.; Lowrie, R.S.; Stratton, L.E.; Jacobs, D.G.

    1981-12-01

    The Third Annual Participants Information Meeting of the Low-Level Waste Management Program was held in New Orleans, Louisiana, November 4-6, 1981 The specific purpose was to bring together appropriate representatives of industry, USNRC, program management, participating field offices, and contractors to: (1) exchange information and analyze program needs, and (2) involve participants in planning, developing and implementing technology for low-level waste management. One hundred seven registrants participated in the meeting. Presentation and workshop findings are included in these proceedings under the following headings: low-level waste activities; waste treatment; shallow land burial; remedial action; greater confinement; ORNL reports; panel workshops; and summary. Forty-six papers have been abstracted and indexed for the data base.

  13. Department of Energy treatment capabilities for greater-than-Class C low-level radioactive waste

    SciTech Connect (OSTI)

    Morrell, D.K.; Fischer, D.K.

    1995-01-01

    This report provides brief profiles for 26 low-level and high-level waste treatment capabilities available at the Idaho National Engineering Laboratory (INEL), Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), Pacific Northwest Laboratory (PNL), Rocky Flats Plant (RFP), Savannah River Site (SRS), and West Valley Demonstration Plant (WVDP). Six of the treatments have potential use for greater-than-Class C low-level waste (GTCC LLW). They include: (a) the glass ceramic process and (b) the Waste Experimental Reduction Facility incinerator at INEL; (c) the Super Compaction and Repackaging Facility and (d) microwave melting solidification at RFP; (e) the vitrification plant at SRS; and (f) the vitrification plant at WVDP. No individual treatment has the capability to treat all GTCC LLW streams. It is recommended that complete physical and chemical characterizations be performed for each GTCC waste stream, to permit using multiple treatments for GTCC LLW.

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

    DOE Patents [OSTI]

    Boatner, Lynn A.; Sales, Brian C.

    1989-01-01

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

  15. Cost savings associated with landfilling wastes containing very low levels of uranium

    SciTech Connect (OSTI)

    Boggs, C.J.; Shaddoan, W.T.

    1996-03-01

    The Paducah Gaseous Diffusion Plant (PGDP) has operated captive landfills (both residential and construction/demolition debris) in accordance with the Commonwealth of Kentucky regulations since the early 1980s. Typical waste streams allowed in these landfills include nonhazardous industrial and municipal solid waste (such as paper, plastic, cardboard, cafeteria waste, clothing, wood, asbestos, fly ash, metals, and construction debris). In July 1992, the U.S. Environmental Protection Agency issued new requirements for the disposal of sanitary wastes in a {open_quotes}contained landfill.{close_quotes} These requirements were promulgated in the 401 Kentucky Administrative Record Chapters 47 and 48 that became effective 30 June 1995. The requirements for a new contained landfill include a synthetic liner made of high-density polyethylene in addition to the traditional 1-meter (3-foot) clay liner and a leachate collection system. A new landfill at Paducah would accept waste streams similar to those that have been accepted in the past. The permit for the previously existing landfills did not include radioactivity limits; instead, these levels were administratively controlled. Typically, if radioactivity was detected above background levels, the waste was classified as low-level waste (LLW), which would be sent off-site for disposal.

  16. Pretreatment of neutralized cladding removal waste sludge: Status Report

    SciTech Connect (OSTI)

    Lumetta, G J; Swanson, J L

    1993-03-01

    This report describes the status of process development for pretreating Hanford neutralized cladding removal waste (NCRW) sludge, of which [approximately] 3.3 [times] 10[sup 6] L is stored in Tanks 103-AW and 105-AW at the Hanford Site. The initial baseline process chosen for pretreating NCRW sludge is to dissolve the sludge in nitric acid and extract the -transuranic (MU) elements from the dissolved sludge solution with octyl(phenyl)-N,N-diisobutylcarbamoyl methyl phosphine oxide (CNWO). This process converts the NCRW sludge into a relatively large volume of low-level waste (LLW) to be disposed of as grout, leaving only a small volume of high-level waste (HLW) requiring vitrification in the Hanford Waste Vitrification Plant (HWVP).

  17. Separation of americium, curium, and rare earths from high-level wastes by oxalate precipitation: experiments with synthetic waste solutions

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1980-01-01

    The separation of trivalent actinides and rare earths from other fission products in high-level nuclear wastes by oxalate precipitation followed by ion exchange (OPIX) was experimentally investigated using synthetic wastes and a small-scale, continuous-flow oxalic acid precipitation and solid-liquid separation system. Trivalent actinide and rare earth oxalates are relatively insoluble in 0.5 to 1.0 M HNO/sub 3/ whereas other fission product oxalates are not. The continuous-flow system consisted of one or two stirred-tank reactors in series for crystal growth. Oxalic acid and waste solutions were mixed in the first tank, with the product solid-liquid slurry leaving the second tank. Solid-liquid separation was tested by filters and by a gravity settler. The experiments determined the fraction of rare earths precipitated and separated from synthetic waste streams as a function of number of reactors, system temperature, oxalic acid concentration, liquid residence time in the process, power input to the stirred-tank reactors, and method of solid-liquid separation. The crystalline precipitate was characterized with respect to form, size, and chemical composition. These experiments are only the first step in converting a proposed chemical flowsheet into a process flowsheet suitable for large-scale remote operations at high activity levels.

  18. Silicon-Polymer Encapsulation of High-Level Calcine Waste for Transportation or Disposal

    SciTech Connect (OSTI)

    G. G. Loomis; C. M. Miller; J. A. Giansiracusa; R. Kimmel; S. V. Prewett

    2000-01-01

    This report presents the results of an experimental study investigating the potential uses for silicon-polymer encapsulation of High Level Calcine Waste currently stored within the Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Engineering and Environmental Laboratory (INEEL). The study investigated two different applications of silicon polymer encapsulation. One application uses silicon polymer to produce a waste form suitable for disposal at a High Level Radioactive Waste Disposal Facility directly, and the other application encapsulates the calcine material for transportation to an offsite melter for further processing. A simulated waste material from INTEC, called pilot scale calcine, which contained hazardous materials but no radioactive isotopes was used for the study, which was performed at the University of Akron under special arrangement with Orbit Technologies, the originators of the silicon polymer process called Polymer Encapsulation Technology (PET). This document first discusses the PET process, followed by a presentation of past studies involving PET applications to waste problems. Next, the results of an experimental study are presented on encapsulation of the INTEC calcine waste as it applies to transportation or disposal of calcine waste. Results relating to long-term disposal include: (1) a characterization of the pilot calcine waste; (2) Toxicity Characteristic Leaching Procedure (TCLP) testing of an optimum mixture of pilot calcine, polysiloxane and special additives; and, (3) Material Characterization Center testing MCC-1P evaluation of the optimum waste form. Results relating to transportation of the calcine material for a mixture of maximum waste loading include: compressive strength testing, 10-m drop test, melt testing, and a Department of Transportation (DOT) oxidizer test.

  19. Polysiloxane Encapsulation of High Level Calcine Waste for Transportation or Disposal

    SciTech Connect (OSTI)

    Loomis, Guy George

    2000-03-01

    This report presents the results of an experimental study investigating the potential uses for silicon-polymer encapsulation of High Level Calcine Waste currently stored within the Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Engineering and Environmental Laboratory (INEEL). The study investigated two different applications of silicon polymer encapsulation. One application uses silicon polymer to produce a waste form suitable for disposal at a High Level Radioactive Waste Disposal Facility directly, and the other application encapsulates the calcine material for transportation to an offsite melter for further processing. A simulated waste material from INTEC, called pilot scale calcine, which contained hazardous materials but no radioactive isotopes was used for the study, which was performed at the University of Akron under special arrangement with Orbit Technologies, the originators of the silicon polymer process called Polymer Encapsulation Technology (PET). This document first discusses the PET process, followed by a presentation of past studies involving PET applications to waste problems. Next, the results of an experimental study are presented on encapsulation of the INTEC calcine waste as it applies to transportation or disposal of calcine waste. Results relating to long-term disposal include: 1) a characterization of the pilot calcine waste; 2) Toxicity Characteristic Leaching Procedure (TCLP) testing of an optimum mixture of pilot calcine, polysiloxane and special additives; and, 3) Material Characterization Center testing MCC-1P evaluation of the optimum waste form. Results relating to transportation of the calcine material for a mixture of maximum waste loading include: compressive strength testing, 10-m drop test, melt testing, and a Department of Transportation (DOT) oxidizer test.

  20. SOLIDIFICATION OF THE HANFORD LAW WASTE STREAM PRODUCED AS A RESULT OF NEAR-TANK CONTINUOUS SLUDGE LEACHING AND SODIUM HYDROXIDE RECOVERY

    SciTech Connect (OSTI)

    Reigel, M.; Johnson, F.; Crawford, C.; Jantzen, C.

    2011-09-20

    The U.S. Department of Energy (DOE), Office of River Protection (ORP), is responsible for the remediation and stabilization of the Hanford Site tank farms, including 53 million gallons of highly radioactive mixed wasted waste contained in 177 underground tanks. The plan calls for all waste retrieved from the tanks to be transferred to the Waste Treatment Plant (WTP). The WTP will consist of three primary facilities including pretreatment facilities for Low Activity Waste (LAW) to remove aluminum, chromium and other solids and radioisotopes that are undesirable in the High Level Waste (HLW) stream. Removal of aluminum from HLW sludge can be accomplished through continuous sludge leaching of the aluminum from the HLW sludge as sodium aluminate; however, this process will introduce a significant amount of sodium hydroxide into the waste stream and consequently will increase the volume of waste to be dispositioned. A sodium recovery process is needed to remove the sodium hydroxide and recycle it back to the aluminum dissolution process. The resulting LAW waste stream has a high concentration of aluminum and sodium and will require alternative immobilization methods. Five waste forms were evaluated for immobilization of LAW at Hanford after the sodium recovery process. The waste forms considered for these two waste streams include low temperature processes (Saltstone/Cast stone and geopolymers), intermediate temperature processes (steam reforming and phosphate glasses) and high temperature processes (vitrification). These immobilization methods and the waste forms produced were evaluated for (1) compliance with the Performance Assessment (PA) requirements for disposal at the IDF, (2) waste form volume (waste loading), and (3) compatibility with the tank farms and systems. The iron phosphate glasses tested using the product consistency test had normalized release rates lower than the waste form requirements although the CCC glasses had higher release rates than the quenched glasses. However, the waste form failed to meet the vapor hydration test criteria listed in the WTP contract. In addition, the waste loading in the phosphate glasses were not as high as other candidate waste forms. Vitrification of HLW waste as borosilicate glass is a proven process; however the HLW and LAW streams at Hanford can vary significantly from waste currently being immobilized. The ccc glasses show lower release rates for B and Na than the quenched glasses and all glasses meet the acceptance criterion of < 4 g/L. Glass samples spiked with Re{sub 2}O{sub 7} also passed the PCT test. However, further vapor hydration testing must be performed since all the samples cracked and the test could not be performed. The waste loading of the iron phosphate and borosilicate glasses are approximately 20 and 25% respectively. The steam reforming process produced the predicted waste form for both the high and low aluminate waste streams. The predicted waste loadings for the monolithic samples is approximately 39%, which is higher than the glass waste forms; however, at the time of this report, no monolithic samples were made and therefore compliance with the PA cannot be determined. The waste loading in the geopolymer is approximately 40% but can vary with the sodium hydroxide content in the waste stream. Initial geopolymer mixes revealed compressive strengths that are greater than 500 psi for the low aluminate mixes and less than 500 psi for the high aluminate mixes. Further work testing needs to be performed to formulate a geopolymer waste form made using a high aluminate salt solution. A cementitious waste form has the advantage that the process is performed at ambient conditions and is a proven process currently in use for LAW disposal. The Saltstone/Cast Stone formulated using low and high aluminate salt solutions retained at least 97% of the Re that was added to the mix as a dopant. While this data is promising, additional leaching testing must be performed to show compliance with the PA. Compressive strength tests must also be performed on the Cast Ston

  1. Estimate of the Potential Amount of Low-Level Waste from the Fukushima Prefecture - 12370

    SciTech Connect (OSTI)

    Hill, Carolyn; Olson, Eric A.J.; Elmer, John

    2012-07-01

    The amount of waste generated by the cleanup of the Fukushima Prefecture (Fukushima-ken) following the releases from the Fukushima Daiichi nuclear power plant accident (March 2011) is dependent on many factors, including: - Contamination amounts; - Cleanup levels determined for the radioisotopes contaminating the area; - Future land use expectations and human exposure scenarios; - Groundwater contamination considerations; - Costs and availability of storage areas, and eventually disposal areas for the waste; and - Decontamination and volume reduction techniques and technologies used. For the purposes of estimating these waste volumes, Fukushima-ken is segregated into zones of similar contamination level and expected future use. Techniques for selecting the appropriate cleanup methods for each area are shown in a decision tree format. This approach is broadly applied to the 20 km evacuation zone and the total amounts and types of waste are estimated; waste resulting from cleanup efforts outside of the evacuation zone is not considered. Some of the limits of future use and potential zones where residents must be excluded within the prefecture are also described. The size and design of the proposed intermediate storage facility is also discussed and the current situation, cleanup, waste handling, and waste storage issues in Japan are described. The method for estimating waste amounts outlined above illustrates the large amount of waste that could potentially be generated by remediation of the 20 km evacuation zone (619 km{sup 2} total) if the currently proposed cleanup goals are uniformly applied. The Japanese environment ministry estimated in early October that the 1 mSv/year exposure goal would make the government responsible for decontaminating about 8,000 km{sup 2} within Fukushima-ken and roughly 4,900 km{sup 2} in areas outside the prefecture. The described waste volume estimation method also does not give any consideration to areas with localized hot spots. Land use and area dose rate estimates for the 20 km evacuation zone indicate there are large areas where doses to the public can be mitigated through methods other than removal and disposal of soil and other wastes. Several additional options for waste reduction can also be considered, including: - Recycling/reusing or disposing of as municipal waste material that can be unconditionally cleared; - Establishing additional precautionary (e.g., liners) and monitoring requirements for municipal landfills to dispose of some conditionally-cleared material; and - Using slightly-contaminated material in construction of reclamations, banks and roads. Waste estimates for cleanup will continue to evolve as decontamination plans are drafted and finalized. (authors)

  2. Low-level radioactive waste management at the Nevada Test Site -- Current status

    SciTech Connect (OSTI)

    Becker, B.D.; Crowe, B.M.; Gertz, C.P.; Clayton, W.A.

    1999-04-01

    The performance objectives of the Department of Energy`s Low-Level Radioactive Waste (LLW) disposal facilities located at the Nevada Test Site transcend those of any other radioactive waste disposal site in the US. Situated at the southern end of the Great Basin, 800 feet above the water table, the Area 5 Radioactive Waste Management Site (RWMS) has utilized a combination of engineered shallow land disposal cells and deep augured shafts to dispose a variety of waste streams. These include high volume low-activity wastes, classified materials, and high-specific-activity special case wastes. Twenty miles north of Area 5 is the Area 3 RWMS. Here bulk LLW disposal takes place in subsidence craters formed from underground testing of nuclear weapons. Earliest records indicate that documented LLW disposal activities have occurred at the Area 5 and Area 3 RWMS`s since 1961 and 1968, respectively. However, these activities have only been managed under a formal program since 1978. This paper describes the technical attributes of the facilities, present and future capacities and capabilities, and provides a description of the process from waste approval to final disposition. The paper also summarizes the current status of the waste disposal operations.

  3. BENEFITS OF VIBRATION ANALYSIS FOR DEVELOPMENT OF EQUIPMENT IN HLW TANKS - 12341

    SciTech Connect (OSTI)

    Stefanko, D.; Herbert, J.

