Sample records for hlw high-level waste

  1. PAIRWISE BLENDING OF HIGH LEVEL WASTE (HLW)

    SciTech Connect (OSTI)

    CERTA, P.J.

    2006-02-22T23:59:59.000Z

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

  2. High level waste characterization in support of low level waste certification. I. HLW supernate radionuclide characterization

    SciTech Connect (OSTI)

    Jamison, M.E.; d`Entremont, P.D.; Clemmons, J.S.; Bess, C.E.; Brown, D.F.

    1994-07-08T23:59:59.000Z

    High Level Waste Programs has radioactive waste storage, treatment and processing facilities that are located in the F and H Areas at the Savannah River Site. These facilities include the Effluent Treatment Facility (ETF), F and H Area Tank Farms, Extended Sludge Processing (ESP), and In-Tank Precipitation (ITP). Job wastes are generated from operation, maintenance, and construction activities inside radiological areas. These items may have been contaminated with radioactive supernate, salt, and sludge material. Most of these wastes will be disposed of in the E-area Vaults. Therefore, an isotopic and hazardous characterization must be performed. The characterization of HLW supernate radionuclides is discussed in Chapter I. The characterization for salt and sludge phases, which can also contaminate LLW, will be included in other Chapters.

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

    SciTech Connect (OSTI)

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

    2010-01-04T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Aurah, Mirwaise Y.; Roberts, Mark A.

    2013-12-12T23:59:59.000Z

    Washington River Protection Solutions (WRPS), operator of High Level Radioactive Waste (HLW) Tank Farms at the Hanford Site, is taking an over 20-year leap in technology, replacing systems that were monitored with clipboards and obsolete computer systems, as well as solving major operations and maintenance hurdles in the area of process automation and information management. While WRPS is fully compliant with procedures and regulations, the current systems are not integrated and do not share data efficiently, hampering how information is obtained and managed.

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

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

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  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

    2010-01-04T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Fox, K. M.

    2014-02-27T23:59:59.000Z

    processing strategy for the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The basis of this alternative approach is an empirical model predicting the crystal accumulation in the WTP glass discharge riser and melter bottom as a function of glass composition, time, and temperature. When coupled with an associated operating limit (e.g., the maximum tolerable thickness of an accumulated layer of crystals), this model could then be integrated into the process control algorithms to formulate crystal tolerant high level waste (HLW) glasses targeting higher waste loadings while still meeting process related limits and melter lifetime expectancies. This report provides a review of the scaled melter testing that was completed in support of the Defense Waste Processing Facility (DWPF) melter. Testing with scaled melters provided the data to define the DWPF operating limits to avoid bulk (volume) crystallization in the un-agitated DWPF melter and provided the data to distinguish between spinels generated by K-3 refractory corrosion versus spinels that precipitated from the HLW glass melt pool. This report includes a review of the crystallization observed with the scaled melters and the full scale DWPF melters (DWPF Melter 1 and DWPF Melter 2). Examples of actual DWPF melter attainment with Melter 2 are given. The intent is to provide an overview of lessons learned, including some example data, that can be used to advance the development and implementation of an empirical model and operating limit for crystal accumulation for WTP. Operation of the first and second (current) DWPF melters has demonstrated that the strategy of using a liquidus temperature predictive model combined with a 100 °C offset from the normal melter operating temperature of 1150 °C (i.e., the predicted liquidus temperature (TL) of the glass must be 1050 °C or less) has been successful in preventing any detrimental accumulation of spinel in the DWPF melt pool, and spinel has not been observed in any of the pour stream glass samples. Spinel was observed at the bottom of DWPF Melter 1 as a result of K-3 refractory corrosion. Issues have occurred with accumulation of spinel in the pour spout during periods of operation at higher waste loadings. Given that both DWPF melters were or have been in operation for greater than 8 years, the service life of the melters has far exceeded design expectations. It is possible that the DWPF liquidus temperature approach is conservative, in that it may be possible to successfully operate the melter with a small degree of allowable crystallization in the glass. This could be a viable approach to increasing waste loading in the glass assuming that the crystals are suspended in the melt and swept out through the riser and pour spout. Additional study is needed, and development work for WTP might be leveraged to support a different operating limit for the DWPF. Several recommendations are made regarding considerations that need to be included as part of the WTP crystal tolerant strategy based on the DWPF development work and operational data reviewed here. These include: Identify and consider the impacts of potential heat sinks in the WTP melter and glass pouring system; Consider the contributions of refractory corrosion products, which may serve to nucleate additional crystals leading to further accumulation; Consider volatilization of components from the melt (e.g., boron, alkali, halides, etc.) and determine their impacts on glass crystallization behavior; Evaluate the impacts of glass REDuction/OXidation (REDOX) conditions and the distribution of temperature within the WTP melt pool and melter pour chamber on crystal accumulation rate; Consider the impact of precipitated crystals on glass viscosity; Consider the impact of an accumulated crystalline layer on thermal convection currents and bubbler effectiveness within the melt pool; Evaluate the impact of spinel accumulation on Joule heating of the WTP melt pool; and Include noble metals in glass melt experiments because of their potential to act as nucleation site

  8. HLW-OVP-97-0068 High Level Waste Management Division High-Level Waste System Plan

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

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

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

    (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 longer...

  10. Interactions of simulated high level waste (HLW) calcine with alkali borosilicate glass. S. Morgan, R. J. Hand, N. C. Hyatt and W. E. Lee

    E-Print Network [OSTI]

    Sheffield, University of

    Interactions of simulated high level waste (HLW) calcine with alkali borosilicate glass. S. Morgan, Mappin St, Sheffield, S1 3JD, UK ABSTRACT This study looks at the interactions between simulated calcined and a solution of lithium nitrate. The addition of sucrose reacts with the free nitric acid in the system

  11. Hanford high level waste (HLW) tank mixer pump safe operating envelope reliability assessment

    SciTech Connect (OSTI)

    Fischer, S.R. [Los Alamos National Lab., NM (United States); Clark, J. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)

    1993-10-01T23:59:59.000Z

    The US Department of Energy and its contractor, Westinghouse Corp., are responsible for the management and safe storage of waste accumulated from processing defense reactor irradiated fuels for plutonium recovery at the Hanford Site. These wastes, which consist of liquids and precipitated solids, are stored in underground storage tanks pending final disposition. Currently, 23 waste tanks have been placed on a safety watch list because of their potential for generating, storing, and periodically releasing various quantities of hydrogen and other gases. Tank 101-SY in the Hanford SY Tank Farm has been found to release hydrogen concentrations greater than the lower flammable limit (LFL) during periodic gas release events. In the unlikely event that an ignition source is present during a hydrogen release, a hydrogen burn could occur with a potential to release nuclear waste materials. To mitigate the periodic gas releases occurring from Tank 101-SY, a large mixer pump currently is being installed in the tank to promote a sustained release of hydrogen gas to the tank dome space. An extensive safety analysis (SA) effort was undertaken and documented to ensure the safe operation of the mixer pump after it is installed in Tank 101-SY.1 The SA identified a need for detailed operating, alarm, and abort limits to ensure that analyzed safety limits were not exceeded during pump operations.

  12. HIGH LEVEL WASTE (HLW) VITRIFICATION EXPERIENCE IN THE US: APPLICATION OF GLASS PRODUCT/PROCESS CONTROL TO OTHERHLW AND HAZARDOUS WASTES

    SciTech Connect (OSTI)

    Jantzen, C; James Marra, J

    2007-09-17T23:59:59.000Z

    Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. At the Savannah River Site (SRS) actual HLW tank waste has successfully been processed to stringent product and process constraints without any rework into a stable borosilicate glass waste since 1996. A unique 'feed forward' statistical process control (SPC) has been used rather than statistical quality control (SQC). In SPC, the feed composition to the melter is controlled prior to vitrification. In SQC, the glass product is sampled after it is vitrified. Individual glass property models form the basis for the 'feed forward' SPC. The property models transform constraints on the melt and glass properties into constraints on the feed composition. The property models are mechanistic and depend on glass bonding/structure, thermodynamics, quasicrystalline melt species, and/or electron transfers. The mechanistic models have been validated over composition regions well outside of the regions for which they were developed because they are mechanistic. Mechanistic models allow accurate extension to radioactive and hazardous waste melts well outside the composition boundaries for which they were developed.