    2012-01-10

    Vibration analyses of equipment intended for use in the Savannah River Site (SRS) radioactive liquid waste storage tanks are performed during pre-deployment testing and has been demonstrated to be effective in reducing the life-cycle costs of the equipment. Benefits of using vibration analysis to identify rotating machinery problems prior to deployment in radioactive service will be presented in this paper. Problems encountered at SRS and actions to correct or lessen the severity of the problem are discussed. In short, multi-million dollar cost saving have been realized at SRS as a direct result of vibration analysis on existing equipment. Vibration analysis of equipment prior to installation can potentially reduce inservice failures, and increases reliability. High-level radioactive waste is currently stored in underground carbon steel waste tanks at the United States Department of Energy (DOE) Savannah River Site and at the Hanford Site, WA. Various types of rotating machinery (pumps and separations equipment) are used to manage and retrieve the tank contents. Installation, maintenance, and repair of these pumps and other equipment are expensive. In fact, costs to remove and replace a single pump can be as high as a half million dollars due to requirements for radioactive containment. Problems that lead to in-service maintenance and/or equipment replacement can quickly exceed the initial investment, increase radiological exposure, generate additional waste, and risk contamination of personnel and the work environment. Several different types of equipment are considered in this paper, but pumps provide an initial example for the use of vibration analysis. Long-shaft (45 foot long) and short-shaft (5-10 feet long) equipment arrangements are used for 25-350 horsepower slurry mixing and transfer pumps in the SRS HLW tanks. Each pump has a unique design, operating characteristics and associated costs, sometimes exceeding a million dollars. Vibration data are routinely collected during pre-installation tests and screened for: Critical speeds or resonance, Imbalance of rotating parts, Shaft misalignment, Fluid whirl or lubrication break down, Bearing damages, and Other component abnormalities. Examples of previous changes in operating parameters and fabrication tolerances and extension of equipment life resulting from the SRS vibration analysis program include: (1) Limiting operational speeds for some pumps to extend service life without design or part changes; (2) Modifying manufacturing methods (tightening tolerances) for impellers on slurry mixing pumps based on vibration data that indicated hydraulic imbalance; (3) Identifying rolling element mounting defects and replacing those components in pump seals before installation; and (4) Identifying the need for bearing design modification for SRS long-shaft mixing pump designs to eliminate fluid whirl and critical speeds which significantly increased the equipment service life. In addition, vibration analyses and related analyses have been used during new equipment scale-up tests to identify the need for design improvements for full-scale operation / deployment of the equipment in the full size tanks. For example, vibration analyses were recently included in the rotary micro-filtration scale-up test program at SRNL.

  4. US Army facility for the consolidation of low-level radioactive waste

    SciTech Connect (OSTI)

    Stein, S.L.; Tanner, J.E.; Murphy, B.L.; Gillings, J.C.; Hadley, R.T.; Lyso, O.M.; Gilchrist, R.L.; Murphy, D.W.

    1983-12-01

    A preliminary study of a waste consolidation facility for the Department of the Army's low-level radioactive waste was carried out to determine a possible site and perform a cost-benefit analysis. Four sites were assessed as possible locations for such a facility, using predetermined site selection criteria. To assist in the selection of a site, an evaluation of environmental issues was included as part of each site review. In addition, a preliminary design for a waste consolidation facility was developed, and facilities at each site were reviewed for their availability and suitability for this purpose. Currently available processes for volume reduction, as well as processes still under development, were then investigated, and the support and handling equipment and the staff needed for the safe operation of a waste consolidation facility were studied. Using current costs for the transportation and burial of low-level waste, a cost comparison was then made between waste disposal with and without the utilization of volume reduction. Finally, regulations that could affect the operation of a waste consolidation facility were identified and their impact was assessed. 11 references, 5 figures, 16 tables.

  5. Preliminary design requirements document for Project W-378, low-level waste vitrification plant

    SciTech Connect (OSTI)

    Swanson, L.M.

    1995-03-31

    The scope of this preliminary Design Requirements Document (DRD) is to identify and define the functions, with associated requirements, which must be performed to accomplish vitrification and disposal of the pretreated low-level waste (LLW) fraction of the Hanford Site tank waste. This document sets forth function requirements, performance requirements and design constraints necessary to begin conceptual design for the Low-Level Waste Vitrification Plant (LLWVP). System and physical interfaces between the LLWVP Project and the Tank Waste Remediation System (TWRS) are identified. The constraints, performance requirements, and transfer of information and data across a technical interface will be documented in an Interface Control Document. The design requirements provided in this document will be augmented by additional detailed design data to be documented by the project.

  6. Waste Treatment Plant - 12508

    SciTech Connect (OSTI)

    Harp, Benton; Olds, Erik

    2012-07-01

    The Waste Treatment Plant (WTP) will immobilize millions of gallons of Hanford's tank waste into solid glass using a proven technology called vitrification. The vitrification process will turn the waste into a stable glass form that is safe for long-term storage. Our discussion of the WTP will include a description of the ongoing design and construction of this large, complex, first-of-a-kind project. The concept for the operation of the WTP is to separate high-level and low-activity waste fractions, and immobilize those fractions in glass using vitrification. The WTP includes four major nuclear facilities and various support facilities. Waste from the Tank Farms is first pumped to the Pretreatment Facility at the WTP through an underground pipe-in-pipe system. When construction is complete, the Pretreatment Facility will be 12 stories high, 540 feet long and 215 feet wide, making it the largest of the four major nuclear facilities that compose the WTP. The total size of this facility will be more than 490,000 square feet. More than 8.2 million craft hours are required to construct this facility. Currently, the Pretreatment Facility is 51 percent complete. At the Pretreatment Facility the waste is pumped to the interior waste feed receipt vessels. Each of these four vessels is 55-feet tall and has a 375,000 gallon capacity, which makes them the largest vessels inside the Pretreatment Facility. These vessels contain a series of internal pulse-jet mixers to keep incoming waste properly mixed. The vessels are inside the black-cell areas, completely enclosed behind thick steel-laced, high strength concrete walls. The black cells are designed to be maintenance free with no moving parts. Once hot operations commence the black-cell area will be inaccessible. Surrounded by black cells, is the 'hot cell canyon'. The hot cell contains all the moving and replaceable components to remove solids and extract liquids. In this area, there is ultrafiltration equipment, cesium-ion exchange columns, evaporator boilers and recirculation pumps, and various mechanical process pumps for transferring process fluids. During the first phase of pretreatment, the waste will be concentrated using an evaporation process. Solids will be filtered out, and the remaining soluble, highly radioactive isotopes will be removed using an ion-exchange process. The high-level solids will be sent to the High-Level Waste (HLW) Vitrification Facility, and the low activity liquids will be sent to the Low-Activity Waste (LAW) Vitrification Facility for further processing. The high-level waste will be transferred via underground pipes to the HLW Facility from the Pretreatment Facility. The waste first arrives at the wet cell, which rests inside a black-cell area. The pretreated waste is transferred through shielded pipes into a series of melter preparation and feed vessels before reaching the melters. Liquids from various facility processes also return to the wet cell for interim storage before recycling back to the Pretreatment Facility. (authors)

  7. Technology, Safety and Costs of Decommissioning a Reference Low-Level Waste Burial Ground. Appendices

    SciTech Connect (OSTI)

    1980-06-01

    Safety and cost information are developed for the conceptual decommissioning of commercial low-level waste (LLW) burial grounds. Two generic burial grounds, one located on an arid western site and the other located on a humid eastern site, are used as reference facilities for the study. The two burial grounds are assumed to have the same site capacity for waste, the same radioactive waste inventory, and similar trench characteristics and operating procedures. The climate, geology. and hydrology of the two sites are chosen to be typical of real western and eastern sites. Volume 2 (Appendices) contains the detailed analyses and data needed to support the results given in Volume 1.

  8. LOW-LEVEL RADIOACTIVE & MIXED WASTE SHIPMENTS TO THE NEVADA NATIONAL SECURITY SITE

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

    NV NY NY NV OH TN TN TN, WA, CA TN TN TN TX y ct 2 Plant Perma-Fix LOW-LEVEL RADIOACTIVE & MIXED WASTE SHIPMENTS TO THE NEVADA NATIONAL SECURITY SITE FIRST QUARTER REPORT, FY 2011 (OCTOBER, NOVEMBER, DECEMBER 2010) Livermore National Laboratory ergy Alliance onal Laboratory Mixed Waste Treatment Project gonne National Laboratory ational Laboratory Gaseous Diffusion Plant Proving Ground terra, LLC en National Laboratory y Environmental Services ecurity Technologies, Inc. h Gaseous Diffusion

  9. EIS-0110: Central Waste Disposal Facility for Low-Level Radioactive Waste, Oak Ridge Reservation, Oak Ridge, Tennessee

    Broader source: Energy.gov [DOE]

    This EIS assessed the environmental impacts of alternatives for the disposal of low-level waste and by-product materials generated by the three major plants on the Oak Ridge Reservation (ORR). In addition to the no-action alternative, two classes of alternatives were evaluated: facility design alternatives and siting alternatives. This project was cancelled after the Draft Environmental Impact Statement was issued.

  10. Performance Assessment for the Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Facility

    SciTech Connect (OSTI)

    Annette L. Schafer; A. Jeffrey Sondrup; Arthur S. Rood

    2012-05-01

    This performance assessment for the Remote-Handled Low-Level Radioactive Waste Disposal Facility at the Idaho National Laboratory documents the projected radiological dose impacts associated with the disposal of low-level radioactive waste at the facility. This assessment evaluates compliance with the applicable radiological criteria of the U.S. Department of Energy and the U.S. Environmental Protection Agency for protection of the public and the environment. The calculations involve modeling transport of radionuclides from buried waste to surface soil and subsurface media, and eventually to members of the public via air, groundwater, and food chain pathways. Projections of doses are calculated for both offsite receptors and individuals who inadvertently intrude into the waste after site closure. The results of the calculations are used to evaluate the future performance of the low-level radioactive waste disposal facility and to provide input for establishment of waste acceptance criteria. In addition, one-factor-at-a-time, Monte Carlo, and rank correlation analyses are included for sensitivity and uncertainty analysis. The comparison of the performance assessment results to the applicable performance objectives provides reasonable expectation that the performance objectives will be met

  11. Glass optimization for vitrification of Hanford Site low-level tank waste

    SciTech Connect (OSTI)

    Feng, X.; Hrma, P.R.; Westsik, J.H. Jr.

    1996-03-01

    The radioactive defense wastes stored in 177 underground single-shell tanks (SST) and double-shell tanks (DST) at the Hanford Site will be separated into low-level and high-level fractions. One technology activity underway at PNNL is the development of glass formulations for the immobilization of the low-level tank wastes. A glass formulation strategy has been developed that describes development approaches to optimize glass compositions prior to the projected LLW vitrification facility start-up in 2005. Implementation of this strategy requires testing of glass formulations spanning a number of waste loadings, compositions, and additives over the range of expected waste compositions. The resulting glasses will then be characterized and compared to processing and performance specifications yet to be developed. This report documents the glass formulation work conducted at PNL in fiscal years 1994 and 1995 including glass formulation optimization, minor component impacts evaluation, Phase 1 and Phase 2 melter vendor glass development, liquidus temperature and crystallization kinetics determination. This report also summarizes relevant work at PNNL on high-iron glasses for Hanford tank wastes conducted through the Mixed Waste Integrated Program and work at Savannah River Technology Center to optimize glass formulations using a Plackett-Burnam experimental design.

  12. Low-level waste management program and interim waste operations technologies

    SciTech Connect (OSTI)

    Mezga, L.J.

    1983-01-01

    The Department of Energy currently supports an integrated technology development and transfer program aimed at ensuring that the technology necessary for the safe management and disposal of LLW by the commercial and defense sectors is available. The program focuses on five technical areas: (1) corrective measures technology, (2) improved shallow land burial technology, (3) greater confinement disposal technology, (4) model development and validation, and (5) treatment methods for problem wastes. The results of activities in these areas are reported in the open literature and the Proceedings of the LLWMP Annual Participants Information Meeting.

  13. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM - 2011

    SciTech Connect (OSTI)

    West, B.; Waltz, R.

    2012-06-21

    Aqueous radioactive wastes from Savannah River Site (SRS) separations and vitrification processes are contained in large underground carbon steel tanks. Inspections made during 2011 to evaluate these vessels and other waste handling facilities along with evaluations based on data from previous inspections are the subject of this report. The 2011 inspection program revealed that the structural integrity and waste confinement capability of the Savannah River Site waste tanks were maintained. All inspections scheduled per SRR-LWE-2011-00026, HLW Tank Farm Inspection Plan for 2011, were completed. Ultrasonic measurements (UT) performed in 2011 met the requirements of C-ESR-G-00006, In-Service Inspection Program for High Level Waste Tanks, Rev. 3, and WSRC-TR-2002-00061, Rev.6. UT inspections were performed on Tanks 25, 26 and 34 and the findings are documented in SRNL-STI-2011-00495, Tank Inspection NDE Results for Fiscal Year 2011, Waste Tanks 25, 26, 34 and 41. A total of 5813 photographs were made and 835 visual and video inspections were performed during 2011. A potential leaksite was discovered at Tank 4 during routine annual inspections performed in 2011. The new crack, which is above the allowable fill level, resulted in no release to the environment or tank annulus. The location of the crack is documented in C-ESR-G-00003, SRS High Level Waste Tank Leaksite Information, Rev.6.

  14. Waste Package Component Design Methodology Report

    SciTech Connect (OSTI)

    D.C. Mecham

    2004-07-12

    This Executive Summary provides an overview of the methodology being used by the Yucca Mountain Project (YMP) to design waste packages and ancillary components. This summary information is intended for readers with general interest, but also provides technical readers a general framework surrounding a variety of technical details provided in the main body of the report. The purpose of this report is to document and ensure appropriate design methods are used in the design of waste packages and ancillary components (the drip shields and emplacement pallets). The methodology includes identification of necessary design inputs, justification of design assumptions, and use of appropriate analysis methods, and computational tools. This design work is subject to ''Quality Assurance Requirements and Description''. The document is primarily intended for internal use and technical guidance for a variety of design activities. It is recognized that a wide audience including project management, the U.S. Department of Energy (DOE), the U.S. Nuclear Regulatory Commission, and others are interested to various levels of detail in the design methods and therefore covers a wide range of topics at varying levels of detail. Due to the preliminary nature of the design, readers can expect to encounter varied levels of detail in the body of the report. It is expected that technical information used as input to design documents will be verified and taken from the latest versions of reference sources given herein. This revision of the methodology report has evolved with changes in the waste package, drip shield, and emplacement pallet designs over many years and may be further revised as the design is finalized. Different components and analyses are at different stages of development. Some parts of the report are detailed, while other less detailed parts are likely to undergo further refinement. The design methodology is intended to provide designs that satisfy the safety and operational requirements of the YMP. Four waste package configurations have been selected to illustrate the application of the methodology during the licensing process. These four configurations are the 21-pressurized water reactor absorber plate waste package (21-PWRAP), the 44-boiling water reactor waste package (44-BWR), the 5 defense high-level radioactive waste (HLW) DOE spent nuclear fuel (SNF) codisposal short waste package (5-DHLWDOE SNF Short), and the naval canistered SNF long waste package (Naval SNF Long). Design work for the other six waste packages will be completed at a later date using the same design methodology. These include the 24-boiling water reactor waste package (24-BWR), the 21-pressurized water reactor control rod waste package (21-PWRCR), the 12-pressurized water reactor waste package (12-PWR), the 5 defense HLW DOE SNF codisposal long waste package (5-DHLWDOE SNF Long), the 2 defense HLW DOE SNF codisposal waste package (2-MC012-DHLW), and the naval canistered SNF short waste package (Naval SNF Short). This report is only part of the complete design description. Other reports related to the design include the design reports, the waste package system description documents, manufacturing specifications, and numerous documents for the many detailed calculations. The relationships between this report and other design documents are shown in Figure 1.

  15. Low-level radioactive-waste compacts. Status report as of July 1982

    SciTech Connect (OSTI)

    Not Available

    1982-07-01

    The Low-Level Radioactive Waste Policy Act (P.L. 96-573), enacted in December 1980, established as federal policy that states take responsibility for providing disposal capacity for low-level radioactive waste (LLW) generated within their borders, except for defense waste and Federal R and D. At the request of Senator James A. McClure, Chairman of the Senate Committee on Energy and Natural Resources, DOE has documented the progress of states individually and collectively in fulfilling their responsibilities under the Public Law. Regionalization through formation of low-level waste compacts has been the primary vehicle by which many states are assuming this responsibility. To date seven low-level waste compacts have been drafted and six have been enacted by state legislatures or ratified by a governor. As indicated by national progress to date, DOE considers the task of compacting achievable by the January 1, 1986, exclusionary date set in law, although several states and NRC questioned this.

  16. Source team evaluation for radioactive low-level waste disposal performance assessment

    SciTech Connect (OSTI)

    Cowgill, M.G.; Sullivan, T.M.