  13. High Level Waste System Plan Revision 9

    SciTech Connect (OSTI)

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

    1998-04-01T23:59:59.000Z

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

  14. Optimizing High Level Waste Disposal

    SciTech Connect (OSTI)

    Dirk Gombert

    2005-09-01T23:59:59.000Z

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

  15. Effect of composition and temperature on the properties of High-Level Waste (HLW) glasses melting above 1200{degrees}C (Draft)

    SciTech Connect (OSTI)

    Vienna, J.D.; Hrma, P.R.; Schweiger, M.J. [and others

    1996-02-01T23:59:59.000Z

    Increasing the melting temperature of HLW glass allows an increase of waste loading (thus reducing product volume) and the production of more durable glasses at a faster melting rate. However, HLW glasses that melt at high temperatures differ in composition from glasses formulated for low temperature ({approximately}1150{degree}C). Consequently, the composition of high-temperature glasses falls in a region previously not well tested or understood. This report represents a preliminary study of property/composition relationships of high-temperature Hanford HLW glasses using a one-component-at-a-time change approach. A test matrix has been designed to explore a composition region expected for high-temperature high-waste loading HLW glasses to be produced at Hanford. This matrix was designed by varying several key components (SiO{sub 2}, B{sub 2}O{sub 3}, Na{sub 2}O, Li{sub 2}O, Fe{sub 2}O{sub 3}, Al{sub 2}O{sub 3}, ZrO{sub 2}, Bi{sub 2}O{sub 3}, P{sub 2}O{sub 5}, UO{sub 2}, TiO{sub 2}, Cr{sub 2}O{sub 3}, and others) starting from a glass based on a Hanford HLW all-blend waste. Glasses were fabricated and tested for viscosity, glass transition temperature, electrical conductivity, crystallinity, liquidus temperature, and PCT release. The effect of individual components on glass properties was assessed using first- and second- order empirical models. The first-order component effects were compared with those from low-temperature HLW glasses.

  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

    2009-11-05T23:59:59.000Z

    Metal corrosion tests were conducted with four high waste loading non-Fe-limited HLW glass compositions. The results at 1150 C (the WTP nominal melter operating temperature) show corrosion performance for all four glasses that is comparable to that of other typical borosilicate waste glasses, including HLW glass compositions that have been developed for iron-limited WTP streams. Of the four glasses tested, the Bi-limited composition shows the greatest extent of corrosion, which may be related to its higher phosphorus content. Tests at higher suggest that a moderate elevation of the melter operating temperature (up to 1200 C) should not result in any significant increase in Inconel corrosion. However, corrosion rates did increase significantly at yet higher temperatures (1230 C). Very little difference was observed with and without the presence of an electric current density of 6 A/inch{sup 2}, which is the typical upper design limit for Inconel electrodes. The data show a roughly linear relationship between the thickness of the oxide scale on the coupon and the Cr-depletion depth, which is consistent with the chromium depletion providing the material source for scale growth. Analysis of the time dependence of the Cr depletion profiles measured at 1200 C suggests that diffusion of Cr in the Ni-based Inconel alloy controls the depletion depth of Cr inside the alloy. The diffusion coefficient derived from the experimental data agrees within one order of magnitude with the published diffusion coefficient data for Cr in Ni matrices; the difference is likely due to the contribution from faster grain boundary diffusion in the tested Inconel alloy. A simple diffusion model based on these data predicts that Inconel 690 alloy will suffer Cr depletion damage to a depth of about 1 cm over a five year service life at 1200 C in these glasses.

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

    SciTech Connect (OSTI)

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

    2009-02-07T23:59:59.000Z

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

  18. Citizen Contributions to the Closure of High-Level Waste (HLW) Tanks 18 and 19 at the Department of Energy's (DOE) Savannah River Site (SRS) - 13448

    SciTech Connect (OSTI)

    Lawless, W.F. [Paine College, Departments of Math and Psychology, 1235 15th Street, Augusta, GA 30901 (United States)] [Paine College, Departments of Math and Psychology, 1235 15th Street, Augusta, GA 30901 (United States)

    2013-07-01T23:59:59.000Z

    Citizen involvement in DOE's decision-making for the environmental cleanup from DOE's management of its nuclear wastes across the DOE complex has had a positive effect on the cleanup of its SRS site, characterized by an acceleration of cleanup not only for the Transuranic wastes at SRS, but also for DOE's first two closures of HLW tanks, both of which occurred at SRS. The Citizens around SRS had pushed successfully for the closures of Tanks 17 and 20 in 1997, becoming the first closures of HLW tanks under regulatory guidance in the USA. However, since then, HLW tank closures ceased due to a lawsuit, the application of new tank clean-up technology, interagency squabbling between DOE and NRC over tank closure criteria, and finally and almost fatally, from budget pressures. Despite an agreement with its regulators for the closure of Tanks 18 and 19 by the end of calendar year 2012, the outlook in Fall 2011 to close these two tanks had dimmed. It was at this point that the citizens around SRS became reengaged with tank closures, helping DOE to reach its agreed upon milestone. (authors)

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

    E-Print Network [OSTI]

    Ghosh, Suchandra Tina, 1973-

    2004-01-01T23:59:59.000Z

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

  20. High-Level Waste Requirements

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

    1999-07-09T23:59:59.000Z

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

  1. High-level waste melter alternatives assessment report

    SciTech Connect (OSTI)

    Calmus, R.B.

    1995-02-01T23:59:59.000Z

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

  2. 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...

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

    SciTech Connect (OSTI)

    Williams, M

    2008-05-09T23:59:59.000Z

    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.

  4. High-level waste issues and resolutions document

    SciTech Connect (OSTI)

    Not Available

    1994-05-01T23:59:59.000Z

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

  5. DEVELOPMENT OF GLASS MATRICES FOR HLW RADIOACTIVE WASTES

    SciTech Connect (OSTI)

    Jantzen, C.

    2010-03-18T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Sheffield, University of

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

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

    SciTech Connect (OSTI)

    Burgard, K.C.

    1998-04-09T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Burgard, K.C.

    1998-06-02T23:59:59.000Z

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

  9. Enclosure 3 DOE Response to EPA Question Regarding "High-Level Liquid Radioactive Waste"

    E-Print Network [OSTI]

    to date, which is from the definitions in the Nuclear Waste Policy Act: The term "high-level radioactive waste" means-- (A) the highly radioactive material resulting from the reprocessing of spent nuclear fuel of waste streams as from the applicable definition of HLW in the Nuclear Waste Policy Act. 5/11/20051 #12

  10. Feed specification for the double-shell tank/single shell tank waste blend for high-level waste vitrification process and melter testing

    SciTech Connect (OSTI)

    Tracey, E.M.; Merz, M.D.; Patello, G.K.; Wiemers, K.D.

    1996-02-01T23:59:59.000Z

    The High-Level Waste (HLW) Vitrification Program is developing technology for the Department of Energy to immobilize high-level and transuranic waste as glass for permanent disposal. In support of the program, Pacific Northwest Laboratory (PNL) is conducting laboratory-scale melter feed preparation studies and HLW melter testing which require a simulated HLW feed. The simulant HLW feed represents a blend of the waste from 177 single shell and double shell tanks. The waste blend composition is based on normalized track radionuclide components (TRAC), historical tank data, and assumptions on the pretreatment of the waste. The HLW simulant feed specification for the waste blend composition provides direction for the preparation of laboratory-scale and large-scale HLW blend simulant to be used in melter feed preparation studies and melter testing.

  11. Production of a High-Level Waste Glass from Hanford Waste Samples

    SciTech Connect (OSTI)

    Crawford, C.L. [Westinghouse Savannah River Company, AIKEN, SC (United States); Farrara, D.M.; Ha, B.C.; Bibler, N.E.

    1998-09-01T23:59:59.000Z

    The HLW glass was produced from a HLW sludge slurry (Envelope D Waste), eluate waste streams containing high levels of Cs-137 and Tc-99, solids containing both Sr-90 and transuranics (TRU), and glass-forming chemicals. The eluates and Sr-90/TRU solids were obtained from ion-exchange and precipitation pretreatments, respectively, of other Hanford supernate samples (Envelopes A, B and C Waste). The glass was vitrified by mixing the different waste streams with glass-forming chemicals in platinum/gold crucibles and heating the mixture to 1150 degree C. Resulting glass analyses indicated that the HLW glass waste form composition was close to the target composition. The targeted waste loading of Envelope D sludge solids in the HLW glass was 30.7 wt percent, exclusive of Na and Si oxides. Condensate samples from the off-gas condenser and off-gas dry-ice trap indicated that very little of the radionuclides were volatilized during vitrification. Microstructure analysis of the HLW glass using Scanning Electron Microscopy (SEM) and Energy Dispersive X-Ray Analysis (EDAX) showed what appeared to be iron spinel in the HLW glass. Further X-Ray Diffraction (XRD) analysis confirmed the presence of nickel spinel trevorite (NiFe2O4). These crystals did not degrade the leaching characteristics of the glass. The HLW glass waste form passed leach tests that included a standard 90 degree C Product Consistency Test (PCT) and a modified version of the United States Environmental Protection Agency Toxicity Characteristic Leaching Procedure (TCLP).

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

    SciTech Connect (OSTI)

    SK Sundaram; ML Elliott; D Bickford

    1999-11-19T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1995-05-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1999-01-01T23:59:59.000Z

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

  15. High-Level Waste Melter Study Report

    SciTech Connect (OSTI)

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

    2001-07-13T23:59:59.000Z

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

  16. High-level radioactive wastes. Supplement 1

    SciTech Connect (OSTI)

    McLaren, L.H. (ed.)

    1984-09-01T23:59:59.000Z

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

  17. High-Level Waste Tank Cleaning and Field Characterization at the West Valley Demonstration Project

    SciTech Connect (OSTI)

    Drake, J. L.; McMahon, C. L.; Meess, D. C.

    2002-02-26T23:59:59.000Z

    The West Valley Demonstration Project (WVDP) is nearing completion of radioactive high-level waste (HLW) retrieval from its storage tanks and subsequent vitrification of the HLW into borosilicate glass. Currently, 99.5% of the sludge radioactivity has been recovered from the storage tanks and vitrified. Waste recovery of cesium-137 (Cs-137) adsorbed on a zeolite media during waste pretreatment has resulted in 97% of this radioactivity being vitrified. Approximately 84% of the original 1.1 x 1018 becquerels (30 million curies) of radioactivity was efficiently vitrified from July 1996 to June 1998 during Phase I processing. The recovery of the last 16% of the waste has been challenging due to a number of factors, primarily the complex internal structural support system within the main 2.8 million liter (750,000 gallon) HLW tank designated 8D-2. Recovery of this last waste has become exponentially more challenging as less and less HLW is available to mobilize and transfer to the Vitrification Facility. This paper describes the progressively more complex techniques being utilized to remove the final small percentage of radioactivity from the HLW tanks, and the multiple characterization technologies deployed to determine the quantity of Cs-137, strontium-90 (Sr-90), and alpha-transuranic (alpha-TRU) radioactivity remaining in the tanks.

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

    SciTech Connect (OSTI)

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

    1999-05-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2009-10-05T23:59:59.000Z

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

  20. High-level waste qualification: Managing uncertainty

    SciTech Connect (OSTI)

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

    1993-12-31T23:59:59.000Z

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

  1. 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.

    2013-01-09T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2013-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    G. Radulesscu; J.S. Tang

    2000-06-07T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2010-12-17T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    THIEN MG; WELLS BE; ADAMSON DJ

    2010-01-14T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    N /A

    2002-10-11T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Sandra M Birk

    2010-10-01T23:59:59.000Z

    ABSTRACT This paper provides an overview of the Department of Energy's (DOE) spent nuclear fuel (SNF) and high level waste (HLW) storage management. Like commercial reactor fuel, DOE's SNF and HLW were destined for the Yucca Mountain repository. In March 2010, the DOE filed a motion with the Nuclear Regulatory Commission (NRC) to withdraw the license application for the repository at Yucca Mountain. A new repository is now decades away. The default for the commercial and DOE research reactor fuel and HLW is on-site storage for the foreseeable future. Though the motion to withdraw the license application and delay opening of a repository signals extended storage, DOE's immediate plans for management of its SNF and HLW remain the same as before Yucca Mountain was designated as the repository, though it has expanded its research and development efforts to ensure safe extended storage. This paper outlines some of the proposed research that DOE is conducting and will use to enhance its storage systems and facilities.

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

    SciTech Connect (OSTI)

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

    2009-12-30T23:59:59.000Z

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

  9. Savannah River Site High-Level Waste Tank Closure Final Environmental Impact Statement

    SciTech Connect (OSTI)

    N /A

    2002-05-31T23:59:59.000Z

    The U.S. Atomic Energy Commission, a U.S. Department of Energy (DOE) predecessor agency, established the Savannah River Site (SRS) near Aiken, South Carolina, in the early 1950s. The primary mission of SRS was to produce nuclear materials for national defense. With the end of the Cold War and the reduction in the size of the United States stockpile of nuclear weapons, the SRS mission has changed. While national defense is still an important facet of the mission, SRS no longer produces nuclear materials and the mission is focused on material stabilization, environmental restoration, waste management, and decontamination and decommissioning of facilities that are no longer needed. As a result of its nuclear materials production mission, SRS generated large quantities of high-level radioactive waste (HLW). The HLW resulted from dissolving spent reactor fuel and nuclear targets to recover the valuable radioactive isotopes. DOE had stored the HLW in 51 large underground storage tanks located in the F- and H-Area Tank Farms at SRS. DOE has emptied and closed two of those tanks. DOE is treating the HLW, using a process called vitrification. The highly radioactive portion of the waste is mixed with a glass like material and stored in stainless steel canisters at SRS, pending shipment to a geologic repository for disposal. This process is currently underway at SRS in the Defense Waste Processing Facility (DWPF). The HLW tanks at SRS are of four different types, which provide varying degrees of protection to the environment due to different degrees of containment. The tanks are operated under the authority of the Atomic Energy Act of 1954 (AEA) and DOE Orders issued under the AEA. The tanks are permitted by the South Carolina Department of Health and Environmental Control (SCDHEC) under South Carolina wastewater regulations, which require permitted facilities to be closed after they are removed from service. DOE has entered into an agreement with the U.S. Environmental Protection Agency (EPA) and SCDHEC to close the HLW tanks after they have been removed from service. Closure of the HLW tanks would comply with DOE's responsibilities under the AEA and the South Carolina closure requirements and be carried out under a schedule agreed to by DOE, EPA, and SCDHEC. There are several ways to close the HLW tanks. DOE has prepared this Environmental Impact Statement (EIS) to ensure that the public and DOE's decision makers have a thorough understanding of the potential environmental impacts of alternative means of closing the tanks. This Summary: (1) describes the HLW tanks and the closure process, (2) describes the National Environmental Policy Act (NEPA) process that DOE is using to aid in decision making, (3) summarizes the alternatives for closing the HLW tanks and identifies DOE.s preferred alternative, and (4) identifies the major conclusions regarding environmental impacts, areas of controversy, and issues that remain to be resolved as DOE proceeds with the HLW tank closure process.

  10. C-106 High-Level Waste Solids: Washing/Leaching and Solubility Versus Temperature Studies

    SciTech Connect (OSTI)

    GJ Lumetta; DJ Bates; PK Berry; JP Bramson; LP Darnell; OT Farmer III; LR Greenwood; FV Hoopes; RC Lettau; GF Piepel; CZ Soderquist; MJ Steele; RT Steele; MW Urie; JJ Wagner

    2000-01-26T23:59:59.000Z

    This report describes the results of a test conducted by Battelle to assess the effects of inhibited water washing and caustic leaching on the composition of the Hanford tank C-106 high-level waste (HLW) solids. The objective of this work was to determine the composition of the C-106 solids remaining after washing with 0.01M NaOH or leaching with 3M NaOH. Another objective of this test was to determine the solubility of various C-106 components as a function of temperature. The work was conducted according to test plan BNFL-TP-29953-8,Rev. 0, Determination of the Solubility of HLW Sludge Solids. The test went according to plan, with only minor deviations from the test plan. The deviations from the test plan are discussed in the experimental section.

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

    SciTech Connect (OSTI)

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

    2008-01-15T23:59:59.000Z

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

  12. The French national program for spent fuel and high-level waste management

    SciTech Connect (OSTI)

    Giraud, J.P.; Demontalembert, J.A. [COGEMA, Velizy-Villacoublay (France)

    1993-12-31T23:59:59.000Z

    From its very beginning, the French national program for spent fuel and HLW management is aimed at the recycling of energetic materials and the safe disposal of nuclear waste. Spent fuel reprocessing is the cornerstone of this program, since it directly opens the way to energetic material recycling, waste minimization and safe conditioning. It is complemented by the HLW management program which is defined by the HLW disposal regulation and the Waste Act issued in 1991.

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

    Broader source: Energy.gov [DOE]

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

  14. Handbook of high-level radioactive waste transportation

    SciTech Connect (OSTI)

    Sattler, L.R.

    1992-10-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Russell, E.W.; Clarke, W. [Lawrence Livermore National Lab., CA (United States)] [Lawrence Livermore National Lab., CA (United States); Domian, H.A. [Babcock and Wilcox Co., Lynchburg, VA (United States)] [Babcock and Wilcox Co., Lynchburg, VA (United States); Madson, A.A. [Kaiser Engineers California Corp., Oakland, CA (United States)] [Kaiser Engineers California Corp., Oakland, CA (United States)

    1991-08-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Cabeche, Dion Tunick

    2011-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2002-02-27T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    THIEN MG; GREER DA; TOWNSON P

    2011-01-13T23:59:59.000Z

    The ability to effectively mix, sample, certify, and deliver consistent batches of high level waste (HLW) feed from the Hanford double shell tanks (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.