    1993-01-01

    Information compiled on the low-level radioactive waste disposed at the three currently operating commercial disposal sites during the period 1987--1989 have been reviewed and processed in order to determine the total activity distribution in terms of waste stream, waste classification and waste form. The review identified deficiencies in the information currently being recorded on shipping manifests and the development of a uniform manifest is recommended (the NRC is currently developing a rule to establish a uniform manifest). The data from waste disposed during 1989 at one of the sites (Richland, WA) were more detailed than the data available during other years and at other sites, and thus were amenable to a more in-depth treatment. This included determination of the distribution of activity for each radionuclide by waste form, and thus enabled these data to be evaluated in terms of the specific needs for improved modeling of releases from waste packages. From the results, preliminary lists have been prepared of the isotopes which might be the most significant from the aspect of the development of a source term model.

  17. Leaching mechanisms of solidified low-level waste. The literature survey

    SciTech Connect (OSTI)

    Dougherty, D.; Colombo, P.

    1985-06-01

    A literature survey on leaching mechanisms, available mathematical models and factors that affect leaching from solidified low-level radioactive waste (LLW) was compiled. Physicochemical mechanisms identified include diffusion, dissolution, ion exchange, corrosion and surface effects. Diffusion was generally considered to be the predominant mechanism in LLW leachability. However, this hierarchy of importance has been strongly questioned for waste forms containing soluble salts and has been shown to be invalid for waste forms incorporating sorbents which control the release of radionuclides by ion exchange. Leaching behavior was modeled both mathematically for curve fitting to leaching data and by consideration of physical and chemical interactions within and between solidification agents, waste materials and additives, if any. Physicochemical analyses of bitumen and polymer solidification agents have considered them to be inert encapsulants with limited water permeability. All of the mathematical models are derived from solutions to the diffusion equation. Other mechanistic processes are included as additional terms in the equation. No comprehensive evaluations of mathematical models for LLW based on curve fitting to data were found in the literature. Factors that affect leaching have been categorized as system factors, leachant factors and waste form factors. System factors include temperature, pressure, radiation, time and the ratio of waste form area to leachant volume. Leachant factors include pH, Eh, flow or replacement frequency and composition while waste form factors include composition, surface condition, porosity and surface area to volume ratio. Information from the literature is reported for each of these factors. 75 refs., 2 figs., 4 tabs.

  18. ECONOMIC ASSESSMENT ON VITRIFICATION FACILITY OF LOW-AND INTERMEDIATE-LEVEL RADIOACTIVE WASTES IN KOREA

    SciTech Connect (OSTI)

    Kim, Sung Il; Lee, Kun Jai; Ji, Pyung Kook; Park, Jong Kil; Ha, Jong Hyun; Song, Myung Jae

    2003-02-27

    The usefulness of vitrification technology for low-and intermediate-level radioactive wastes was demonstrated with high volume reduction capability and good mechanical and chemical stability of final waste forms, and commercial vitrification facility is expected to be constructed at Ulchin site of Korean Nuclear Power Plant Ulchin Unit 5 and 6. Hence, overall economic assessment was necessary to find out the economic advantage of the vitrification facility and to predict the construction and operation costs of the facility on the preliminary bases. Additionally, the generation characteristics of radioactive wastes were investigated. The results of the cost analysis showed that the disposal cost of radioactive wastes treated by vitrification facility reduced to 85 percent compared with that by current waste treatment system. And the present worth analysis was performed through the cost-benefit analysis method for the commercial vitrification facility. The results showed that the vitrification facility combining cold crucible melter (CCM) for treatment of combustible DAW, spent resin, and borated liquid waste concentrate and plasma torch melter (PTM) for non-combustible DAW and spent filter is more economical than current waste treatment system when the escalation rate of disposal cost of more than 10 percent per year was applied.

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

    SciTech Connect (OSTI)

    David Duncan

    2011-03-01

    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.

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

    SciTech Connect (OSTI)

    David Duncan

    2009-10-01

    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.

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

    SciTech Connect (OSTI)

    David Duncan

    2011-04-01

    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.

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

    SciTech Connect (OSTI)

    David Duncan

    2010-06-01

    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.

  3. EM's West Valley Cleanup Finds Success in History-Making Waste Relocation

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

    | Department of Energy Workers load an overpack container into a vertical storage cask. Workers load an overpack container into a vertical storage cask. WEST VALLEY, N.Y. - For the first time in U.S. history, high-level waste (HLW) was placed in long-term, outdoor storage. The unprecedented accomplishment occurred in EM's cleanup at the West Valley Demonstration Project (WVDP). The liquid waste, which had been immobilized in glass in the vitrification process, is being relocated from the

  4. Estimating heel retrieval costs for underground storage tank waste at Hanford. Draft

    SciTech Connect (OSTI)

    DeMuth, S.

    1996-08-26

    Approximately 100 million gallons ({approx}400,000 m{sup 3}) of existing U.S. Department of Energy (DOE) owned radioactive waste stored in underground tanks can not be disposed of as low-level waste (LLW). The current plan for disposal of UST waste which can not be disposed of as LLW is immobilization as glass and permanent storage in an underground repository. Disposal of LLW generally can be done sub-surface at the point of origin. Consequently, LLW is significantly less expensive to dispose of than that requiring an underground repository. Due to the lower cost for LLW disposal, it is advantageous to separate the 100 million gallons of waste into a small volume of high-level waste (HLW) and a large volume of LLW.

  5. RAIL ROUTING - CURRENT PRACTICES FOR SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTE SHIPMENTS,

    Office of Environmental Management (EM)

    4.31 Final 7/30 RAIL ROUTING - CURRENT PRACTICES FOR SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTE SHIPMENTS, AND A COMPARATIVE ANALYSIS OF HIGHWAY REGULATORY GUIDELINES APPLIED TO RAIL A DISCUSSION PAPER PREPARED BY THE RAIL TOPIC GROUP OF THE TRANSPORTATION EXTERNAL COORDINATION WORKING GROUP JULY 2004 1 Revision 4.31 Final 7/30 Rail Routing- Current Practices for Spent Nuclear Fuel And High-Level Waste Shipments, And a Comparative Analysis Of Highway Regulatory Guidelines Applied to Rail I.

  6. SRS Crosses Halfway Mark on Closing Another High-Level Waste Tank |

    Energy Savers [EERE]

    Department of Energy SRS Crosses Halfway Mark on Closing Another High-Level Waste Tank SRS Crosses Halfway Mark on Closing Another High-Level Waste Tank February 25, 2016 - 12:20pm Addthis Work is more than halfway complete on grouting Tank 12, the eighth to be operationally closed at Savannah River Site. Work is more than halfway complete on grouting Tank 12, the eighth to be operationally closed at Savannah River Site. AIKEN, S.C. - Savannah River Site (SRS) moved past the halfway mark

  7. Improvement to low-level radioactive-waste vitrification processes. Master's thesis

    SciTech Connect (OSTI)

    Horton, W.S.

    1986-05-01

    Low-level radioactive waste vitrification (LLWV) is a technically feasible and cost-competitive alternative to the traditional immobilization options, i.e., cementation or bituminization. This thesis analyzes cementation, bituminization and vitrification, reviews the impact of the low-level Waste-stream composition on the vitrification process, then proposes and discusses several techniques to control the volatile radionuclides in a Process Improved LLWV system (PILLWV). The techniques that control the volatile radionuclides include chemical precipitation, electrodialysis, and ion exchange. Ion exchange is preferred. A comparison of the technical specifications, of the regulatory compliance, and of the cost considerations shows the PILLWV to be the superior LLW immobilization option.

  8. Low-level Waste Safely Dispositioned Under Runoff Cover at SRS | Department

    Office of Environmental Management (EM)

    of Energy Low-level Waste Safely Dispositioned Under Runoff Cover at SRS Low-level Waste Safely Dispositioned Under Runoff Cover at SRS April 26, 2011 - 12:00pm Addthis The liner installer heat-welds a sand anchor closed. The sand anchors are installed under the liner and across the length of the slit trench to keep the liner in place and minimize the effects of wind lift. The liner installer heat-welds a sand anchor closed. The sand anchors are installed under the liner and across the

  9. Development of a Rotary Microfilter for SRS HLW Applications

    SciTech Connect (OSTI)

    MICHAEL, POIRIER

    2004-11-24

    The processing rate of Savannah River Site high level waste decontamination processes are limited by the flow rate of the solid-liquid separation. The baseline process, using a 0.1 micron cross flow filter, produces 0.02 gpm/ft2 of filtrate under expected operating conditions. Savannah River National Laboratory personnel identified the rotary microfilter as a technology that could significantly increase filter flux, with throughput improvements of as much as 10X for that specific operation. With funding from the Department of Energy Office of Cleanup Technologies, SRNL personnel are evaluating and developing the rotary microfilter for radioactive service at SRS. This work includes pilot-scale and actual waste testing to evaluate system reliability, the impact of radiation on system components, the filter flux for a variety of waste streams, and relative performance for alternative filter media.

  10. DEMONSTRATION OF THE NEXT-GENERATION CAUSTIC-SIDE SOLVENT EXTRACTION SOLVENT WITH 2-CM CENTRIGUGAL CONTRACTORS USING TANK 49H WASTE AND WASTE SIMULANT

    SciTech Connect (OSTI)

    Pierce, R.; Peters, T.; Crowder, M.; Pak, D.; Fink, S.; Blessing, R.; Washington, A.; Caldwell, T.

    2011-11-29

    Researchers successfully demonstrated the chemistry and process equipment of the Caustic-Side Solvent Extraction (CSSX) flowsheet using MaxCalix for the decontamination of high level waste (HLW). The demonstration was completed using a 12-stage, 2-cm centrifugal contactor apparatus at the Savannah River National Laboratory (SRNL). This represents the first CSSX process demonstration of the MaxCalix solvent system with Savannah River Site (SRS) HLW. Two tests lasting 24 and 27 hours processed non-radioactive simulated Tank 49H waste and actual Tank 49H HLW, respectively. A solvent extraction system for removal of cesium from alkaline solutions was developed utilizing a novel solvent invented at the Oak Ridge National Laboratory (ORNL). This solvent consists of a calix[4]arene-crown-6 extractant dissolved in an inert hydrocarbon matrix. A modifier is added to the solvent to enhance the extraction power of the calixarene and to prevent the formation of a third phase. An additional additive is used to improve stripping performance and to mitigate the effects of any surfactants present in the feed stream. The process that deploys this solvent system is known as Caustic Side Solvent Extraction (CSSX). The solvent system has been deployed at the Savannah River Site (SRS) in the Modular CSSX Unit (MCU) since 2008.

  11. Conceptual Design Report for the Remote-Handled Low-Level Waste Disposal Project

    SciTech Connect (OSTI)

    David Duncan

    2011-05-01

    This conceptual design report addresses development of replacement remote-handled low-level waste disposal capability for the Idaho National Laboratory. Current disposal capability at the Radioactive Waste Management Complex is planned until the facility is full or until it must be closed in preparation for final remediation (approximately at the end of Fiscal Year 2017). This conceptual design report includes key project assumptions; design options considered in development of the proposed onsite disposal facility (the highest ranked alternative for providing continued uninterrupted remote-handled low level waste disposal capability); process and facility descriptions; safety and environmental requirements that would apply to the proposed facility; and the proposed cost and schedule for funding, design, construction, and operation of the proposed onsite disposal facility.

  12. Conceptual Design Report for the Remote-Handled Low-Level Waste Disposal Project

    SciTech Connect (OSTI)

    Lisa Harvego; David Duncan; Joan Connolly; Margaret Hinman; Charles Marcinkiewicz; Gary Mecham

    2011-03-01

    This conceptual design report addresses development of replacement remote-handled low-level waste disposal capability for the Idaho National Laboratory. Current disposal capability at the Radioactive Waste Management Complex is planned until the facility is full or until it must be closed in preparation for final remediation (approximately at the end of Fiscal Year 2017). This conceptual design report includes key project assumptions; design options considered in development of the proposed onsite disposal facility (the highest ranked alternative for providing continued uninterrupted remote-handled low level waste disposal capability); process and facility descriptions; safety and environmental requirements that would apply to the proposed facility; and the proposed cost and schedule for funding, design, construction, and operation of the proposed onsite disposal facility.

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

    SciTech Connect (OSTI)

    Eaton, W.C. [Westinghouse Hanford Co., Richland, WA (United States)] [Westinghouse Hanford Co., Richland, WA (United States)

    1995-11-01

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

  14. Conceptual Design Report for Remote-Handled Low-Level Waste Disposal Facility

    SciTech Connect (OSTI)

    Lisa Harvego; David Duncan; Joan Connolly; Margaret Hinman; Charles Marcinkiewicz; Gary Mecham

    2010-10-01

    This conceptual design report addresses development of replacement remote-handled low-level waste disposal capability for the Idaho National Laboratory. Current disposal capability at the Radioactive Waste Management Complex is planned until the facility is full or until it must be closed in preparation for final remediation (approximately at the end of Fiscal Year 2017). This conceptual design report includes key project assumptions; design options considered in development of the proposed onsite disposal facility (the highest ranked alternative for providing continued uninterrupted remote-handled low level waste disposal capability); process and facility descriptions; safety and environmental requirements that would apply to the proposed facility; and the proposed cost and schedule for funding, design, construction, and operation of the proposed onsite disposal facility.

  15. Waste Treatment and Immobilation Plant Pretreatment Facility | Department

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

    of Energy Pretreatment Facility PDF icon Summary - WTP Pretreatment Facility More Documents & Publications Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility Compilation of TRA Summaries Hanford ETR Tank Waste Treatment and Immobilization Plant - Hanford Tank Waste Treatment and Immobilization Plant Technical Review - External Flowsheet Review Team (Technical) Report

  16. Low-level radioactive waste management: transitioning to off-site disposal at Los Alamos National Laboratory

    SciTech Connect (OSTI)

    Dorries, Alison M

    2010-11-09

    Facing the closure of nearly all on-site management and disposal capability for low-level radioactive waste (LLW), Los Alamos National Laboratory (LANL) is making ready to ship the majority of LLW off-site. In order to ship off-site, waste must meet the Treatment, Storage, and Disposal Facility's (TSDF) Waste Acceptance Criteria (WAC). In preparation, LANL's waste management organization must ensure LANL waste generators characterize and package waste compliantly and waste characterization documentation is complete and accurate. Key challenges that must be addressed to successfully make the shift to off-site disposal of LLW include improving the detail, accuracy, and quality of process knowledge (PK) and acceptable knowledge (AK) documentation, training waste generators and waste management staff on the higher standard of data quality and expectations, improved WAC compliance for off-site facilities, and enhanced quality assurance throughout the process. Certification of LANL generators will allow direct off-site shipping of LLW from their facilities.

  17. Low Temperature Aluminum Dissolution Of Sludge Waste

    SciTech Connect (OSTI)

    Keefer, M.T.; Hamm, B.A.; Pike, J.A. [Washington Savannah River Company, Aiken, SC (United States)

    2008-07-01

    High Level Waste (HLW) at the Savannah River Site (SRS) is currently stored in aging underground storage tanks. This waste is a complex mixture of insoluble solids, referred to as sludge, and soluble salts. Continued long-term storage of these radioactive wastes poses an environmental risk. The sludge is currently being stabilized in the Defense Waste Processing Facility (DWPF) through a vitrification process immobilizing the waste in a borosilicate glass matrix for long-term storage in a federal repository. Without additional treatment, the existing volume of sludge would produce nearly 8000 canisters of vitrified waste. Aluminum compounds, along with other non-radioactive components, represent a significant portion of the sludge mass currently planned for vitrification processing in DWPF. Removing the aluminum from the waste stream reduces the volume of sludge requiring vitrification and improves production rates. Treating the sludge with a concentrated sodium hydroxide (caustic) solution at elevated temperatures (>90 deg. C) to remove aluminum is part of an overall sludge mass reduction effort to reduce the number of vitrified canisters, shorten the life cycle for the HLW system, and reduce the risk associated with the long term storage of radioactive wastes at SRS. A projected reduction of nearly 900 canisters will be achieved by performing aluminum dissolution on six targeted sludge batches; however, a project to develop and install equipment will not be ready for operation until 2013. The associated upgrades necessary to implement a high temperature process in existing facilities are costly and present many technical challenges. Efforts to better understand the characteristics of the sludge mass and dissolution kinetics are warranted to overcome these challenges. Opportunities to further reduce the amount of vitrified waste and increase production rates should also be pursued. Sludge staged in Tank 51 as the next sludge batch for feed to DWPF consisted primarily of radioactive wastes containing a very high aluminum concentration. Based on initial laboratory testing and previous sludge characterization, aluminum in this sludge could be dissolved at low temperature (no more than 65 deg. C) in a concentrated caustic solution. The amount of aluminum predicted to dissolve under these conditions ranged from 25% to 80%. An opportunity existed to remove a significant amount of aluminum prior to vitrification in DWPF and increase the level of understanding of the effects of caustic dissolution of aluminum at lower temperatures. This paper presents the results of a real waste laboratory demonstration and full-scale implementation of a low temperature aluminum dissolution process which should be considered as a viable means to reduce radioactive sludge mass and reduce the amount of waste to be vitrified. (authors)

  18. Lead-iron phosphate glass as a containment medium for the disposal of high-level nuclear wastes

    DOE Patents [OSTI]

    Boatner, L.A.; Sales, B.C.