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

    SciTech Connect (OSTI)

    Stefanovsky, Sergey; Marra, James; Lebedev, Vladimir

    2014-01-13T23:59:59.000Z

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

  20. PHYSICAL CHARACTERIZATION OF VITREOUS STATE LABORATORY AY102/C106 AND AZ102 HIGH LEVEL WASTE MELTER FEED SIMULANTS (U)

    SciTech Connect (OSTI)

    Hansen, E

    2005-03-31T23:59:59.000Z

    The objective of this task is to characterize and report specified physical properties and pH of simulant high level waste (HLW) melter feeds (MF) processed through the scaled melters at Vitreous State Laboratories (VSL). The HLW MF simulants characterized are VSL AZ102 straight hydroxide melter feed, VSL AZ102 straight hydroxide rheology adjusted melter feed, VSL AY102/C106 straight hydroxide melter feed, VSL AY102/C106 straight hydroxide rheology adjusted melter feed, and Savannah River National Laboratory (SRNL) AY102/C106 precipitated hydroxide processed sludge blended with glass former chemicals at VSL to make melter feed. The physical properties and pH were characterized using the methods stated in the Waste Treatment Plant (WTP) characterization procedure (Ref. 7).

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

    SciTech Connect (OSTI)

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

    1984-01-01T23:59:59.000Z

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

  2. The Waste Isolation Pilot Plant Deep Geological Repository: A Domestic and Global Blueprint for Safe Disposal of High-Level Radioactive Waste - 12081

    SciTech Connect (OSTI)

    Eriksson, Leif G. [Nuclear Waste Dispositions, Winter Park, Florida 32789 (United States); Dials, George E. [B and W Conversion Services, LLC, Lexington, Kentucky 40513 (United States)

    2012-07-01T23:59:59.000Z

    At the end of 2011, the world's first used/spent nuclear fuel and other long-lived high-level radioactive waste (HLW) repository is projected to open in 2020, followed by two more in 2025. The related pre-opening periods will be at least 40 years, as it also would be if USA's candidate HLW-repository is resurrected by 2013. If abandoned, a new HLW-repository site would be needed. On 26 March 1999, USA began disposing long-lived radioactive waste in a deep geological repository in salt at the Waste Isolation Pilot Plant (WIPP) site. The related pre-opening period was less than 30 years. WIPP has since been re-certified twice. It thus stands to reason the WIPP repository is the global proof of principle for safe deep geological disposal of long-lived radioactive waste. It also stands to reason that the lessons learned since 1971 at the WIPP site provide a unique, continually-updated, blueprint for how the pre-opening period for a new HLW repository could be shortened both in the USA and abroad. (authors)

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

    Office of Environmental Management (EM)

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

  4. Spanish high level radioactive waste management system issues

    SciTech Connect (OSTI)

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

    1993-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Mahidhara, R.K.; Elleman, T.S.; Murty, K.L. [North Carolina State Univ., Raleigh, NC (United States)

    1992-11-01T23:59:59.000Z

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

  6. DOE high-level waste tank safety program. Final report

    SciTech Connect (OSTI)

    NONE

    1998-11-01T23:59:59.000Z

    The overall objective of the work was to provide LANL with support to the DOE High-Level Waste Tank Safety Program. This effort included direct support to the DOE High-Level Waste Tank Working Groups, development of a database to track all identified safety issues, development of requirements for waste tank modernization, evaluation of external comments regarding safety-related guidance/instruction developed previously, examination of current federal and state regulations associated with DOE Tank farm operations, and performance of a conduct of operations review. All tasks which were assigned under this Task Order were completed. Descriptions of the objectives of each task and effort performed to complete each objective is provided.

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

    SciTech Connect (OSTI)

    Thien, Mike G. [Washington River Protection Solutions, LLC, Richland, WA (United States); Barnes, Steve M. [URS, Richland, WA (United States)

    2013-01-17T23:59:59.000Z

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described.

  8. A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests - 13342

    SciTech Connect (OSTI)

    Thien, Mike G. [Washington River Protection Solutions, LLC, P.O Box 850, Richland WA, 99352 (United States)] [Washington River Protection Solutions, LLC, P.O Box 850, Richland WA, 99352 (United States); Barnes, Steve M. [Waste Treatment Plant, 2435 Stevens Center Place, Richland WA 99354 (United States)] [Waste Treatment Plant, 2435 Stevens Center Place, Richland WA 99354 (United States)

    2013-07-01T23:59:59.000Z

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described. (authors)

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

    SciTech Connect (OSTI)

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

    1998-12-01T23:59:59.000Z

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

  10. 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.

    1984-01-01T23:59:59.000Z

    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.

  11. Progress in High-Level Waste Tank Cleaning at the Idaho National Environmental and Engineering Laboratory

    SciTech Connect (OSTI)

    Lockie, K. A.; McNaught, W. B.

    2002-02-26T23:59:59.000Z

    The Department of Energy Idaho Operations Office (DOE-ID) is making preparations to close two underground high-level waste (HLW) storage tanks at the Idaho National Engineering and Environmental Laboratory (INEEL) to meet Resource Conservation and Recovery Act (RCRA) regulations and Department of Energy (DOE) orders. Closure of these two tanks is scheduled for 2004 as the first phase in closure of the eleven 300,000 gallon tanks currently in service at the Idaho Nuclear Technology and Engineering Center (INTEC). Design, development, and deployment of a remotely operated tank cleaning system were completed in August 2001. The system incorporates many commercially available components, which have been adapted for application in cleaning high-level waste tanks. The system also uses existing waste transfer technology (steam-jets) to remove tank heel solids from the tank bottoms during the cleaning operations. By using this existing transfer system and commercially available equipment, the cost of developing custom designed cleaning equipment can be avoided. Remotely operated directional spray nozzles, automatic rotating wash balls, video monitoring equipment, decontamination spray-rings, and tank specific access interface devices have been integrated to provide a system that efficiently cleans tank walls and heel solids in an acidic, radioactive environment. This system is also compliant with operational and safety performance requirements at INTEC. Through the deployment of the tank cleaning system, the INEEL High Level Waste Program has demonstrated the capability to clean tanks to meet RCRA clean closure standards and DOE closure performance measures. The tank cleaning system deployed at the INTEC offers unique advantages over other approaches evaluated at the INEEL and throughout the DOE Complex. The system's ability to agitate and homogenize the tank heel sludge will simplify verification-sampling techniques and reduce the total quantity of samples required to demonstrate compliance with the performance standards. This will reduce tank closure budget requirements and improve closure-planning schedules.

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

    SciTech Connect (OSTI)

    N /A

    2005-06-30T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Reboul, S. H.

    2013-09-30T23:59:59.000Z

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

  14. High-level waste tank farm set point document

    SciTech Connect (OSTI)

    Anthony, J.A. III

    1995-01-15T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2014-05-15T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Quigley, K.D.; Wessman, D

    2003-02-27T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2009-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Mendel, J.E. (compiler)

    1984-08-01T23:59:59.000Z

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

  19. E-Print Network 3.0 - aging high-level waste Sample Search Results

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

    or transportation of high-level radioactive waste... on which the Secretary begins disposal of ... Source: U.S. Nuclear Waste Technical Review Board Collection: Fission and...

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

    SciTech Connect (OSTI)

    WILLIS, W.L.

    2000-06-15T23:59:59.000Z

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

  1. THE RETRIEVAL KNOWLEDGE CENTER EVALUATION OF LOW TANK LEVEL MIXING TECHNOLOGIES FOR DOE HIGH LEVEL WASTE TANK RETRIEVAL 10516

    SciTech Connect (OSTI)

    Fellinger, A.

    2009-12-08T23:59:59.000Z

    The Department of Energy (DOE) Complex has over two-hundred underground storage tanks containing over 80-million gallons of legacy waste from the production of nuclear weapons. The majority of the waste is located at four major sites across the nation and is planned for treatment over a period of almost forty years. The DOE Office of Technology Innovation & Development within the Office of Environmental Management (DOE-EM) sponsors technology research and development programs to support processing advancements and technology maturation designed to improve the costs and schedule for disposal of the waste and closure of the tanks. Within the waste processing focus area are numerous technical initiatives which included the development of a suite of waste removal technologies to address the need for proven equipment and techniques to remove high level radioactive wastes from the waste tanks that are now over fifty years old. In an effort to enhance the efficiency of waste retrieval operations, the DOE-EM Office of Technology Innovation & Development funded an effort to improve communications and information sharing between the DOE's major waste tank locations as it relates to retrieval. The task, dubbed the Retrieval Knowledge Center (RKC) was co-lead by the Savannah River National Laboratory (SRNL) and the Pacific Northwest National Laboratory (PNNL) with core team members representing the Oak Ridge and Idaho sites, as well as, site contractors responsible for waste tank operations. One of the greatest challenges to the processing and closure of many of the tanks is complete removal of all tank contents. Sizeable challenges exist for retrieving waste from High Level Waste (HLW) tanks; with complications that are not normally found with tank retrieval in commercial applications. Technologies currently in use for waste retrieval are generally adequate for bulk removal; however, removal of tank heels, the materials settled in the bottom of the tank, using the same technology have proven to be difficult. Through the RKC, DOE-EM funded an evaluation of adaptable commercial technologies that could assist with the removal of the tank heels. This paper will discuss the efforts and results of developing the RKC to improve communications and discussion of tank waste retrieval through a series of meetings designed to identify technical gaps in retrieval technologies at the DOE Hanford and Savannah River Sites. This paper will also describe the results of an evaluation of commercially available technologies for low level mixing as they might apply to HLW tank heel retrievals.