    1984-04-11

    Disclosed are lead-iron phosphate glasses containing a high level of Fe/sub 2/O/sub 3/ for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste

  19. Potential problem areas: extended storage of low-level radioactive waste

    SciTech Connect (OSTI)

    Siskind, B.

    1985-01-01

    If a state or regional compact does not have adequate disposal capacity for low-level radioactive waste (LLRW), then extended storage of certain LLRW may be necessary. The Nuclear Regulatory Commission (NRC) has contracted with Brookhaven National Laboratory to address the technical issues of extended storage. The dual objectives of this study are (1) to provide practical technical assessments for NRC to consider in evaluating specific proposals for extended storage and (2) to help ensure adequate consideration by NRC, Agreement States, and licensees of potential problems that may arise from existing or proposed extended storage practices. Storage alternatives are considered in order to characterize the likely storage environments for these wastes. In particular, the range of storage alternatives considered and being implemented by the nuclear power plant utilities is described. The properties of the waste forms and waste containers are discussed. An overview is given of the performance of the waste package and its contents during storage (e.g., radiolytic gas generation, corrosion) and of the effects of extended storage on the performance of the waste package after storage (e.g., radiation-induced embrittlement of polyethylene, the weakening of steel containers by corrosion). Additional information and actions required to address these concerns, including possible mitigative measures, are discussed. 26 refs., 1 tab.

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

    SciTech Connect (OSTI)

    RAMSEY WG; GRAY MF; CALMUS RB; EDGE JA; GARRETT BG

    2011-01-13

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

  1. Pilot-scale grout production test with a simulated low-level waste

    SciTech Connect (OSTI)

    Fow, C.L.; Mitchell, D.H.; Treat, R.L.; Hymas, C.R.

    1987-05-01

    Plans are underway at the Hanford Site near Richland, Washington, to convert the low-level fraction of radioactive liquid wastes to a grout form for permanent disposal. Grout is a mixture of liquid waste and grout formers, including portland cement, fly ash, and clays. In the plan, the grout slurry is pumped to subsurface concrete vaults on the Hanford Site, where the grout will solidify into large monoliths, thereby immobilizing the waste. A similar disposal concept is being planned at the Savannah River Laboratory site. The underground disposal of grout was conducted at Oak Ridge National Laboratory between 1966 and 1984. Design and construction of grout processing and disposal facilities are underway. The Transportable Grout Facility (TGF), operated by Rockwell Hanford Operations (Rockwell) for the Department of Energy (DOE), is scheduled to grout Phosphate/Sulfate N Reactor Operations Waste (PSW) in FY 1988. Phosphate/Sulfate Waste is a blend of two low-level waste streams generated at Hanford's N Reactor. Other wastes are scheduled to be grouted in subsequent years. Pacific Northwest Laboratory (PNL) is verifying that Hanford grouts can be safely and efficiently processed. To meet this objective, pilot-scale grout process equipment was installed. On July 29 and 30, 1986, PNL conducted a pilot-scale grout production test for Rockwell. During the test, 16,000 gallons of simulated nonradioactive PSW were mixed with grout formers to produce 22,000 gallons of PSW grout. The grout was pumped at a nominal rate of 15 gpm (about 25% of the nominal production rate planned for the TGF) to a lined and covered trench with a capacity of 30,000 gallons. Emplacement of grout in the trench will permit subsequent evaluation of homogeneity of grout in a large monolith. 12 refs., 34 figs., 5 tabs.

  2. Proceedings of the Fifth Annual Participants' Information Meeting: DOE Low-Level Waste Management Program

    SciTech Connect (OSTI)

    Not Available

    1983-12-01

    The meeting consisted of the following six sessions: (1) plenary session I; (2) disposal technology; (3) characteristics and treatment of low-level waste; (4) environmental aspects and performance prediction; (5) overall summary sessions; and (6) plenary session II. Fifty two papers of the papers presented were processed for inclusion in the Energy Data Base. (ATT)

  3. LOW-LEVEL WASTE DISPOSAL FACILITY FEDERAL REVIEW GROUP EXECUTION PLAN

    Office of Environmental Management (EM)

    LOW-LEVEL WASTE DISPOSAL FACILITY FEDERAL REVIEW GROUP EXECUTION PLAN Los Alamos National Laboratory Paducah Idaho National Laboratory Hanford Site Savannah River Site Nevada National Security Site Oak Ridge Portsmouth Revision 0 September 2015 3 Contents Acronyms ........................................................................................................................................ 5 1.0 Introduction

  4. 12/2000 Low-Level Waste Disposal Capacity Report Version 2

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

    Current and Planned Low-Level Waste Disposal Capacity Report Revision 2 December 2000 U.S. Department of Energy Office of Environmental Management i TABLE OF CONTENTS EXECUTIVE SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ES-1 1.0 INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.1 Summary of Report Sections . . . . . . . . . . . . . . . . . . . . . .

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

    SciTech Connect (OSTI)

    Han, F.C.

    1996-09-30

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

  6. Environmental Assessment Offsite Thermal Treatment of Low-Level Mixed Waste

    SciTech Connect (OSTI)

    N /A

    1999-05-06

    The U.S. Department of Energy (DOE), Richland Operations Office (RL) needs to demonstrate the economics and feasibility of offsite commercial treatment of contact-handled low-level mixed waste (LLMW), containing polychlorinated biphenyls (PCBS) and other organics, to meet existing regulatory standards for eventual disposal.

  7. Geologic Descriptions for the Solid-Waste Low Level Burial Grounds

    SciTech Connect (OSTI)

    Bjornstad, Bruce N.; Lanigan, David C.

    2007-09-23

    This document provides the stratigraphic framework and six hydrogeologic cross sections and interpretations for the solid-waste Low Level Burial Grounds on the Hanford Site. Four of the new cross sections are located in the 200 West Area while the other two are located within the 200 East Area. The cross sections display sediments of the vadose zone and uppermost unconfined aquifer.

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

    SciTech Connect (OSTI)

    Larson, D.E.

    1980-09-01

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

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

    SciTech Connect (OSTI)

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

    1994-03-01

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

  10. New York State Low-Level Radioactive Waste Status Report for 1992

    SciTech Connect (OSTI)

    Attridge, T.; Rapaport, S.; Yang, Qian

    1993-06-01

    This report summarizes data on low-level radioactive waste (LLRW) generation in New York State for calendar year 1992. It is based on reports from generators that must be filed annually with the New York State Energy Research and Development Authority (Energy Authority) and on data from the US Department of Energy. The New York State Low-Level Radioactive Waste Management Act (State Act) requires LLRW generators in the State to submit annual reports detailing the classes and quantities of waste generated. This is the seventh year generators have been required to submit reports on their waste to the Energy Authority. The data are summarized in a series of tables and figures. There are three sections in the report. Section 1 covers volume, radioactivity and other characteristics of waste generated in 1992. Section 2 shows historical LLRW generation over the years and includes generators` projections for the next five years. Section 3 provides a list of all facilities for which 1992 LLRW reports were received.

  11. New York State low-level radioactive waste status report for 1998

    SciTech Connect (OSTI)

    Voelk, H.

    1999-06-01

    This report summarizes data on low-level radioactive waste (LLRW) generated in New York State: it is based on reports from generators that must be filed annually with the New York State Energy Research and Development Authority (NYSERDA) and on data from the US Department of Energy (US DOE). The New York State Low-Level Radioactive Waste Management Act (State Act) requires LLRW generators in the State to submit annual reports detailing the classes and quantities of waste generated. This is the 13th year generators have been required to submit these reports to NYSERDA. The data are summarized in a series of tables and figures. There are four sections in the report. Section 1 covers volume, activity, and other characteristics of waste shipped for disposal in 1998. Activity is the measure of a material`s radioactivity, or the number of radiation-emitting events occurring each second. Section 2 summarizes volume, activity, and other characteristics of waste held for storage as of December 31, 1998. Section 3 shows historical LLRW generation and includes generators` projections for the next five years. Section 4 provides a list, by county, of all facilities from which 1998 LLRW reports were received. 2 figs., 23 tabs.

  12. One size fits all? An assessment tool for solid waste management at local and national levels

    SciTech Connect (OSTI)

    Broitman, Dani; Ayalon, Ofira; Kan, Iddo

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Waste management schemes are generally implemented at national or regional level. Black-Right-Pointing-Pointer Local conditions characteristics and constraints are often neglected. Black-Right-Pointing-Pointer We developed an economic model able to compare multi-level waste management options. Black-Right-Pointing-Pointer A detailed test case with real economic data and a best-fit scenario is described. Black-Right-Pointing-Pointer Most efficient schemes combine clear National directives with local level flexibility. - Abstract: As environmental awareness rises, integrated solid waste management (WM) schemes are increasingly being implemented all over the world. The different WM schemes usually address issues such as landfilling restrictions (mainly due to methane emissions and competing land use), packaging directives and compulsory recycling goals. These schemes are, in general, designed at a national or regional level, whereas local conditions and constraints are sometimes neglected. When national WM top-down policies, in addition to setting goals, also dictate the methods by which they are to be achieved, local authorities lose their freedom to optimize their operational WM schemes according to their specific characteristics. There are a myriad of implementation options at the local level, and by carrying out a bottom-up approach the overall national WM system will be optimal on economic and environmental scales. This paper presents a model for optimizing waste strategies at a local level and evaluates this effect at a national level. This is achieved by using a waste assessment model which enables us to compare both the economic viability of several WM options at the local (single municipal authority) level, and aggregated results for regional or national levels. A test case based on various WM approaches in Israel (several implementations of mixed and separated waste) shows that local characteristics significantly influence WM costs, and therefore the optimal scheme is one under which each local authority is able to implement its best-fitting mechanism, given that national guidelines are kept. The main result is that strict national/regional WM policies may be less efficient, unless some type of local flexibility is implemented. Our model is designed both for top-down and bottom-up assessment, and can be easily adapted for a wide range of WM option comparisons at different levels.

  13. Analysis of accident sequences and source terms at treatment and storage facilities for waste generated by US Department of Energy waste management operations

    SciTech Connect (OSTI)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.; Folga, S.; Policastro, A.; Freeman, W.; Jackson, R.; Mishima, J.; Turner, S.

    1996-12-01

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the US Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies assessed, and the resultant radiological and chemical source terms evaluated. A personal-computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for the calculation of human health risk impacts. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated, and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. Key assumptions in the development of the source terms are identified. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also discuss specific accident analysis data and guidance used or consulted in this report.

  14. Separation, Concentration, and Immobilization of Technetium and Iodine from Alkaline Supernate Waste

    SciTech Connect (OSTI)

    James Harvey; Michael Gula

    1998-12-07

    Development of remediation technologies for the characterization, retrieval, treatment, concentration, and final disposal of radioactive and chemical tank waste stored within the Department of Energy (DOE) complex represents an enormous scientific and technological challenge. A combined total of over 90 million gallons of high-level waste (HLW) and low-level waste (LLW) are stored in 335 underground storage tanks at four different DOE sites. Roughly 98% of this waste is highly alkaline in nature and contains high concentrations of nitrate and nitrite salts along with lesser concentrations of other salts. The primary waste forms are sludge, saltcake, and liquid supernatant with the bulk of the radioactivity contained in the sludge, making it the largest source of HLW. The saltcake (liquid waste with most of the water removed) and liquid supernatant consist mainly of sodium nitrate and sodium hydroxide salts. The main radioactive constituent in the alkaline supernatant is cesium-137, but strontium-90, technetium-99, and transuranic nuclides are also present in varying concentrations. Reduction of the radioactivity below Nuclear Regulatory Commission (NRC) limits would allow the bulk of the waste to be disposed of as LLW. Because of the long half-life of technetium-99 (2.1 x 10 5 y) and the mobility of the pertechnetate ion (TcO 4 - ) in the environment, it is expected that technetium will have to be removed from the Hanford wastes prior to disposal as LLW. Also, for some of the wastes, some level of technetium removal will be required to meet LLW criteria for radioactive content. Therefore, DOE has identified a need to develop technologies for the separation and concentration of technetium-99 from LLW streams. Eichrom has responded to this DOE-identified need by demonstrating a complete flowsheet for the separation, concentration, and immobilization of technetium (and iodine) from alkaline supernatant waste.

  15. Radioactive Waste Conditioning, Immobilisation, And Encapsulation Processes And Technologies: Overview And Advances (Chapter 7)

    SciTech Connect (OSTI)

    Jantzen, Carol M.; Lee, William E.; Ojovan, Michael I.

    2012-10-19

    The main immobilization technologies that are available commercially and have been demonstrated to be viable are cementation, bituminization, and vitrification. 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 alkali borosilicate glass or alkali aluminophosphate glass. The exact compositions of nuclear waste glasses are tailored for easy preparation and melting, avoidance of glass-in-glass phase separation, avoidance of uncontrolled crystallization, and acceptable chemical durability, e.g., leach resistance. Glass has also been used to stabilize a variety of low level wastes (LLW) and mixed (radioactive and hazardous) low level wastes (MLLW) from other sources such as fuel rod cladding/decladding processes, chemical separations, radioactive sources, radioactive mill tailings, contaminated soils, medical research applications, and other commercial processes. The sources of radioactive waste generation are captured in other chapters in this book regarding the individual practices in various countries (legacy wastes, currently generated wastes, and future waste generation). Future waste generation is primarily driven by interest in sources of clean energy and this has led to an increased interest in advanced nuclear power production. The development of advanced wasteforms is a necessary component of the new nuclear power plant (NPP) flowsheets. Therefore, advanced nuclear wasteforms are being designed for robust disposal strategies. A brief summary is given of existing and advanced wasteforms: glass, glass-ceramics, glass composite materials (GCM’s), and crystalline ceramic (mineral) wasteforms that chemically incorporate radionuclides and hazardous species atomically in their structure. Cementitious, geopolymer, bitumen, and other encapsulant wasteforms and composites that atomically bond and encapsulate wastes are also discussed. The various processing technologies are cross-referenced to the various types of wasteforms since often a particular type of wasteform can be made by a variety of different processing technologies.

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

    SciTech Connect (OSTI)

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

    2011-10-01

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

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

    SciTech Connect (OSTI)

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

    1996-02-21

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

  18. A data base for low-level radioactive waste disposal sites

    SciTech Connect (OSTI)

    Daum, M.L.; Moskowitz, P.D.

    1989-07-01

    A computerized database was developed to assist the US Environmental Protection Agency (EPA) in evaluating methods and data for characterizing health hazards associated with land and ocean disposal options for low-level radioactive wastes. The data cover 1984 to 1987. The types of sites considered include Nuclear Regulatory Commission (NRC) licensed commercial disposal sites, EPA National Priority List (NPL) sites, US Department of Energy (DOE) Formerly Utilized Sites Remedial Action Project (FUSRAP) and DOE Surplus Facilities Management Program (SFMP) sites, inactive US ocean disposal sites, and DOE/Department of Defense facilities. Sources of information include reports from EPA, the US Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC), as well as direct communication with individuals associated with specific programs. The data include site descriptions, waste volumes and activity levels, and physical and radiological characterization of low-level wastes. Additional information on mixed waste, packaging forms, and disposal methods were compiled, but are not yet included in the database. 55 refs., 4 figs., 2 tabs.

  19. Gas generation from low-level radioactive waste: Concerns for disposal

    SciTech Connect (OSTI)

    Siskind, B.