  2. Transmutation of high-level radioactive waste - Perspectives

    E-Print Network [OSTI]

    Junghans, Arnd; Grosse, Eckart; Hannaske, Roland; Kögler, Toni; Massarczyk, Ralf; Schwengner, Ronald; Wagner, Andreas

    2014-01-01T23:59:59.000Z

    In a fast neutron spectrum essentially all long-lived actinides (e.g. Plutonium) undergo fission and thus can be transmuted into generally short lived fission products. Innovative nuclear reactor concepts e.g. accelerator driven systems (ADS) are currently in development that foresee a closed fuel cycle. The majority of the fissile nuclides (uranium, plutonium) shall be used for power generation and only fission products will be put into final disposal that needs to last for a historical time scale of only 1000 years. For the transmutation of high-level radioactive waste a lot of research and development is still required. One aspect is the precise knowledge of nuclear data for reactions with fast neutrons. Nuclear reactions relevant for transmutation are being investigated in the framework of the european project ERINDA. First results from the new neutron time-of-flight facility nELBE at Helmholtz-Zentrum Dresden-Rossendorf will be presented.

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

    SciTech Connect (OSTI)

    KETUSKY, EDWARD

    2006-04-20T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2012-01-10T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2002-02-25T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1996-08-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2006-04-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Anthony, J.A. III

    1995-01-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    D.C. Richardson

    2003-03-19T23:59:59.000Z

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

  10. Final Report - High Level Waste Vitrification System Improvements, VSL-07R1010-1, Rev 0, dated 04/16/07

    SciTech Connect (OSTI)

    Kruger, Albert A.; Gan, H.; Pegg, I. L.; Gong, W.; Champman, C. C.; Joseph, I.; Matlack, K. S.

    2013-11-13T23:59:59.000Z

    This report describes work conducted to support the development and testing of new glass formulations that extend beyond those that have been previously investigated for the Hanford Waste Treatment and Immobilization Plant (WTP). The principal objective was to investigate maximization of the incorporation of several waste components that are expected to limit waste loading and, consequently, high level waste (HLW) processing rates and canister count. The work was performed with four waste compositions specified by the Office of River Protection (ORP); these wastes contain high concentrations of bismuth, chromium, aluminum, and aluminum plus sodium. The tests were designed to identify glass formulations that maximize waste loading while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM100) tests designed to examine the effects of enhanced glass formulations, increased glass processing temperature, increased crystallinity, and feed solids content on waste processing rate and product quality.

  11. 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

    2005-09-19T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Lee, S.

    2011-07-05T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2013-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2012-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Kelly, Steven E.

    2013-11-11T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2008-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    BRIGGS, M.G.

    2000-09-22T23:59:59.000Z

    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.

  18. Amended Record of Decision for the Idaho High-Level Waste (HLW) and

    Office of Environmental Management (EM)

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

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion |EnergyonSupport0.pdf5 OPAM SEMIANNUAL REPORT TOJaredKansas1Increase Profits

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

    SciTech Connect (OSTI)

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

    1994-12-01T23:59:59.000Z

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

  1. Hanford high-level waste evaporator/crystallizer corrosion evaluation

    SciTech Connect (OSTI)

    Ohl, P.C.; Carlos, W.C.

    1993-10-01T23:59:59.000Z

    The US Department of Energy, Hanford Site nuclear reservation, located in Southeastern Washington State, is currently home to 61 Mgal of radioactive waste stored in 177 large underground storage tanks. As an intermediate waste volume reduction, the 242-A Evaporator/Crystallizer processes waste solutions from most of the operating laboratories and plants on the Hanford Site. The waste solutions are concentrated in the Evaporator/Crystallizer to a slurry of liquid and crystallized salts. This concentrated slurry is returned to Hanford Site waste tanks at a significantly reduced volume. The Washington State Department of Ecology Dangerous Waste Regulations, WAC 173-393 require that a tank system integrity assessment be completed and maintained on file at the facility for all dangerous waste tank systems. This corrosion evaluation was performed in support of the 242-A Evaporator/Crystallizer Tank System Integrity Assessment Report. This corrosion evaluation provided a comprehensive compatibility study of the component materials and corrosive environments. Materials used for the Evaporator components and piping include austenitic stainless steels (SS) (primarily ASTM A240, Type 304L) and low alloy carbon steels (CS) (primarily ASTM A53 and A106) with polymeric or asbestos gaskets at flanged connections. Building structure and secondary containment is made from ACI 301-72 Structural Concrete for Buildings and coated with a chemically resistant acrylic coating system.

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

    SciTech Connect (OSTI)

    Decker, L.A.

    1998-04-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Aaron, G.; Wilmarth, B.

    2011-09-19T23:59:59.000Z

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

  4. Northeast High-Level Radioactive Waste Transportation Task Force Agenda

    Office of Environmental Management (EM)

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  5. High-Level Waste Corporate Board, Mark Gilbertson

    Office of Environmental Management (EM)

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  6. West Valley Demonstration Project High-Level Waste Management

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergy Cooperation | Department ofEnergy Is Everywhere! Webinar: EnergyDRAFT_19507_1 High-Level

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

    SciTech Connect (OSTI)

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

    2012-09-30T23:59:59.000Z

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

  8. 3-D MAPPING TECHNOLOGIES FOR HIGH LEVEL WASTE TANKS

    SciTech Connect (OSTI)

    Marzolf, A.; Folsom, M.

    2010-08-31T23:59:59.000Z

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

  9. Review of Alternative Technologies for Pretreatment of Accumulated HLW

    SciTech Connect (OSTI)

    Romanovsky, V.; Rimski-Korsakov, A.

    2002-02-26T23:59:59.000Z

    Accumulated liquid high-level wastes (HLW) from nuclear centers in Russia and the United States (U. S.) contain great amounts of nonradioactive salts; it is prudent to vitrify not the entire volume of these wastes, but only the concentrates of radionuclides recovered from them. For this purpose, different pretreatment technologies based on liquid-liquid extraction are under development.

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

    Office of Environmental Management (EM)

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  11. High-Level Waste Corporate Board Meeting Agenda

    Office of Environmental Management (EM)

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  12. High-Level Waste Corporate Board Performance Assessment Subcommittee

    Office of Environmental Management (EM)

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

    SciTech Connect (OSTI)

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

    1994-05-01T23:59:59.000Z

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

  14. High-level waste vitrification off-gas cleanup technology

    SciTech Connect (OSTI)

    Hanson, M.S.

    1980-01-01T23:59:59.000Z

    This brief overview is intended to be a basis for discussion of needs and problems existing in the off-gas clean-up technology. A variety of types of waste form and processes are being developed in the United States and abroad. A description of many of the processes can be found in the Technical Alternative Documents (TAD). Concurrently, off-gas processing systems are being developed with most of the processes. An extensive review of methodology as well as decontamination factors can be found in the literature. Since it is generally agreed that the most advanced solidification process is vitrification, discussion here centers about the off-gas problems related to vitrification. With a number of waste soldification facilities around the world in operation, it can be shown that present technology can satisfy the present requirement for off-gas control. However, a number of areas within the technology base show potential for improvement. Fundamental as well as verification studies are needed to obtain the improvements.

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

    E-Print Network [OSTI]

    Driscoll, Michael J.

    1988-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Peters, T.B.

    2001-09-10T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Gibbs, Jonathan Sutton

    2010-01-01T23:59:59.000Z

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

  18. Defense HLW Glass Degradation Model

    SciTech Connect (OSTI)

    D. Strachan

    2004-10-20T23:59:59.000Z

    The purpose of this report is to document the development of a model for calculating the release rate for radionuclides and other key elements from high-level radioactive waste (HLW) glasses under exposure conditions relevant to the performance of the repository. Several glass compositions are planned for the repository, some of which have yet to be identified (i.e., glasses from Hanford and Idaho National Engineering and Environmental Laboratory). The mechanism for glass dissolution is the same for these glasses and the glasses yet to be developed for the disposal of DOE wastes. All of these glasses will be of a quality consistent with the glasses used to develop this report.

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

    SciTech Connect (OSTI)

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

    2011-01-04T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2012-12-13T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Boyer, Edmond

    Sequential Thermo-Hydraulic Modeling of Variably Saturated Flow in High-Level Radioactive Waste-Malabry, France Key words: waste repository, geological disposal, thermo- hydraulic modeling Introduction The most developed a sequential model to predict the coupled thermo-hydraulic processes at a cell-scale radioactive

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

    SciTech Connect (OSTI)

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

    1981-10-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2002-02-25T23:59:59.000Z

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

  4. Criticality Safety Evaluation of Hanford Site High Level Waste Storage Tanks

    SciTech Connect (OSTI)

    ROGERS, C.A.

    2000-02-17T23:59:59.000Z

    This criticality safety evaluation covers operations for waste in underground storage tanks at the high-level waste tank farms on the Hanford site. This evaluation provides the bases for criticality safety limits and controls to govern receipt, transfer, and long-term storage of tank waste. Justification is provided that a nuclear criticality accident cannot occur for tank farms operations, based on current fissile material and operating conditions.

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

    SciTech Connect (OSTI)

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

    2008-09-23T23:59:59.000Z

    The objective of this task was to develop glass formulations for (Department of Energy) DOE waste streams with high aluminum concentrations to avoid nepheline formation while maintaining or meeting waste loading and/or waste throughput expectations as well as satisfying critical process and product performance related constraints. Liquidus temperatures and crystallization behavior were carefully characterized to support model development for higher waste loading glasses. The experimental work, characterization, and data interpretation necessary to meet these objectives were performed among three partnering laboratories: the V.G. Khlopin Radium Institute (KRI), Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL). Projected glass compositional regions that bound anticipated Defense Waste Processing Facility (DWPF) and Hanford high level waste (HLW) glass regions of interest were developed and used to generate glass compositions of interest for meeting the objectives of this study. A thorough statistical analysis was employed to allow for a wide range of waste glass compositions to be examined while minimizing the number of glasses that had to be fabricated and characterized in the laboratory. The glass compositions were divided into two sets, with 45 in the test matrix investigated by the U.S. laboratories and 30 in the test matrix investigated by KRI. Fabrication and characterization of the US and KRI-series glasses were generally handled separately. This report focuses mainly on the US-series glasses. Glasses were fabricated and characterized by SRNL and PNNL. Crystalline phases were identified by X-ray diffraction (XRD) in the quenched and canister centerline cooled (CCC) glasses and were generally iron oxides and spinels, which are not expected to impact durability of the glass. Nepheline was detected in five of the glasses after the CCC heat treatment. Chemical composition measurements for each of the glasses were conducted following an analytical plan. A review of the individual oxides for each glass revealed that there were no errors in batching significant enough to impact the outcome of the study. A comparison of the measured compositions of the replicates indicated an acceptable degree of repeatability as the percent differences for most of the oxides were less than 5% and percent differences for all of the oxides were less than 10 wt%. Chemical durability was measured using the Product Consistency Test (PCT). All but two of the study glasses had normalized leachate for boron (NL [B]) values that were well below that of the Environmental Assessment (EA) reference glass. The two highest NL [B] values were for the CCC versions of glasses US-18 and US-27 (10.498 g/L and 15.962 g/L, respectively). Nepheline crystallization was identified by qualitative XRD in five of the US-series glasses. Each of these five glasses (US-18, US-26, US-27, US-37 and US-43) showed a significant increase in NL [B] values after the CCC heat treatment. This reduction in durability can be attributed to the formation of nepheline during the slow cooling cycle and the removal of glass formers from the residual glass network. The liquidus temperature (T{sub L}) of each glass in the study was determined by both optical microscopy and XRD methods. The correlation coefficient of the measured XRD TL data versus the measured optical TL data was very good (R{sup 2} = 0.9469). Aside from a few outliers, the two datasets aligned very well across the entire temperature range (829 C to 1312 C for optical data and 813 C to 1310 C for XRD crystal fraction data). The data also correlated well with the predictions of a PNNL T{sub L} model. The correlation between the measured and calculated data had a higher degree of merit for the XRD crystal fraction data than for the optical data (higher R{sup 2} value of 0.9089 versus 0.8970 for the optical data). The SEM-EDS analysis of select samples revealed the presence of undissolved RuO{sub 2} in all glasses due to the low solubility of RuO{sub 2} in borosilicate glass. These

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

    SciTech Connect (OSTI)

    Manuel, A.F.

    1996-02-05T23:59:59.000Z

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

  7. The Savannah River Site Replacement High Level Radioactive Waste Evaporator Project

    SciTech Connect (OSTI)

    Presgrove, S.B. (Bechtel Savannah River, Inc., North Augusta, SC (United States))

    1992-01-01T23:59:59.000Z

    The Replacement High Level Waste Evaporator Project was conceived in 1985 to reduce the volume of the high level radioactive waste Process of the high level waste has been accomplished up to this time using Bent Tube type evaporators and therefore, that type evaporator was selected for this project. The Title I Design of the project was 70% completed in late 1990. The Department of Energy at that time hired an independent consulting firm to perform a complete review of the project. The DOE placed a STOP ORDER on purchasing the evaporator in January 1991. Essentially, no construction was to be done on this project until all findings and concerns dealing with the type and design of the evaporator are resolved. This report addresses two aspects of the DOE design review; (1) Comparing the Bent Tube Evaporator with the Forced Circulation Evaporator, (2) The design portion of the DOE Project Review - concentrated on the mechanical design properties of the evaporator. 1 ref.

  8. The Savannah River Site Replacement High Level Radioactive Waste Evaporator Project

    SciTech Connect (OSTI)

    Presgrove, S.B. [Bechtel Savannah River, Inc., North Augusta, SC (United States)

    1992-08-01T23:59:59.000Z

    The Replacement High Level Waste Evaporator Project was conceived in 1985 to reduce the volume of the high level radioactive waste Process of the high level waste has been accomplished up to this time using Bent Tube type evaporators and therefore, that type evaporator was selected for this project. The Title I Design of the project was 70% completed in late 1990. The Department of Energy at that time hired an independent consulting firm to perform a complete review of the project. The DOE placed a STOP ORDER on purchasing the evaporator in January 1991. Essentially, no construction was to be done on this project until all findings and concerns dealing with the type and design of the evaporator are resolved. This report addresses two aspects of the DOE design review; (1) Comparing the Bent Tube Evaporator with the Forced Circulation Evaporator, (2) The design portion of the DOE Project Review - concentrated on the mechanical design properties of the evaporator. 1 ref.

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

    SciTech Connect (OSTI)

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

    1998-06-18T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    R.A. Levich; J.S. Stuckless

    2006-09-25T23:59:59.000Z

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

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

    DOE Patents [OSTI]

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

    1989-01-01T23:59:59.000Z

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

  12. 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.

    1984-04-11T23:59:59.000Z

    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

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

    SciTech Connect (OSTI)

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

    1984-04-01T23:59:59.000Z

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

  14. 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."

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

    SciTech Connect (OSTI)

    Easterline, D.; Kunreuther, H.

    1995-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Han, F.C.

    1996-09-30T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1994-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1994-04-01T23:59:59.000Z

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

  19. 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

    2011-08-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1995-02-01T23:59:59.000Z

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

  1. COMPUTER MODELING OF HIGH-LEVEL WASTE GLASS TEMPERATURES WITHIN DWPF CANISTERS DURING POURING AND COOL DOWN

    SciTech Connect (OSTI)

    Amoroso, J.

    2011-10-09T23:59:59.000Z

    This report describes the results of a computer simulation study to predict the temperature of the glass at any location inside a DWPF canister during pouring and subsequent cooling. These simulations are an integral part of a larger research focus aimed at developing methods to predict, evaluate, and ultimately suppress nepheline formation in HLW glasses. That larger research focus is centered on holistically understanding nepheline formation in HLW glass by exploring the fundamental thermal and chemical driving forces for nepheline crystallization with respect to realistic processing conditions. Through experimental work, the goal is to integrate nepheline crystallization potential in HLW glass with processing capability to ultimately optimize waste loading and throughput while maintaining an acceptable product with respect to durability. The results of this study indicated severe temperature gradients and prolonged temperature dwell times exist throughout different locations in the canister and that the time and temperatures that HLW glass is subjected to during processing is a function of pour rate. The simulations indicate that crystallization driving forces are not uniform throughout the glass volume in a DWPF (or DWPF-like) canister and illustrate the importance of considering overall kinetics (chemical and thermal driving forces) of nepheline formation when developing methods to predict and suppress its formation in HLW glasses. The intended path forward is to use the simulation data both as a driver for future experimental work and, as an investigative tool for evaluating the impact of experimental results. Simulation data will be used to develop laboratory experiments to more acutely evaluate nepheline formation in HLW glass by incorporating the simulated temperatures throughout the canister into the laboratory experiments. Concurrently, laboratory experiments will be performed to identify nepheline crystallization potential in HLW glass as a function of time and temperature, the results of which will be fed back into simulations to evaluate the potential impacts. Through an iterative process involving computer simulations and experimental results, the potential for nepheline crystallization in HLW glass can be predicted, evaluated, and suppressed to maximize waste loading and throughput of canisters.