    1992-01-01

    The Advisory Committee on Nuclear Waste (ACNW) has urged the Nuclear Regulatory Commission (NRC) to reexamine the topic of hydrogen gas generation from low-level radioactive waste (LLW) in closed spaces to ensure that the slow buildup of hydrogen from water-bearing wastes in sealed containers does not become a problem for long-term safe disposal. Brookhaven National Laboratory (BNL) has prepared a report, summarized in this paper, for the NRC to respond to these concerns. The paper discusses the range of values for G(H{sub 2}) reported for materials of relevance to LLW disposal; most of these values are in the range of 0.1 to 0.6. Most studies of radiolytic hydrogen generation indicate a leveling off of pressurization, probably because of chemical kinetics involving, in many cases, the radiolysis of water within the waste. Even if no leveling off occurs, realistic gas leakage rates (indicating poor closure by gaskets on drums and liners) will result in adequate relief of pressure for radiolytic gas generation from the majority of commercial sector LLW packages. Biodegradative gas generation, however, could pose a pressurization hazard even at realistic gas leakage rates. Recommendations include passive vents on LLW containers (as already specified for high integrity containers) and upper limits to the G values and/or the specific activity of the LLW.

  20. Gas generation from low-level radioactive waste: Concerns for disposal

    SciTech Connect (OSTI)

    Siskind, B.

    1992-04-01

    The Advisory Committee on Nuclear Waste (ACNW) has urged the Nuclear Regulatory Commission (NRC) to reexamine the topic of hydrogen gas generation from low-level radioactive waste (LLW) in closed spaces to ensure that the slow buildup of hydrogen from water-bearing wastes in sealed containers does not become a problem for long-term safe disposal. Brookhaven National Laboratory (BNL) has prepared a report, summarized in this paper, for the NRC to respond to these concerns. The paper discusses the range of values for G(H{sub 2}) reported for materials of relevance to LLW disposal; most of these values are in the range of 0.1 to 0.6. Most studies of radiolytic hydrogen generation indicate a leveling off of pressurization, probably because of chemical kinetics involving, in many cases, the radiolysis of water within the waste. Even if no leveling off occurs, realistic gas leakage rates (indicating poor closure by gaskets on drums and liners) will result in adequate relief of pressure for radiolytic gas generation from the majority of commercial sector LLW packages. Biodegradative gas generation, however, could pose a pressurization hazard even at realistic gas leakage rates. Recommendations include passive vents on LLW containers (as already specified for high integrity containers) and upper limits to the G values and/or the specific activity of the LLW.

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

    SciTech Connect (OSTI)

    S.M. Frank

    2011-09-01

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

  2. Greater-than-Class C low-level radioactive waste characterization. Appendix A-3: Basis for greater-than-Class C low-level radioactive waste light water reactor projections

    SciTech Connect (OSTI)

    Mancini, A.; Tuite, P.; Tuite, K.; Woodberry, S.

    1994-09-01

    This study characterizes low-level radioactive waste types that may exceed Class C limits at light water reactors, estimates the amounts of waste generated, and estimates radionuclide content and distribution within the waste. Waste types that may exceed Class C limits include metal components that become activated during operations, process wastes such as cartridge filters and decontamination resins, and activated metals from decommissioning activities. Operating parameters and current management practices at operating plants are reviewed and used to estimate the amounts of low-level waste exceeding Class C limits that is generated per fuel cycle, including amounts of routinely generated activated metal components and process waste. Radionuclide content is calculated for specific activated metals components. Empirical data from actual low-level radioactive waste are used to estimate radionuclide content for process wastes. Volumes and activities are also estimated for decommissioning activated metals that exceed Class C limits. To estimate activation levels of decommissioning waste, six typical light water reactors are modeled and analyzed. This study does not consider concentration averaging.

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

    Broader source: Energy.gov [DOE]

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

  4. Mixed and low-level waste treatment project: Appendix C, Health and safety criteria for the mixed and low-level waste treatment facility at the Idaho National Engineering Laboratory. Part 1, Waste streams and treatment technologies

    SciTech Connect (OSTI)

    Neupauer, R.M.; Thurmond, S.M.

    1992-09-01

    This report describes health and safety concerns associated with the Mixed and Low-level Waste Treatment Facility at the Idaho National Engineering Laboratory. Various hazards are described such as fire, electrical, explosions, reactivity, temperature, and radiation hazards, as well as the potential for accidental spills, exposure to toxic materials, and other general safety concerns.

  5. Evaluation and compilation of DOE [Department of Energy] waste package test data; Biannual report, February 1988--July 1988

    SciTech Connect (OSTI)

    Interrante, C.; Escalante, E.; Fraker, A.; Plante, E.

    1989-10-01

    This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six month period February 1988 through July 1988. Activities for the DOE Materials Characterization Center are reviewed for the period January 1988 through June 1988. A summary is given of the Yucca Mountain, Nevada disposal site activities. Short discussions relating to the reviewed publications are given and complete reviews and evaluations are included. 20 refs., 1 fig., 1 tab.

  6. Potential for Subsidence at the Low-level Waste Disposal Area

    SciTech Connect (OSTI)

    Keck, Karen Nina; Seitz, Roger Ray

    2002-09-01

    U.S. Department of Energy (DOE) Order 435.1, Radioactive Waste Management requires that DOE low-level radioactive waste (LLW) disposal facilities receive a Disposal Authorization Statement (DAS) from DOE-Headquarters. The DAS for the LLW disposal facility at the Radioactive Waste Management Complex (RWMC) at the Idaho National Engineering and Environmental Laboratory (INEEL) was granted in April 2000 and included a number of conditions that must be addressed. A maintenance plan (Schuman 2000) was prepared that identifies the tasks to be completed to address the conditions in the DAS as well as a schedule for their completion. The need for a subsidence analysis was one of the conditions identified for the DAS, and thus, a task to prepare a subsidence analysis was included in the maintenance plan. This document provides the information necessary to satisfy that requirement.

  7. Potential for Subsidence at the Low-Level Radioactive Waste Disposal Area

    SciTech Connect (OSTI)

    Keck, K.A.; Seitz, R.R.

    2002-09-26

    U.S. Department of Energy (DOE) Order 435.1, Radioactive Waste Management requires that DOE low-level radioactive waste (LLW) disposal facilities receive a Disposal Authorization Statement (DAS) from DOE-Headquarters. The DAS for the LLW disposal facility at the Radioactive Waste Management Complex (RWMC) at the Idaho National Engineering and Environmental Laboratory (INEEL) was granted in April 2000 and included a number of conditions that must be addressed. A maintenance plan (Schuman 2000) was prepared that identifies the tasks to be completed to address the conditions in the DAS as well as a schedule for their completion. The need for a subsidence analysis was one of the conditions identified for the DAS, and thus, a task to prepare a subsidence analysis was included in the maintenance plan. This document provides the information necessary to satisfy that requirement.

  8. A comparison of solidification media for the stabilization of low- level radioactive wastes

    SciTech Connect (OSTI)

    Cowgill, M.G.

    1991-10-01

    When requirements exist to stabilize low-level radioactive waste (LLW) prior to disposal, efforts to achieve this stability often center on the mixing of the waste with a solidification medium. Although historically the medium of choice has been based on the use of portland cement as the binder material, several other options have been developed and subsequently implemented. These include thermoplastic polymers, thermosetting polymers and gypsum. No one medium has thus far been successful in providing stability to all forms of LLW. The characteristics and attributes of these different binder materials are reviewed and compared. The aspects examined include availability of information, limitations to use, sensitivity to process or waste chemistry changes, radionuclide retention ability, modeling of radionuclide release processes, ease and safety of use, and relative costs.

  9. SPONTANEOUS CATALYTIC WET AIR OXIDATION DURING PRE-TREATMENT OF HIGH-LEVEL RADIOACTIVE WASTE SLUDGE

    SciTech Connect (OSTI)

    Koopman, D.; Herman, C.; Pareizs, J.; Bannochie, C.; Best, D.; Bibler, N.; Fellinger, T.

    2009-10-01

    Savannah River Remediation, LLC (SRR) operates the Defense Waste Processing Facility for the U.S. Department of Energy at the Savannah River Site. This facility immobilizes high-level radioactive waste through vitrification following chemical pretreatment. Catalytic destruction of formate and oxalate ions to carbon dioxide has been observed during qualification testing of non-radioactive analog systems. Carbon dioxide production greatly exceeded hydrogen production, indicating the occurrence of a process other than the catalytic decomposition of formic acid. Statistical modeling was used to relate the new reaction chemistry to partial catalytic wet air oxidation of both formate and oxalate ions driven by the low concentrations of palladium, rhodium, and/or ruthenium in the waste. Variations in process conditions led to increases or decreases in the total oxidative destruction, as well as partially shifting the preferred species undergoing destruction from oxalate ion to formate ion.

  10. Choosing solidification or vitrification for low-level radioactive and mixed waste treatment

    SciTech Connect (OSTI)

    Gimpel, R.F.

    1992-02-14

    Solidification (making concrete) and vitrification (making glass) are frequently the treatment methods recommended for treating inorganic or radioactive wastes. Solidification is generally perceived as the most economical treatment method. Whereas, vitrification is considered (by many) as the most effective of all treatment methods. Unfortunately, vitrification has acquired the stigma that it is too expensive to receive further consideration as an alternative to solidification in high volume treatment applications. Ironically, economic studies, as presented in this paper, show that vitrification may be more competitive in some high volume applications. Ex-situ solidification and vitrification are the competing methods for treating in excess of 450 000m{sup 3} of low-level radioactive and mixed waste at the Fernald Environmental Management Project (FEMP or simply, Fernald) located near Cincinnati, Ohio. This paper summarizes how Fernald is choosing between solidification and vitrification as the primary waste treatment method.

  11. New York State low-level radioactive waste status report for 1997

    SciTech Connect (OSTI)

    1998-06-01

    This report summarizes data on low-level radioactive waste (LLRW) generated in New York State. It is based on reports from generators that must be filed annually with the New York State Energy Research and Development Authority (NYSERDA) and on data from the US Department of Energy (US DOE). The data are summarized in a series of tables and figures. There are four sections in this report. Section 1 covers volume, activity, and other characteristics of waste shipped for disposal in 1997. (Activity is the measure of a material`s radioactivity, or the number of radiation-emitting events occurring each second.) Section 2 summarizes volume, activity, and other characteristics of waste held for storage as of December 31, 1997. Section 3 shows historical LLRW generation and includes generators` projections for the next five years. Section 4 provides a list, by county, of all facilities from which 1997 LLRW reports were received.

  12. Estimating costs of low-level radioactive waste disposal alternatives for the Commonwealth of Massachusetts

    SciTech Connect (OSTI)

    Not Available

    1994-02-01

    This report was prepared for the Commonwealth of Massachusetts by the Idaho National Engineering Laboratory, National Low-Level Waste Management Program. It presents planning life-cycle cost (PLCC) estimates for four sizes of in-state low-level radioactive waste (LLRW) disposal facilities. These PLCC estimates include preoperational and operational expenditures, all support facilities, materials, labor, closure costs, and long-term institutional care and monitoring costs. It is intended that this report bc used as a broad decision making tool for evaluating one of the several complex factors that must be examined when deciding between various LLRW management options -- relative costs. Because the underlying assumptions of these analyses will change as the Board decides how it will manage Massachusett`s waste and the specific characteristics any disposal facility will have, the results of this study are not absolute and should only be used to compare the relative costs of the options presented. The disposal technology selected for this analysis is aboveground earth-mounded vaults. These vaults are reinforced concrete structures where low-level waste is emplaced and later covered with a multi-layered earthen cap. The ``base case`` PLCC estimate was derived from a preliminary feasibility design developed for the Illinois Low-Level Radioactive Waste Disposal Facility. This PLCC report describes facility operations and details the procedure used to develop the base case PLCC estimate for each facility component and size. Sensitivity analyses were performed on the base case PLCC estimate by varying several factors to determine their influences upon the unit disposal costs. The report presents the results of the sensitivity analyses for the five most significant cost factors.

  13. Studies Related to Chemical Mechanisms of Gas Formation in Hanford High-Level Nuclear Wastes

    SciTech Connect (OSTI)

    E. Kent Barefield; Charles L. Liotta; Henry M. Neumann

    2002-04-08

    The objective of this work is to develop a more detailed mechanistic understanding of the thermal reactions that lead to gas production in certain high-level waste storage tanks at the Hanford, Washington site. Prediction of the combustion hazard for these wastes and engineering parameters for waste processing depend upon both a knowledge of the composition of stored wastes and the changes that they undergo as a result of thermal and radiolytic decomposition. Since 1980 when Delagard first demonstrated that gas production (H2and N2O initially, later N2 and NH3)in the affected tanks was related to oxidative degradation of metal complexants present in the waste, periodic attempts have been made to develop detailed mechanisms by which the gases were formed. These studies have resulted in the postulation of a series of reactions that account for many of the observed products, but which involve several reactions for which there is limited, or no, precedent. For example, Al(OH)4 has been postulated to function as a Lewis acid to catalyze the reaction of nitrite ion with the metal complexants, NO is proposed as an intermediate, and the ratios of gaseous products may be a result of the partitioning of NO between two or more reactions. These reactions and intermediates have been the focus of this project since its inception in 1996.

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

    SciTech Connect (OSTI)

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

    1991-12-31

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

  15. Next Generation Extractants for Cesium Separation from High-Level Waste

    SciTech Connect (OSTI)

    Moyer, Bruce A; Bazelaire, Eve; Bonnesen, Peter V; Custelcean, Radu; Delmau, Laetitia Helene; Ditto, Mary E; Engle, Nancy L; Gorbunova, Maryna; Haverlock, Tamara; Levitskaia, Tatiana G.; Bartsch, Richard A.; Surowiec, Malgorzata A.; Marquez, Manuel; Zhou, Hui

    2006-01-01

    This project seeks a fundamental understanding and major improvement in cesium separation from high-level waste by cesium-selective calixcrown extractants. Systems of particular interest involve novel solvent-extraction systems containing specific members of the calix[4]arene-crown-6 family, alcohol solvating agents, and alkylamines. Questions being addressed bear upon cesium binding strength, extraction selectivity, cesium stripping, and extractant solubility. Enhanced properties in this regard will specifically benefit applied projects funded by the USDOE Office of Environmental Management to clean up sites such as the Savannah River Site (SRS), Hanford, and the Idaho National Environmental and Engineering Laboratory. The most direct beneficiary will be the SRS Salt Processing Project, which has recently identified the Caustic-Side Solvent Extraction (CSSX) process employing a calixcrown as its preferred technology for cesium removal from SRS high-level tank waste. Disposal of high-level waste is horrendously expensive, in large part because the actual radioactive matter in underground waste tanks at various USDOE sites has been diluted over 1000-fold by ordinary inorganic chemicals. To vitrify the entire mass of the high-level waste would be prohibitively expensive. Accordingly, an urgent need has arisen for technologies to remove radionuclides such as {sup 137}Cs from the high-level waste so that the bulk of it may be diverted to cheaper low-level waste forms and cheaper storage. To address this need in part, chemical research at Oak Ridge National Laboratory (ORNL) has focused on calixcrown extractants, molecules that combine a crown ether with a calixarene. This hybrid possesses a cavity that is highly complementary for the Cs{sup +} ion vs. the Na+ ion, making it possible to cleanly separate cesium from wastes that contain 10,000- to 1,000,000-fold higher concentrations of sodium. Previous EMSP results in Project 55087 elucidated the underlying extraction equilibria in cesium nitrate extraction by the calixcrown used in the CSSX process, calix[4]arene-bis(t-octylbenzo-crown-6), designated here as BOBCalixC6 (see structure). This understanding led to key improvements in the development of the CSSX process under the EM Efficient Separations and Crosscutting Program, entailing a method to back-extract or 'strip' cesium from the calixcrown subsequent to cesium extraction from waste. Having this stripping method allowed the cesium to be concentrated in a relatively pure aqueous stream and the extractant to be regenerated for recycle. Closing the cycle then made possible the design of a process flowsheet and successful demonstration through collaboration with Argonne National Laboratory and Savannah River Technology Center under funding from the USDOE Office of Project Completion and Tanks Focus Area. Despite these successes, the CSSX process represents young technology that can benefit substantially from further fundamental inquiry. First, reversibility of the process (stripping efficiency) still presents the greatest potential for problems and the greatest potential for improvement. Second, although the calixcrown extractants for cesium are two orders of magnitude stronger than the next best simple crown ether, a minor fraction of the extractant capacity is utilized. Third, potassium competes significantly with cesium for the calixcrown binding site, an important issue in dealing with Hanford wastes having potassium concentrations as high as 1 M. Fourth, the calixcrown solubility needs to be improved. And finally, the mechanism of extraction must be understood in detail to provide the base of knowledge from which further development of the technology can be rationally made.