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

    SciTech Connect (OSTI)

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

    2004-08-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    HILL RC PHILIP; REYNOLDS JG; RUTLAND PL

    2011-02-23T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2011-10-01T23:59:59.000Z

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

  5. 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

    2005-10-31T23:59:59.000Z

    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.

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

    Energy Savers [EERE]

    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...

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

    SciTech Connect (OSTI)

    Crawford, T.W.

    1995-09-22T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1994-03-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1996-09-30T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1994-03-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Larson, D.E. (ed.)

    1980-09-01T23:59:59.000Z

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

  12. Safety analysis report vitrified high level waste type B shipping cask

    SciTech Connect (OSTI)

    NONE

    1995-03-01T23:59:59.000Z

    This Safety Analysis Report describes the design, analyses, and principle features of the Vitrified High Level Waste (VHLW) Cask. In preparing this report a detailed evaluation of the design has been performed to ensure that all safety, licensing, and operational goals for the cask and its associated Department of Energy program can be met. The functions of this report are: (1) to fully document that all functional and regulatory requirements of 10CFR71 can be met by the package; and (2) to document the design and analyses of the cask for review by the Nuclear Regulatory Commission. The VHLW Cask is the reusable shipping package designed by GNSI under Department of Energy contract DE-AC04-89AL53-689 for transportation of Vitrified High Level Waste, and to meet the requirements for certification under 10CFR71 for a Type B(U) package. The VHLW cask has been designed as packaging for transport of canisters of Vitrified High Level Waste solidified at Department of Energy facilities.

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

    SciTech Connect (OSTI)

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

    1993-12-31T23:59:59.000Z

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

  14. Progress achieved in HLW vitrification techniques at INE

    SciTech Connect (OSTI)

    Grunewald, W.; Roth, G.; Tobie, W.; Weisenburger, S. [Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Nucleare Entsorgungstechnik

    1993-12-31T23:59:59.000Z

    The progress in the liquid-fed ceramic waste glass melter process for high level waste vitrification is described. The technique has been used in the PAMELA plant in Mol/Belgium from 1985 to 1991. Currently three programs are underway at INE (Institut fuer Nukleare Entsorgungstechnik): a technology program for optimizing the process for noble metals containing high level waste (HLW), a vitrification technology transfer project with China, and a research project on noble metals behavior in an engineering scale melter which is funded by the US Department of Energy with oversight by the Pacific Northwest Laboratory (PNL) in Richland, WA. The status of the programs and results available are described.

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

    SciTech Connect (OSTI)

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

    2008-01-23T23:59:59.000Z

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

  16. RADIOACTIVE HIGH LEVEL WASTE TANK PITTING PREDICTIONS: AN INVESTIGATION INTO CRITICAL SOLUTION CONCENTRATIONS

    SciTech Connect (OSTI)

    Hoffman, E.

    2012-11-08T23:59:59.000Z

    A series of cyclic potentiodynamic polarization tests was performed on samples of ASTM A537 carbon steel in support of a probability-based approach to evaluate the effect of chloride and sulfate on corrosion the steel?s susceptibility to pitting corrosion. Testing solutions were chosen to systemically evaluate the influence of the secondary aggressive species, chloride, and sulfate, in the nitrate based, high-level wastes. The results suggest that evaluating the combined effect of all aggressive species, nitrate, chloride, and sulfate, provides a consistent response for determining corrosion susceptibility. The results of this work emphasize the importance for not only nitrate concentration limits, but also chloride and sulfate concentration limits.

  17. High performance gamma measurements of equipment retrieved from Hanford high-level nuclear waste tanks

    SciTech Connect (OSTI)

    Troyer, G.L.

    1997-03-17T23:59:59.000Z

    The cleanup of high level defense nuclear waste at the Hanford site presents several progressive challenges. Among these is the removal and disposal of various components from buried active waste tanks to allow new equipment insertion or hazards mitigation. A unique automated retrieval system at the tank provides for retrieval, high pressure washing, inventory measurement, and containment for disposal. Key to the inventory measurement is a three detector HPGe high performance gamma spectroscopy system capable of recovering data at up to 90% saturation (200,000 counts per second). Data recovery is based on a unique embedded electronic pulser and specialized software to report the inventory. Each of the detectors have different shielding specified through Monte Carlo simulation with the MCNP program. This shielding provides performance over a dynamic range of eight orders of magnitude. System description, calibration issues and operational experiences are discussed.

  18. Conceptual modular description of the high-level waste management system for system studies model development

    SciTech Connect (OSTI)

    McKee, R.W.; Young, J.R.; Konzek, G.J.

    1992-08-01T23:59:59.000Z

    This document presents modular descriptions of possible alternative components of the federal high-level radioactive waste management system and the procedures for combining these modules to obtain descriptions for alternative configurations of that system. The 20 separate system component modules presented here can be combined to obtain a description of any of the 17 alternative system configurations (i.e., scenarios) that were evaluated in the MRS Systems Studies program (DOE 1989a). First-approximation descriptions of other yet-undefined system configurations could also be developed for system study purposes from this database. The descriptions include, in a modular format, both functional descriptions of the processes in the waste management system, plus physical descriptions of the equipment and facilities necessary for performance of those functions.

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergy Cooperation |South42.2 (April 2012) 1 Documentation and ApprovalAmanda Scott AboutAdditional

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

    SciTech Connect (OSTI)

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

    2005-10-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Jolly, R; Bruce Martin, B

    2008-01-15T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2012-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2011-08-01T23:59:59.000Z

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

  4. C-104 high-level waste solids: Washing/leaching and solubility versus temperature studies

    SciTech Connect (OSTI)

    GJ Lumetta; DJ Bates; JP Bramson; LP Darnell; OT Farmer III; SK Fiskum; LR Greenwood; FV Hoopes; CZ Soderquist; MJ Steele; RT Steele; MW Urie; JJ Wagner

    2000-05-17T23:59:59.000Z

    This report describes the results of a test conducted by Battelle to assess the effects of inhibited water washing and caustic leaching on the composition of the C-104 HLW solids. The objective of this work was to determine the composition of the C-104 solids remaining after washing with 0.01 M NaOH or leaching with 3 M NaOH. Another objective of this test was to determine the solubility of the C-104 solids as a function of temperature. The work was conducted according to test plan BNFL-TP-29953-8, Rev. 0, ``Determination of the Solubility of HLW Sludge Solids.

  5. Hanford Tank Waste Information Enclosure 1 Hanford Tank Waste Information

    E-Print Network [OSTI]

    Hanford Tank Waste Information Enclosure 1 1 Hanford Tank Waste Information 1.0 Summary This information demonstrates the wastes in the twelve Hanford Site tanks meet the definition of transuranic (TRU. The wastes in these twelve (12) tanks are not high-level waste (HLW), and contain more than 100 nanocuries

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

    E-Print Network [OSTI]

    Kammen, Daniel M.

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

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-SeriesFlickr FlickrGuidedCH2M HILLAdministrationHigh SchoolHighHIGH WASTE

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

    SciTech Connect (OSTI)

    Hrma, Pavel R.; Riley, Brian J.; Crum, Jarrod V.; Matyas, Josef

    2014-01-15T23:59:59.000Z

    Spinel crystals precipitate in high-level waste glasses containing Fe, Cr, Ni , Mn, Zn, and Ru. The liquidus temperature (TL) of spinel as the primary crystallization phase is a function of glass composition and the spinel solubility (c0) is a function of both glass composition and temperature (T). Previously reported models of TL as a function of composition are based on TL measured directly, which requires laborious experimental procedures. Viewing the curve of c0 versus T as the liquidus line allows a significant broadening of the composition region for model fitting. This paper estimates TL as a function of composition based on c0 data obtained with the X-ray diffraction technique.

  9. Guidelines for development of structural integrity programs for DOE high-level waste storage tanks

    SciTech Connect (OSTI)

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

    1997-01-01T23:59:59.000Z

    Guidelines are provided for developing programs to promote the structural integrity of high-level waste storage tanks and transfer lines at the facilities of the Department of Energy. Elements of the program plan include a leak-detection system, definition of appropriate loads, collection of data for possible material and geometric changes, assessment of the tank structure, and non-destructive examination. Possible aging degradation mechanisms are explored for both steel and concrete components of the tanks, and evaluated to screen out nonsignificant aging mechanisms and to indicate methods of controlling the significant aging mechanisms. Specific guidelines for assessing structural adequacy will be provided in companion documents. Site-specific structural integrity programs can be developed drawing on the relevant portions of the material in this document.