  16. FINAL REPORT REGULATORY OFF GAS EMISSIONS TESTING ON THE DM1200 MELTER SYSTEM USING HLW AND LAW SIMULANTS VSL-05R5830-1 REV 0 10/31/05

    SciTech Connect (OSTI)

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

    2011-12-29

    The operational requirements for the River Protection Project - Waste Treatment Plant (RPP-WTP) Low Activity Waste (LAW) and High Level Waste (HLW) melter systems, together with the feed constituents, impose a number of challenges to the off-gas treatment system. The system must be robust from the standpoints of operational reliability and minimization of maintenance. The system must effectively control and remove a wide range of solid particulate matter, acid mists and gases, and organic constituents (including those arising from products of incomplete combustion of sugar and organics in the feed) to concentration levels below those imposed by regulatory requirements. The baseline design for the RPP-WTP LAW primary off-gas system includes a submerged bed scrubber (SBS), a wet electrostatic precipitator (WESP), and a high efficiency particulate air (HEPA) filter. The secondary off-gas system includes a sulfur-impregnated activated carbon bed (AC-S), a thermal catalytic oxidizer (TCO), a single-stage selective catalytic reduction NOx treatment system (SCR), and a packed-bed caustic scrubber (PBS). The baseline design for the RPP-WTP HLW primary off-gas system includes an SBS, a WESP, a high efficiency mist eliminator (HEME), and a HEPA filter. The HLW secondary off-gas system includes a sulfur-impregnated activated carbon bed, a silver mordenite bed, a TCO, and a single-stage SCR. The one-third scale HLW DM1200 Pilot Melter installed at the Vitreous State Laboratory (VSL) was equipped with a prototypical off-gas train to meet the needs for testing and confirmation of the performance of the baseline off-gas system design. Various modifications have been made to the DM1200 system as the details of the WTP design have evolved, including the installation of a silver mordenite column and an AC-S column for testing on a slipstream of the off-gas flow; the installation of a full-flow AC-S bed for the present tests was completed prior to initiation of testing. The DM1200 system was reconfigured to enable testing of the baseline HLW or LAW off-gas trains to perform off-gas emissions testing with both LAW and HLW simulants in the present work. During 2002 and 2003, many of these off-gas components were tested individually and in an integrated manner with the DM1200 Pilot Melter. Data from these tests are being used to support engineering design confirmation and to provide data to support air permitting activities. In fiscal year 2004, the WTP Project was directed by the Office of River Protection (ORP) to comply with Environmental Protection Agency (EPA) Maximum Achievable Control Technology (MACT) requirements for organics. This requires that the combined melter and off-gas system have destruction and removal efficiency (DRE) of >99.99% for principal organic dangerous constituents (PODCs). In order to provide confidence that the melter and off-gas system are able to achieve the required DRE, testing has been directed with both LAW and HLW feeds. The tests included both 'normal' and 'challenge' WTP melter conditions in order to obtain data for the potential range of operating conditions for the WTP melters and off-gas components. The WTP Project, Washington State Department of Ecology, and ORP have agreed that naphthalene will be used for testing to represent semi-volatile organics and allyl alcohol will be used to represent volatile organics. Testing was also performed to determine emissions of halides, metals, products of incomplete combustion (PICs), dioxins, furans, coplanar PCBs, total hydrocarbons, and COX and NOX, as well as the particle size distribution (PSD) of particulate matter discharged at the end of the off-gas train. A description of the melter test requirements and analytical methods used is provided in the Test Plan for this work. Test Exceptions were subsequently issued which changed the TCO catalyst, added total organic emissions (TOE) to exhaust sampling schedule, and allowing modification of the test conditions in response to attainable plenum temperatures as well as temperature increases in the sulfur impregnated activated carbon (AC-S) column. Data are provided in this final report for all the required emission samples as well as melter and off-gas conditions during all the sampling periods. Appended to this report are previously issued VSL Letter Reports on method development for monitoring allyl alcohol in melter exhaust streams, on the results of characterization of the selected AC-S carbon media (Donnau BAT37), and on DM1200 off-line tests on the AC-S bed; also appended are reports from Air Tech on emissions sampling, and reports from Keika Ventures on validation of analytical data provided by Severn Trent Laboratories of Knoxville, Tennessee.

  17. The residuals analysis project: Evaluating disposal options for treated mixed low-level waste

    SciTech Connect (OSTI)

    Waters, R.D.; Gruebel, M.M.; Case, J.T.; Letourneau, M.J.

    1997-03-01

    For almost four years, the U.S. Department of Energy (DOE) through its Federal Facility Compliance Act Disposal Workgroup has been working with state regulators and governors` offices to develop an acceptable configuration for disposal of its mixed low-level waste (MLLW). These interactions have resulted in screening the universe of potential disposal sites from 49 to 15 and conducting ``performance evaluations`` for those fifteen sites to estimate their technical capabilities for disposal of MLLW. In the residuals analysis project, we estimated the volume of DOE`s MLLW that will require disposal after treatment and the concentrations of radionuclides in the treated waste. We then compared the radionuclide concentrations with the disposal limits determined in the performance evaluation project for each of the fifteen sites. The results are a scoping-level estimate of the required volumetric capacity for MLLW disposal and the identification of waste streams that may pose problems for disposal based on current treatment plans. The analysis provides technical information for continued discussions between the DOE and affected States about disposal of MLLW and systematic input to waste treatment developers on disposal issues.

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

    SciTech Connect (OSTI)

    NONE

    1996-08-01

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

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

    SciTech Connect (OSTI)

    NONE

    1996-01-01

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

  20. E AREA LOW LEVEL WASTE FACILITY DOE 435.1 PERFORMANCE ASSESSMENT

    SciTech Connect (OSTI)

    Wilhite, E

    2008-03-31

    This Performance Assessment for the Savannah River Site E-Area Low-Level Waste Facility was prepared to meet requirements of Chapter IV of the Department of Energy Order 435.1-1. The Order specifies that a Performance Assessment should provide reasonable assurance that a low-level waste disposal facility will comply with the performance objectives of the Order. The Order also requires assessments of impacts to water resources and to hypothetical inadvertent intruders for purposes of establishing limits on radionuclides that may be disposed near-surface. According to the Order, calculations of potential doses and releases from the facility should address a 1,000-year period after facility closure. The point of compliance for the performance measures relevant to the all pathways and air pathway performance objective, as well as to the impact on water resources assessment requirement, must correspond to the point of highest projected dose or concentration beyond a 100-m buffer zone surrounding the disposed waste following the assumed end of active institutional controls 100 years after facility closure. During the operational and institutional control periods, the point of compliance for the all pathways and air pathway performance measures is the SRS boundary. However, for the water resources impact assessment, the point of compliance remains the point of highest projected dose or concentration beyond a 100-m buffer zone surrounding the disposed waste during the operational and institutional control periods. For performance measures relevant to radon and inadvertent intruders, the points of compliance are the disposal facility surface for all time periods and the disposal facility after the assumed loss of active institutional controls 100 years after facility closure, respectively. The E-Area Low-Level Waste Facility is located in the central region of the SRS known as the General Separations Area. It is an elbow-shaped, cleared area, which curves to the northwest, situated immediately north of the Mixed Waste Management Facility. The E-Area Low-Level Waste Facility is comprised of 200 acres for waste disposal and a surrounding buffer zone that extends out to the 100-m point of compliance. Disposal units within the footprint of the low-level waste facilities include the Slit Trenches, Engineered Trenches, Component-in-Grout Trenches, the Low-Activity Waste Vault, the Intermediate-Level Vault, and the Naval Reactor Component Disposal Area. Radiological waste disposal operations at the E-Area Low-Level Waste Facility began in 1994. E-Area Low-Level Waste Facility closure will be conducted in three phases: operational closure, interim closure, and final closure. Operational closure will be conducted during the 25-year operation period (30-year period for Slit and Engineered Trenches) as disposal units are filled; interim closure measures will be taken for some units. Interim closure will take place following the end of operations and will consist of an area-wide runoff cover along with additional grading over the trench units. Final closure of all disposal units in the E-Area Low-Level Waste Facility will take place at the end of the 100-year institutional control period and will consist of the installation of an integrated closure system designed to minimize moisture contact with the waste and to serve as a deterrent to intruders. Radiological dose to human receptors is analyzed in this PA in the all-pathways analysis, the inadvertent intruder analysis and the air pathway analysis, and the results are compared to the relevant performance measures. For the all-pathways analysis, the performance measure of relevance is a 25-mrem/yr EDE to representative members of the public, excluding dose from radon and its progeny in air. For the inadvertent intruder, the applicable performance measures are 100-mrem/yr EDE and 500 mrem/yr EDE for chronic and exposure scenarios, respectively. The relevant performance measure for the air pathway is 10-mrem/yr EDE via the air pathway, excluding dose from radon and its progeny in air. Protecti

  1. Protecting Lake Ontario - Treating Wastewater from the Remediated Low-Level Radioactive Waste Management Facility - 13227

    SciTech Connect (OSTI)

    Freihammer, Till; Chaput, Barb; Vandergaast, Gary; Arey, Jimi

    2013-07-01

    The Port Granby Project is part of the larger Port Hope Area Initiative, a community-based program for the development and implementation of a safe, local, long-term management solution for historic low level radioactive waste (LLRW) and marginally contaminated soils (MCS). The Port Granby Project involves the relocation and remediation of up to 0.45 million cubic metres of such waste from the current Port Granby Waste Management Facility located in the Municipality of Clarington, Ontario, adjacent to the shoreline of Lake Ontario. The waste material will be transferred to a new suitably engineered Long-Term Waste Management Facility (LTWMF) to be located inland approximately 700 m from the existing site. The development of the LTWMF will include construction and commissioning of a new Wastewater Treatment Plant (WWTP) designed to treat wastewater consisting of contaminated surface run off and leachate generated during the site remediation process at the Port Granby Waste Management Facility as well as long-term leachate generated at the new LTWMF. Numerous factors will influence the variable wastewater flow rates and influent loads to the new WWTP during remediation. The treatment processes will be comprised of equalization to minimize impacts from hydraulic peaks, fine screening, membrane bioreactor technology, and reverse osmosis. The residuals treatment will comprise of lime precipitation, thickening, dewatering, evaporation and drying. The distribution of the concentration of uranium and radium - 226 over the various process streams in the WWTP was estimated. This information was used to assess potential worker exposure to radioactivity in the various process areas. A mass balance approach was used to assess the distribution of uranium and radium - 226, by applying individual contaminant removal rates for each process element of the WTP, based on pilot scale results and experience-based assumptions. The mass balance calculations were repeated for various flow and load scenarios. (authors)

  2. Melter technology evaluation for vitrification of Hanford Site low-level waste

    SciTech Connect (OSTI)

    Wilson, C.N.; Burgard, K.C.; Weber, E.T.; Brown, N.R.

    1995-04-01

    The current plan at the Hanford Site, in accordance with the Tri-Party Agreement among Washington State, the US Environmental Protection Agency, and the US Department of Energy, is to convert the low-level tank waste fraction into a silicate glass. The low-level waste will be composed primarily of sodium nitrate and nitrite salts concentrated in a highly alkaline aqueous solution. The capability to process up to 200 metric tons/day off glass will be established to produce an estimated 210,000 m{sup 3} for onsite disposal. A program to test and evaluate high-capacity melter technologies is in progress. Testing performed by seven different industrial sources using Joule heating, combustion, plasma, and carbon arc melters is described.

  3. Polybrominated diphenyl ethers in e-waste: Level and transfer in a typical e-waste recycling site in Shanghai, Eastern China

    SciTech Connect (OSTI)

    Li, Yue; Duan, Yan-Ping, E-mail: duanyanping@tongji.edu.cn; Huang, Fan; Yang, Jing; Xiang, Nan; Meng, Xiang-Zhou; Chen, Ling

    2014-06-01

    Highlights: • PBDEs were detected in the majority of e-waste. • PBDEs were found in TVs made in China after 1990. • The levels of ?PBDEs in e-waste made in Japan far exceed the threshold limit of RoHS. • The inappropriate recycling and disposal of e-waste is an important source of PBDEs. - Abstract: Very few data for polybrominated diphenyl ethers (PBDEs) were available in the electronic waste (e-waste) as one of the most PBDEs emission source. This study reported concentrations of PBDEs in e-waste including printer, rice cooker, computer monitor, TV, electric iron and water dispenser, as well as dust from e-waste, e-waste dismantling workshop and surface soil from inside and outside of an e-waste recycling plant in Shanghai, Eastern China. The results showed that PBDEs were detected in the majority of e-waste, and the concentrations of ?PBDEs ranged from not detected to 175 g/kg, with a mean value of 10.8 g/kg. PBDEs were found in TVs made in China after 1990. The mean concentrations of ?PBDEs in e-waste made in Korea, Japan, Singapore and China were 1.84 g/kg, 20.5 g/kg, 0.91 g/kg, 4.48 g/kg, respectively. The levels of ?PBDEs in e-waste made in Japan far exceed the threshold limit of RoHS (1.00 g/kg). BDE-209 dominated in e-waste, accounting for over 93%. The compositional patterns of PBDEs congeners resembled the profile of Saytex 102E, indicating the source of deca-BDE. Among the samples of dust and surface soil from a typical e-waste recycling site, the highest concentrations of ?{sub 18}PBDEs and BDE-209 were found in dust in e-waste, ranging from 1960 to 340,710 ng/g and from 910 to 320,400 ng/g, which were 1–2 orders of magnitude higher than other samples. It suggested that PBDEs released from e-waste via dust, and then transferred to surrounding environment.

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

    SciTech Connect (OSTI)

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

    1996-09-30

    The Failure Modes and Release Quantities of the Hanford High Level Waste Tanks due to postulated accident loads were established by a DELPHI Expert Panel consisting of both on-site and off-site experts in the field of Structure and Release. The Report presents the evaluation process, accident loads, tank structural failure conclusion reached by the panel during the two-day meeting.

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

    SciTech Connect (OSTI)

    Ross, B.

    1989-02-01

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

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

    SciTech Connect (OSTI)

    Gdowski, G.E.; Bullen, D.B. )

    1988-08-01

    Six alloys are being considered as possible materials for the fabrication of containers for the disposal of high-level radioactive waste. Three of these candidate materials are copper-based alloys: CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). The other three are iron- to nickel-based austenitic materials: Types 304L and 316L stainless steels and Alloy 825. Radioactive waste will include spent-fuel assemblies from reactors as well as waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The waste-package containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, the containers must be retrievable from the disposal site. Shortly after emplacement of the containers in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. This radiation will promote the radiolytic decomposition of moist air to hydrogen. This volume surveys the available data on the effects of hydrogen on the six candidate alloys for fabrication of the containers. For copper, the mechanism of hydrogen embrittlement is discussed, and the effects of hydrogen on the mechanical properties of the copper-based alloys are reviewed. The solubilities and diffusivities of hydrogen are documented for these alloys. For the austenitic materials, the degradation of mechanical properties by hydrogen is documented. The diffusivity and solubility of hydrogen in these alloys are also presented. For the copper-based alloys, the ranking according to resistance to detrimental effects of hydrogen is: CDA 715 (best) > CDA 613 > CDA 102 (worst). For the austenitic alloys, the ranking is: Type 316L stainless steel {approx} Alloy 825 > Type 304L stainless steel (worst). 87 refs., 19 figs., 8 tabs.

  7. Compliance with Waste Acceptance Criteria of WIPP and NTS for Vitrified Low-Level and TRU Waste Forms

    SciTech Connect (OSTI)

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

    1998-07-01

    A joint project between the Oak Ridge National Laboratory (ORNL) and the Savannah River Technology Center (SRTC) has been established to evaluate vitrification as an option for the immobilization of waste within ORNL tank farms. This paper presents details of calculations based on current best available analyses of the Oak Ridge Tanks on the limits for waste loadings imposed by the waste acceptance criteria.

  8. Comparison of selected foreign plans and practices for spent fuel and high-level waste management

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.; Lakey, L.T.; Johnson, A.B. Jr.; Hazelton, R.F.; Bradley, D.J.

    1990-04-01

    This report describes the major parameters for management of spent nuclear fuel and high-level radioactive wastes in selected foreign countries as of December 1989 and compares them with those in the United States. The foreign countries included in this study are Belgium, Canada, France, the Federal Republic of Germany, Japan, Sweden, Switzerland, and the United Kingdom. All the countries are planning for disposal of spent fuel and/or high-level wastes in deep geologic repositories. Most countries (except Canada and Sweden) plan to reprocess their spent fuel and vitrify the resultant high-level liquid wastes; in comparison, the US plans direct disposal of spent fuel. The US is planning to use a container for spent fuel as the primary engineered barrier. The US has the most developed repository concept and has one of the earliest scheduled repository startup dates. The repository environment presently being considered in the US is unique, being located in tuff above the water table. The US also has the most prescriptive regulations and performance requirements for the repository system and its components. 135 refs., 8 tabs.

  9. Framework for DOE mixed low-level waste disposal: Site fact sheets

    SciTech Connect (OSTI)

    Gruebel, M.M.; Waters, R.D.; Hospelhorn, M.B.; Chu, M.S.Y.

    1994-11-01

    The Department of Energy (DOE) is required to prepare and submit Site Treatment Plans (STPS) pursuant to the Federal Facility Compliance Act (FFCAct). Although the FFCAct does not require that disposal be addressed in the STPS, the DOE and the States recognize that treatment of mixed low-level waste will result in residues that will require disposal in either low-level waste or mixed low-level waste disposal facilities. As a result, the DOE is working with the States to define and develop a process for evaluating disposal-site suitability in concert with the FFCAct and development of the STPS. Forty-nine potential disposal sites were screened; preliminary screening criteria reduced the number of sites for consideration to twenty-six. The DOE then prepared fact sheets for the remaining sites. These fact sheets provided additional site-specific information for understanding the strengths and weaknesses of the twenty-six sites as potential disposal sites. The information also provided the basis for discussion among affected States and the DOE in recommending sites for more detailed evaluation.

  10. Siting process for disposal site of low level radiactive waste in Thailand

    SciTech Connect (OSTI)

    Yamkate, P.; Sriyotha, P.; Thiengtrongjit, S.; Sriyotha, K. )

    1992-01-01

    The radioactive waste in Thailand is composed of low level waste from the application of radioisotopes in medical treatment and industry, the operation of the 2 MW TRIGA Mark III Research Reactor and the production of radioisotopes at OAEP. In addition, the high activity of sealed radiation sources i.e. Cs-137 Co-60 and Ra-226 are also accumulated. Since the volume of treated waste has been gradually increased, the general needs for a repository become apparent. The near surface disposal method has been chosen for this aspect. The feasibility study on the underground disposal site has been done since 1982. The site selection criteria have been established, consisting of the rejection criteria, the technical performance criteria and the economic criteria. About 50 locations have been picked for consideration and 5 candidate sites have been selected and subsequent investigated. After thoroughly investigation, a definite location in Ratchburi Province, about 180 kilometers southwest of Bangkok, has been selected as the most suitable place for the near surface disposal of radioactive waste in Thailand.

  11. Complete decay of radionuclides: Implications for low-level waste disposal in municipal landfills

    SciTech Connect (OSTI)

    Meck, R.A.

    1996-05-01

    The time required for the complete decay of a radioactive source can be quantified by specifying an acceptable probability and using an original derivation. The physical phenomenon of complete decay may be used as the technical basis to change regulations and permit, with public acceptance, the inexpensive disposal of short half-lived radioactive waste into municipal landfills. Current regulations require isolation of trash form the biosphere for 30 years during the post-closure control period for municipal landfills. Thirty years is sufficient time for complete decay of significant quantities of short-lived radionuclides, and there is a large decay capacity in the nation`s landfills. As the major generators of low-level radioactive waste with relatively short half-lives, the academic, medical, and research communities likely would benefit most from such regulatory relief. Disposal of such waste is prohibited or costly. The waste must be specially packaged, stored, transported, and disposed in designated repositories. Regulatory relief can be initiated by citizens since the Administrative Procedures Act gives citizens the right to petition for regulatory change. 10 refs., 2 tabs.

  12. Preliminary Hazard Analysis for the Remote-Handled Low-Level Waste Disposal Facility

    SciTech Connect (OSTI)

    Lisa Harvego; Mike Lehto

    2010-02-01

    The need for remote handled low level waste (LLW) disposal capability has been identified. A new onsite, remote-handled LLW disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled LLW disposal capability for remote-handled LLW that is generated as part of the nuclear mission of the Idaho National Laboratory and from spent nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled LLW in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This document supports the conceptual design for the proposed remote-handled LLW disposal facility by providing an initial nuclear facility hazard categorization and by identifying potential hazards for processes associated with onsite handling and disposal of remote-handled LLW.

  13. Preliminary Hazard Analysis for the Remote-Handled Low-Level Waste Disposal Facility

    SciTech Connect (OSTI)

    Lisa Harvego; Mike Lehto

    2010-05-01

    The need for remote handled low level waste (LLW) disposal capability has been identified. A new onsite, remote-handled LLW disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled LLW disposal capability for remote-handled LLW that is generated as part of the nuclear mission of the Idaho National Laboratory and from spent nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled LLW in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This document supports the conceptual design for the proposed remote-handled LLW disposal facility by providing an initial nuclear facility hazard categorization and by identifying potential hazards for processes associated with onsite handling and disposal of remote-handled LLW.

  14. Preliminary Hazard Analysis for the Remote-Handled Low-Level Waste Disposal Project

    SciTech Connect (OSTI)

    Lisa Harvego; Mike Lehto

    2010-10-01

    The need for remote handled low level waste (LLW) disposal capability has been identified. A new onsite, remote-handled LLW disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled LLW disposal capability for remote-handled LLW that is generated as part of the nuclear mission of the Idaho National Laboratory and from spent nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled LLW in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This document supports the conceptual design for the proposed remote-handled LLW disposal facility by providing an initial nuclear facility hazard categorization and by identifying potential hazards for processes associated with onsite handling and disposal of remote-handled LLW.

  15. Optimization of Deep Borehole Systems for HLW Disposal

    SciTech Connect (OSTI)

    Driscoll, Michael; Baglietto, Emilio; Buongiorno, Jacopo; Lester, Richard; Brady, Patrick; Arnold, B. W.

    2015-09-09

    This is the final report on a project to update and improve the conceptual design of deep boreholes for high level nuclear waste disposal. The effort was concentrated on application to intact US legacy LWR fuel assemblies, but conducted in a way in which straightforward extension to other waste forms, host rock types and countries was preserved. The reference fuel design version consists of a vertical borehole drilled into granitic bedrock, with the uppermost kilometer serving as a caprock zone containing a diverse and redundant series of plugs. There follows a one to two kilometer waste canister emplacement zone having a hole diameter of approximately 40-50 cm. Individual holes are spaced 200-300 m apart to form a repository field. The choice of verticality and the use of a graphite based mud as filler between the waste canisters and the borehole wall liner was strongly influenced by the expectation that retrievability would continue to be emphasized in US and worldwide repository regulatory criteria. An advanced version was scoped out using zinc alloy cast in place to fill void space inside a disposal canister and its encapsulated fuel assembly. This excludes water and greatly improves both crush resistance and thermal conductivity. However the simpler option of using a sand fill was found adequate and is recommended for near-term use. Thermal-hydraulic modeling of the low permeability and porosity host rock and its small (≤ 1%) saline water content showed that vertical convection induced by the waste’s decay heat should not transport nuclides from the emplacement zone up to the biosphere atop the caprock. First order economic analysis indicated that borehole repositories should be cost-competitive with shallower mined repositories. It is concluded that proceeding with plans to drill a demonstration borehole to confirm expectations, and to carry out priority experiments, such as retention and replenishment of in-hole water is in order.

  16. Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, task 17208: Final report

    SciTech Connect (OSTI)

    Amoroso, J. W.; Marra, J. C.

    2015-08-26

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

  17. Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208: Final report

    SciTech Connect (OSTI)

    Amoroso, J. W.; Marra, J. C.

    2015-08-26

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

  18. Waste Treatment and Immobilization Plant (WTP) Analytical Laboratory (LAB),

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

    Balance of Facilities (BOF) and Low-Activity Waste Vitrification Facilities (LAW) | Department of Energy Immobilization Plant (WTP) Analytical Laboratory (LAB), Balance of Facilities (BOF) and Low-Activity Waste Vitrification Facilities (LAW) PDF icon Summary - WTP Analytical Lab, BOF and LAW Waste Vitrification Facilities More Documents & Publications Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility Waste Treatment and Immobilation Plant Pretreatment Facility

  19. Proceedings of the tenth annual DOE low-level waste management conference: Session 3: Disposal technology and facility development

    SciTech Connect (OSTI)

    Not Available

    1988-12-01

    This document contains ten papers on various aspects of low-level radioactive waste management. Topics include: design and construction of a facility; alternatives to shallow land burial; the fate of tritium and carbon 14 released to the environment; defense waste management; engineered sorbent barriers; remedial action status report; and the disposal of mixed waste in Texas. Individual papers were processed separately for the data base. (TEM)

  20. Technical Scope and Approach for the 2004 Composite Analysis of Low Level Waste Disposal at the Hanford Site

    SciTech Connect (OSTI)

    Kincaid, Charles T.; Bryce, Robert W.; Buck, John W.

    2004-07-09

    A composite analysis is required by U.S. Department of Energy (DOE) Manual 435.1-1 to ensure public safety through the management of active and planned low-level radioactive waste disposal facilities associated with the Hanford Site (DOE/HQ-Manual 435.1-1). A Composite Analysis is defined as ''a reasonably conservative assessment of the cumulative impact from active and planned low-level waste disposal facilities, and all other sources from radioactive contamination that could interact with the low-level waste disposal facility to affect the dose to future members of the public''. At the Hanford Site, a composite analysis is required for continued disposal authorization for the immobilized low-activity waste, tank waste vitrification plant melters, low level waste in the 200 East and 200 West Solid Waste Burial Grounds, and Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) waste in the Environmental Restoration Disposal Facility. The 2004 Composite Analysis will be a site-wide analysis, considering final remedial actions for the Columbia River corridor and the Central Plateau at the Hanford Site. The river corridor includes waste sites and facilities in each of the 100 Areas as well as the 300, 400, and 600 Areas. The remedial actions for the river corridor are being conducted to meet residential land use standards with the vision of the river corridor being devoted to a combination of recreation and preservation. The ''Central Plateau'' describes the region associated with operations and waste sites of the 200 Areas. DOE is developing a strategy for closure of the Central Plateau area by 2035. At the time of closure, waste management activities will shrink to a Core Zone within the Central Plateau. The Core Zone will contain the majority of Hanford's permanently disposed waste

  1. DESCISION SUPPORT SYSTEM FOR MANAGEMENT OF LOW-LEVEL WASTE DISPOSAL AT THE NEVADA TEST SITE

    National Nuclear Security Administration (NNSA)

    Decision Support System For Management Of Low-Level Radioactive Waste Disposal At The Nevada Test Site G. Shott, V. Yucel, L. Desotell Bechtel Nevada P.O. Box 98521, Las Vegas, NV 89193-8521 USA J.T. Carilli U.S. Department of Energy National Nuclear Security Administration, Nevada Site Office P.O. Box 98518, Las Vegas, NV 89193-8518 USA ABSTRACT The long-term safety of U.S. Department of Energy (DOE) low-level radioactive disposal facilities is assessed by conducting a performance assessment --

  2. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM 2009

    SciTech Connect (OSTI)

    West, B.; Waltz, R.

    2010-06-21

    Aqueous radioactive wastes from Savannah River Site (SRS) separations and vitrification processes are contained in large underground carbon steel tanks. Inspections made during 2009 to evaluate these vessels and other waste handling facilities along with evaluations based on data from previous inspections are the subject of this report. The 2009 inspection program revealed that the structural integrity and waste confinement capability of the Savannah River Site waste tanks were maintained. All inspections scheduled per LWO-LWE-2008-00423, HLW Tank Farm Inspection Plan for 2009, were completed. All Ultrasonic measurements (UT) performed in 2009 met the requirements of C-ESG-00006, In-Service Inspection Program for High Level Waste Tanks, Rev. 1, and WSRC-TR-2002-00061, Rev.4. UT inspections were performed on Tank 29 and the findings are documented in SRNL-STI-2009-00559, Tank Inspection NDE Results for Fiscal Year 2009, Waste Tank 29. Post chemical cleaning UT measurements were made in Tank 6 and the results are documented in SRNL-STI-2009-00560, Tank Inspection NDE Results Tank 6, Including Summary of Waste Removal Support Activities in Tanks 5 and 6. A total of 6669 photographs were made and 1276 visual and video inspections were performed during 2009. Twenty-Two new leaksites were identified in 2009. The locations of these leaksites are documented in C-ESR-G-00003, SRS High Level Waste Tank Leaksite Information, Rev.4. Fifteen leaksites at Tank 5 were documented during tank wall/annulus cleaning activities. Five leaksites at Tank 6 were documented during tank wall/annulus cleaning activities. Two new leaksites were identified at Tank 19 during waste removal activities. Previously documented leaksites were reactivated at Tanks 5 and 12 during waste removal activities. Also, a very small amount of additional leakage from a previously identified leaksite at Tank 14 was observed.

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

    SciTech Connect (OSTI)

    Hulse, R.A.

    1991-08-01

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

  4. Vitrification treatment options for disposal of greater-than-Class-C low-level waste in a deep geologic repository

    SciTech Connect (OSTI)

    Fullmer, K.S.; Fish, L.W.; Fischer, D.K.

    1994-11-01

    The Department of Energy (DOE), in keeping with their responsibility under Public Law 99-240, the Low-Level Radioactive Waste Policy Amendments Act of 1985, is investigating several disposal options for greater-than-Class C low-level waste (GTCC LLW), including emplacement in a deep geologic repository. At the present time vitrification, namely borosilicate glass, is the standard waste form assumed for high-level waste accepted into the Civilian Radioactive Waste Management System. This report supports DOE`s investigation of the deep geologic disposal option by comparing the vitrification treatments that are able to convert those GTCC LLWs that are inherently migratory into stable waste forms acceptable for disposal in a deep geologic repository. Eight vitrification treatments that utilize glass, glass ceramic, or basalt waste form matrices are identified. Six of these are discussed in detail, stating the advantages and limitations of each relative to their ability to immobilize GTCC LLW. The report concludes that the waste form most likely to provide the best composite of performance characteristics for GTCC process waste is Iron Enriched Basalt 4 (IEB4).

  5. Retention of Halogens in Waste Glass

    SciTech Connect (OSTI)

    Hrma, Pavel R.

    2010-05-01

    In spite of their potential roles as melting rate accelerators and foam breakers, halogens are generally viewed as troublesome components for glass processing. Of five halogens, F, Cl, Br, I, and At, all but At may occur in nuclear waste. A nuclear waste feed may contain up to 10 g of F, 4 g of Cl, and ?100 mg of Br and I per kg of glass. The main concern is halogen volatility, producing hazardous fumes and particulates, and the radioactive iodine 129 isotope of 1.7x10^7-year half life. Because F and Cl are soluble in oxide glasses and tend to precipitate on cooling, they can be retained in the waste glass in the form of dissolved constituents or as dispersed crystalline inclusions. This report compiles known halogen-retention data in both high-level waste (HLW) and low-activity waste (LAW) glasses. Because of its radioactivity, the main focus is on I. Available data on F and Cl were compiled for comparison. Though Br is present in nuclear wastes, it is usually ignored; no data on Br retention were found.

  6. Low Level Radioactive Wastes Conditioning during Decommissioning of Salaspils Research Reactor

    SciTech Connect (OSTI)

    Abramenkova, G.; Klavins, M.; Abramenkovs, A.

    2008-01-15

    The decommissioning of Salaspils research reactor is connected with the treatment of 2200 tons different materials. The largest part of all materials ({approx}60 % of all dismantled materials) is connected with low level radioactive wastes conditioning activities. Dismantled radioactive materials were cemented in concrete containers using water-cement mortar. According to elaborated technology, the tritiated water (150 tons of liquid wastes from special canalization tanks) was used for preparation of water-cement mortar. Such approach excludes the emissions of tritiated water into environment and increases the efficiency of radioactive wastes management system for decommissioning of Salaspils research reactor. The Environmental Impact Assessment studies for Salaspils research reactor decommissioning (2004) and for upgrade of repository 'Radons' for decommissioning purposes (2005) induced the investigations of radionuclides release parameters from cemented radioactive waste packages. These data were necessary for implementation of quality assurance demands during conditioning of radioactive wastes and for safety assessment modeling for institutional control period during 300 years. Experimental studies indicated, that during solidification of water- cement samples proceeds the increase of temperature up to 81 deg. C. It is unpleasant phenomena since it can result in damage of concrete container due to expansion differences for mortar and concrete walls. Another unpleasant factor is connected with the formation of bubbles and cavities in the mortar structure which can reduce the mechanical stability of samples and increase the release of radionuclides from solidified cement matrix. The several additives, fly ash and PENETRON were used for decrease of solidification temperature. It was found, that addition of fly ash to the cement-water mortar can reduce the solidification temperature up to 62 deg. C. Addition of PENETRON results in increasing of solidification temperature up to 83 deg. C. Experimental data shows, that water/cement ratio significantly influences on water-cement mortar's viscosity and solidified samples mechanical stability. Increasing of water ratio from 0.45 up to 0.65 decreases water-cement mortar's viscosity from 1100 mPas up to 90 mPas. Significant reduction of viscosity is an important factor, which facilitates the fulfillment all gaps and cavities with the mortar during conditioning of solid radioactive wastes in containers. On the other hand, increase water ratio from 0.45 up to 0.65 decreases mechanical stability of water-cement samples from 23 N/mm{sup 2} to the 12 N/mm{sup 2}. It means that water-cement bulk stability significantly decreases with increasing of water content. Technologically is important to increase the tritiated water content in container with cemented radioactive wastes. It gives a possibility to increase the fulfillment of container with radioactive materials. On the other hand, additional water significantly reduces bulk stability of containers with cemented radioactive wastes, which can result in disintegration of radioactive wastes packages in repository during 300 years. Taking into account the experimental results, it is not recommended to exceed the water/cement ratio more than 0.60. Tritium and Cs{sup 137} leakage tests show, that radionuclides release curves has a complicate structure. Experimental results indicated that addition of fly ash result in facilitation of tritium and cesium release in water phase. This is unpleasant factor, which significantly decreases the safety of disposed radioactive wastes. Despite the positive impact on solidification temperature drop, the addition of fly ash to the cement-water mortar is not recommended in case of cementation of radionuclides in concrete containers. In conclusion: The cementation processes of solid radioactive wastes in concrete containers were investigated. The influence of additives on cementation processes was studied. It was shown, that the increasing of water ratio from 0.45 up to 0.65 decreases water-cement mortar

  7. Melton Valley liquid low-level radioactive waste storage tanks evaluation

    SciTech Connect (OSTI)

    1995-06-01

    The Melton Valley Liquid Low-Level Radioactive Waste Storage Tanks (MVSTs) store the evaporator concentrates from the Liquid Low-Level Radioactive Waste (LLLW) System at the Oak Ridge National Laboratory (ORNL). The eight stainless steel tanks contain approximately 375,000 gallons of liquid and sludge waste. These are some of the newer, better-designed tanks in the LLLW System. They have been evaluated and found by the US Environmental Protection Agency (EPA) and the Tennessee Department of Environment and Conservation to comply with all Federal Facility Agreement requirements for double containment. The operations and maintenance aspects of the tanks were also reviewed by the Defense Nuclear Facilities Safety Board (DNFSB) in September 1994. This document also contains an assessment of the risk to the public and ORNL workers from a leak in one of the MVSTs. Two primary scenarios were investigated: (1) exposure of the public to radiation from drinking Clinch River water contaminated by leaked LLLW, and (2) exposure of on-site workers to radiation by inhaling air contaminated by leaked LLLW. The estimated frequency of a leak from one of the MVSTs is about 8 {times} 10{sup {minus}4} events per year, or about once in 1200 years (with a 95% confidence level). If a leak were to occur, the dose to a worker from inhalation would be about 2.3 {times} 10{sup {minus}1} mrem (with a 95% confidence level). The dose to a member of the public through the drinking water pathway is estimated to be about 7 {times} 10{sup {minus}1} mrem (with a 95% confidence level). By comparison with EPA Safe Drinking Water regulations, the allowable lifetime radiation dose is about 300 mrem. Thus, a postulated LLLW leak from the MVSTs would not add appreciably to an individual`s lifetime radiation dose.

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

    SciTech Connect (OSTI)

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

    2011-08-01

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

  9. Iron-phosphate ceramics for solidification of mixed low-level waste

    DOE Patents [OSTI]

    Aloy, Albert S. (St. Petersburg, RU); Kovarskaya, Elena N. (St. Petersburg, RU); Koltsova, Tatiana I. (St. Petersburg, RU); Macheret, Yevgeny (Idaho Falls, ID); Medvedev, Pavel G. (Ozersk, RU); Todd, Terry (Aberdeen, ID)

    2000-01-01

    A method of immobilizing mixed low-level waste is provided which uses low cost materials and has a relatively long hardening period. The method includes: forming a mixture of iron oxide powders having ratios, in mass %, of FeO:Fe.sub.2 O.sub.3 :Fe.sub.3 O.sub.4 equal to 25-40:40-10:35-50, or weighing a definite amount of magnetite powder. Metallurgical cinder can also be used as the source of iron oxides. A solution of the orthophosphoric acid, or a solution of the orthophosphoric acid and ferric oxide, is formed and a powder phase of low-level waste and the mixture of iron oxide powders or cinder (or magnetite powder) is also formed. The acid solution is mixed with the powder phase to form a slurry with the ratio of components (mass %) of waste:iron oxide powders or magnetite:acid solution=30-60:15-10:55-30. The slurry is blended to form a homogeneous mixture which is cured at room temperature to form the final product.

  10. Preliminary PCT data on glass formulations developed for Hanford Site low-level wastes

    SciTech Connect (OSTI)

    Feng, X.; Schweiger, M.J.; Hrma, P.R.; Palmer, S.E.; Smith, D.E.; Kim, D.; Gong, M.; Westsik, J.H. Jr.

    1995-09-01

    Tank wastes stored at the Hanford Site are to be separated into high-level and low-level waste (LLW) fractions and vitrified for disposal. The high content of Na{sub 2}O in the LLW, averaging about 80 wt% on an oxide basis, necessitates the development of durable high-sodium glasses. Pacific Northwest Laboratory (PNL) is providing glass formulations for the LLW vitrification program. The most important considerations for acceptable LLW waste glass compositions are the following: (1) the capability to incorporate high sodium content from LLW; (2) satisfactory long-term durability; and (3) proper processability, such as desired viscosity at melting temperature. To develop durable high-sodium glasses in a reasonably short time and to supply data needed for modeling the glass`s long-term performance, several short-term test methods, such as 7-day PCT (Product Consistency Test), 28-day Materials Characterization Center MCC-1 test, single-pass flow through test, and vapor-hydration test have been used. A long-term static test (up to one year using PCT) is also being performed for selected glasses. This data report includes only the PCT data available at the time of the publication.

  11. Assessment of microbial processes on gas production at radioactive low-level waste disposal sites

    SciTech Connect (OSTI)

    Weiss, A.J.; Tate, R.L. III; Colombo, P.

    1982-05-01

    Factors controlling gaseous emanations from low level radioactive waste disposal sites are assessed. Importance of gaseous fluxes of methane, carbon dioxide, and possible hydrogen from the site, stems from the inclusion of tritium and/or carbon-14 into the elemental composition of these compounds. In that the primary source of these gases is the biodegradation of organic components of the waste material, primary emphasis of the study involved an examination of the biochemical pathways producing methane, carbon dioxide, and hydrogen, and the environmental parameters controlling the activity of the microbial community involved. Initial examination of the data indicates that the ecosystem is anaerobic. As the result of the complexity of the pathway leading to methane production, factors such as substrate availability, which limit the initial reaction in the sequence, greatly affect the overall rate of methane evolution. Biochemical transformations of methane, hydrogen and carbon dioxide as they pass through the soil profile above the trench are discussed. Results of gas studies performed at three commercial low level radioactive waste disposal sites are reviewed. Methods used to obtain trench and soil gas samples are discussed. Estimates of rates of gas production and amounts released into the atmosphere (by the GASFLOW model) are evaluated. Tritium and carbon-14 gaseous compounds have been measured in these studies; tritiated methane is the major radionuclide species in all disposal trenches studied. The concentration of methane in a typical trench increases with the age of the trench, whereas the concentration of carbon dioxide is similar in all trenches.

  12. Performance assessment for a hypothetical low-level waste disposal facility

    SciTech Connect (OSTI)

    Smith, C.S.; Rohe, M.J.; Ritter, P.D.

    1997-01-01

    Disposing of low-level waste (LLW) is a concern for many states throughout the United States. A common disposal method is below-grade concrete vaults. Performance assessment analyses make predictions of contaminant release, transport, ingestion, inhalation, or other routes of exposure, and the resulting doses for various disposal methods such as the below-grade concrete vaults. Numerous assumptions are required to simplify the processes associated with the disposal facility to make predictions feasible. In general, these assumptions are made conservatively so as to underestimate the performance of the facility. The objective of this report is to describe the methodology used in conducting a performance assessment for a hypothetical waste facility located in the northeastern United States using real data as much as possible. This report consists of the following: (a) a description of the disposal facility and site, (b) methods used to analyze performance of the facility, (c) the results of the analysis, and (d) the conclusions of this study.

  13. Commercial disposal options for Idaho National Engineering Laboratory low-level radioactive waste

    SciTech Connect (OSTI)

    Porter, C.L.; Widmayer, D.A.

    1995-09-01

    The Idaho National Engineering Laboratory (INEL) is a Department of Energy (DOE)-owned, contractor-operated site. Significant quantities of low-level radioactive waste (LLW) have been generated and disposed of onsite at the Radioactive Waste Management Complex (RWMC). The INEL expects to continue generating LLW while performing its mission and as aging facilities are decommissioned. An on-going Performance Assessment process for the RWMC underscores the potential for reduced or limited LLW disposal capacity at the existing onsite facility. In order to properly manage the anticipated amount of LLW, the INEL is investigating various disposal options. These options include building a new facility, disposing the LLW at other DOE sites, using commercial disposal facilities, or seeking a combination of options. This evaluation reports on the feasibility of using commercial disposal facilities.

  14. Overview of Low-Level Waste Disposal Operations at the Nevada Test Site

    SciTech Connect (OSTI)

    DOE /Navarro

    2007-02-01

    The U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office Environmental Management Program is charged with the responsibility to carry out the disposal of on-site and off-site generated low-level radioactive waste at the Nevada Test Site. Core elements of this mission are ensuring that disposal take place in a manner that is safe and cost-effective while protecting workers, the public, and the environment. This paper focuses on giving an overview of the Nevada Test Site facilities regarding currant design of disposal. In addition, technical attributes of the facilities established through the site characterization process will be further described. An update on current waste disposal volumes and capabilities will also be provided. This discussion leads to anticipated volume projections and disposal site requirements as the Nevada Test Site disposal operations look towards the future.

  15. Low-Level waste phase 1 melter testing off gas and mass balance evaluation

    SciTech Connect (OSTI)

    Wilson, C.N.

    1996-06-28

    Commercially available melter technologies were tested during 1994-95 as part of a multiphase program to test candidate technologies for vitrification of the low-level waste (LLW) stream to be derived from retrieval and pretreatment of Hanford Site tank wastes. Seven vendors were selected for Phase 1 testing to demonstrate vitrification of a high sodium content liquid LLW simulant. The tested melter technologies included four Joule-heated melters, a carbon electrode melter, a combustion melter, and a plasma melter. Various dry and slurry melter feed preparation processes were also tested. Various feed material samples, product glass samples, and process offgas streams were characterized to provide data for evaluation of process decontamination factors and material mass balances for each vitrification technology. This report describes the melter mass balance evaluations and results for six of the Phase 1 LLW melter vendor demonstration tests.

  16. Low-level waste vitrification phase 1 vendor test sample analysis data

    SciTech Connect (OSTI)

    Mast, E.S.

    1995-10-04

    A multi-phase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Phase 1 of the melter demonstration tests was performed in vendor test facilities using simulated LLW and was completed during FY-1995. Test samples taken during Phase 1 testing were analyzed by independent laboratories who reported the analyses results to Westinghouse Hanford Company for integration and evaluation. The reported analytical results were integrated into an electronic data base using Microsoft Excel*5.0. This report documents this data base as of the end of FY-1995, and is supplemental to the Phase 1 LLW melter testing summary report, WHC-SD-WM-ER-498, revision 0.

  17. Cold Crucible Induction Melting Technology for Vitrification of High Level Waste: Development and Status in India

    SciTech Connect (OSTI)

    Sugilal, G.; Sengar, P.B.S. [Nuclear Recycle Group, Bhabha Atomic Research Centre, Trombay, Mumbai (India)

    2008-07-01

    Cold crucible induction melting is globally emerging as an alternative technology for the vitrification of high level radioactive waste. The new technology offers several advantages such as high temperature availability with long melter life, high waste loading, high specific capacity etc. Based on the laboratory and bench scale studies, an engineering scale cold crucible induction melter was locally developed in India. The melter was operated continuously to assess its performance. The electrical and thermal efficiencies were found to be in the range of 70-80 % and 10-20 % respectively. Glass melting capacities up to 200 kg m{sup -2} hr{sup -1} were accomplished using the ESCCIM. Industrially adaptable melter operating procedures for start-up, melting and pouring operations were established (author)

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

    SciTech Connect (OSTI)

    Eaton, W.C.

    1995-10-26

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

  19. Proceedings of the tenth annual DOE low-level waste management conference: Session 1: Institutional and regulatory issues

    SciTech Connect (OSTI)

    Not Available

    1988-12-01

    This document contains eleven papers on various aspects of low-level radioactive waste regulation. Topics include: EPA environmental standards; international exemption principles; the concept of below regulatory concern; envirocare activities in Utah; mixed waste; FUSRAP and the Superfund; and a review of various incentive programs. Individual papers are processed separately for the data base. (TEM)

  20. HLW Return from France to Germany - 15 Years of Experience in Public Acceptance and Technical Aspects - 12149

    SciTech Connect (OSTI)

    Graf, Wilhelm

    2012-07-01

    Since in 1984 the national reprocessing concept was abandoned the reprocessing abroad was the only existing disposal route until 1994. With the amendment of the Atomic Energy Act in 2001 spent fuel management changed completely since from 1 June 2005 any delivery of spent fuel to reprocessing plants was prohibited and the direct disposal of spent fuel became mandatory. Until 2005 the total amount of spent fuel to be reprocessed abroad added up to 6080 t HM, 5309 t HM thereof in France. The waste generated from reprocessing - alternatively an equivalent amount of radioactive material - has to be returned to the country of origin according to the commercial contracts signed between the German utilities and COGEMA, now AREVA NC, in France and BNFL, now INS in UK. In addition the German and the French government exchanged notes with the obligation of both sides to enable and support the return of reprocessing residues or equivalents to Germany. The return of high active vitrified waste from La Hague to the interim storage facility at Gorleben was demanding from the technical view i. e. the cask design and the transport. Unfortunately the Gorleben area served as a target for nuclear opponents from the first transport in 1996 to the latest one in 2011. The protection against sabotage of the railway lines and mass protests needed highly improved security measures. In France and Germany special working forces and projects have been set up to cope with this extraordinary situation. A complex transport organization was established to involve all parties in line with the German and French requirements during transport. The last transport of vitrified residues from France has been completed successfully so far thus confirming the efficiency of the applied measures. Over 15 years there was and still is worldwide no comparable situation it is still unique. Summing up, the exceptional project handling challenge that resulted from the continuous anti-nuclear civil disobedience in Germany over the whole 15-year long project running time could be faced efficiently. It has to be concluded that despite of all problems the anti-nuclear activities have caused so far, all transports of vitrified HLW have always been completed successfully by adapting the commonly established safety, security and public acceptance measures to the special conditions and needs in Germany and coordinating the activities of all parties involved but at the expense of high costs for industry and government and a challenging operational complexity. Apart from an anticipatory project planning a good communication between all involved industrial parties and the French and the German government was the key to the effective management of such shipments and to minimize the radiological, economic, environmental, public and political impact. The future will show how efficiently the gained experience can be used for further return projects which are to be realized since no reprocessed waste has yet been returned from UK and neither the medium-level nor the low-level radioactive waste has been transferred from France to Germany. (author)