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

    SciTech Connect (OSTI)

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Riley, Brian J. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hrma, Pavel [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

    2012-11-15T23:59:59.000Z

    Spinel crystals precipitate in high-level waste glasses containing Fe, Cr, Ni, Mn, Zn, and Ru. The liquidus temperature (T{sub L}d) of spinel as the primary crystallization phase is a function of glass composition, and the spinel solubility (c{sub o}) is a function of both glass composition and temperature (T). Previously reported models of T{sub L} as a function of composition are based on T{sub L} measured directly, which requires laborious experimental procedures. Viewing the curve of c{sub o} versus T as the liquidus line allows a significant broadening of the composition region for model fitting. This paper estimates T{sub L} as a function of composition based on c{sub o} data obtained with the X-ray diffraction technique.

  11. Shale disposal of U.S. high-level radioactive waste.

    SciTech Connect (OSTI)

    Sassani, David Carl; Stone, Charles Michael; Hansen, Francis D.; Hardin, Ernest L.; Dewers, Thomas A.; Martinez, Mario J.; Rechard, Robert Paul; Sobolik, Steven Ronald; Freeze, Geoffrey A.; Cygan, Randall Timothy; Gaither, Katherine N.; Holland, John Francis; Brady, Patrick Vane

    2010-05-01T23:59:59.000Z

    This report evaluates the feasibility of high-level radioactive waste disposal in shale within the United States. The U.S. has many possible clay/shale/argillite basins with positive attributes for permanent disposal. Similar geologic formations have been extensively studied by international programs with largely positive results, over significant ranges of the most important material characteristics including permeability, rheology, and sorptive potential. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in shale media. We develop scoping performance analyses, based on the applicable features, events, and processes identified by international investigators, to support a generic conclusion regarding post-closure safety. Requisite assumptions for these analyses include waste characteristics, disposal concepts, and important properties of the geologic formation. We then apply lessons learned from Sandia experience on the Waste Isolation Pilot Project and the Yucca Mountain Project to develop a disposal strategy should a shale repository be considered as an alternative disposal pathway in the U.S. Disposal of high-level radioactive waste in suitable shale formations is attractive because the material is essentially impermeable and self-sealing, conditions are chemically reducing, and sorption tends to prevent radionuclide transport. Vertically and laterally extensive shale and clay formations exist in multiple locations in the contiguous 48 states. Thermal-hydrologic-mechanical calculations indicate that temperatures near emplaced waste packages can be maintained below boiling and will decay to within a few degrees of the ambient temperature within a few decades (or longer depending on the waste form). Construction effects, ventilation, and the thermal pulse will lead to clay dehydration and deformation, confined to an excavation disturbed zone within a few meters of the repository, that can be reasonably characterized. Within a few centuries after waste emplacement, overburden pressures will seal fractures, resaturate the dehydrated zones, and provide a repository setting that strongly limits radionuclide movement to diffusive transport. Coupled hydrogeochemical transport calculations indicate maximum extents of radionuclide transport on the order of tens to hundreds of meters, or less, in a million years. Under the conditions modeled, a shale repository could achieve total containment, with no releases to the environment in undisturbed scenarios. The performance analyses described here are based on the assumption that long-term standards for disposal in clay/shale would be identical in the key aspects, to those prescribed for existing repository programs such as Yucca Mountain. This generic repository evaluation for shale is the first developed in the United States. Previous repository considerations have emphasized salt formations and volcanic rock formations. Much of the experience gained from U.S. repository development, such as seal system design, coupled process simulation, and application of performance assessment methodology, is applied here to scoping analyses for a shale repository. A contemporary understanding of clay mineralogy and attendant chemical environments has allowed identification of the appropriate features, events, and processes to be incorporated into the analysis. Advanced multi-physics modeling provides key support for understanding the effects from coupled processes. The results of the assessment show that shale formations provide a technically advanced, scientifically sound disposal option for the U.S.

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

    E-Print Network [OSTI]

    Sheffield, University of

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

  13. High level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 6

    SciTech Connect (OSTI)

    Not Available

    1994-04-01T23:59:59.000Z

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 6) outlines the standards and requirements for the sections on: Environmental Restoration and Waste Management, Research and Development and Experimental Activities, and Nuclear Safety.

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

    SciTech Connect (OSTI)

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

    1982-11-01T23:59:59.000Z

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

  15. Hanford High-Level Waste Vitrification Program at the Pacific Northwest National Laboratory: technology development - annotated bibliography

    SciTech Connect (OSTI)

    Larson, D.E.

    1996-09-01T23:59:59.000Z

    This report provides a collection of annotated bibliographies for documents prepared under the Hanford High-Level Waste Vitrification (Plant) Program. The bibliographies are for documents from Fiscal Year 1983 through Fiscal Year 1995, and include work conducted at or under the direction of the Pacific Northwest National Laboratory. The bibliographies included focus on the technology developed over the specified time period for vitrifying Hanford pretreated high-level waste. The following subject areas are included: General Documentation; Program Documentation; High-Level Waste Characterization; Glass Formulation and Characterization; Feed Preparation; Radioactive Feed Preparation and Glass Properties Testing; Full-Scale Feed Preparation Testing; Equipment Materials Testing; Melter Performance Assessment and Evaluations; Liquid-Fed Ceramic Melter; Cold Crucible Melter; Stirred Melter; High-Temperature Melter; Melter Off-Gas Treatment; Vitrification Waste Treatment; Process, Product Control and Modeling; Analytical; and Canister Closure, Decontamination, and Handling

  16. TWRS retrieval and disposal mission, immobilized high-level waste storage plan

    SciTech Connect (OSTI)

    Calmus, R.B.

    1998-01-07T23:59:59.000Z

    This project plan has a two fold purpose. First, it provides a plan specific to the Hanford Tank Waste Remediation System (TWRS) Immobilized High-Level Waste (EMW) Storage Subproject for the Washington State Department of Ecology (Ecology) that meets the requirements of Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) milestone M-90-01 (Ecology et al. 1996) and is consistent with the project plan content guidelines found in Section 11.5 of the Tri-Party Agreement action plan. Second, it provides an upper tier document that can be used as the basis for future subproject line item construction management plans. The planning elements for the construction management plans are derived from applicable U.S. Department of Energy (DOE) planning guidance documents (DOE Orders 4700.1 (DOE 1992a) and 430.1 (DOE 1995)). The format and content of this project plan are designed to accommodate the plan`s dual purpose. A cross-check matrix is provided in Appendix A to explain where in the plan project planning elements required by Section 11.5 of the Tri-Party Agreement are addressed.

  17. SUMO, System performance assessment for a high-level nuclear waste repository: Mathematical models

    SciTech Connect (OSTI)

    Eslinger, P.W.; Miley, T.B.; Engel, D.W.; Chamberlain, P.J. II

    1992-09-01T23:59:59.000Z

    Following completion of the preliminary risk assessment of the potential Yucca Mountain Site by Pacific Northwest Laboratory (PNL) in 1988, the Office of Civilian Radioactive Waste Management (OCRWM) of the US Department of Energy (DOE) requested the Performance Assessment Scientific Support (PASS) Program at PNL to develop an integrated system model and computer code that provides performance and risk assessment analysis capabilities for a potential high-level nuclear waste repository. The system model that has been developed addresses the cumulative radionuclide release criteria established by the US Environmental Protection Agency (EPA) and estimates population risks in terms of dose to humans. The system model embodied in the SUMO (System Unsaturated Model) code will also allow benchmarking of other models being developed for the Yucca Mountain Project. The system model has three natural divisions: (1) source term, (2) far-field transport, and (3) dose to humans. This document gives a detailed description of the mathematics of each of these three divisions. Each of the governing equations employed is based on modeling assumptions that are widely accepted within the scientific community.

  18. PERFORMANCE OF A BURIED RADIOACTIVE HIGH LEVEL WASTE GLASS AFTER 24 YEARS

    SciTech Connect (OSTI)

    Jantzen, C; Daniel Kaplan, D; Ned Bibler, N; David Peeler, D; John Plodinec, J

    2008-05-05T23:59:59.000Z

    A radioactive high level waste glass was made in 1980 with Savannah River Site (SRS) Tank 15 waste. This glass was buried in the SRS burial ground for 24 years but lysimeter data was only available for the first 8 years. The glass was exhumed and analyzed in 2004. The glass was predicted to be very durable and laboratory tests confirmed the durability response. The laboratory results indicated that the glass was very durable as did analysis of the lysimeter data. Scanning electron microscopy of the glass burial surface showed no significant glass alteration consistent with the results of the laboratory and field tests. No detectable Pu, Am, Cm, Np, or Ru leached from the glass into the surrounding sediment. Leaching of {beta}/{delta} from {sup 90}Sr and {sup 137}Cs in the glass was diffusion controlled. Less than 0.5% of the Cs and Sr in the glass leached into the surrounding sediment, with >99% of the leached radionuclides remaining within 8 centimeters of the glass pellet.

  19. Design and performance of atomizing nozzles for spray calcination of high-level wastes

    SciTech Connect (OSTI)

    Miller, F.A.; Stout, L.A.

    1981-05-01T23:59:59.000Z

    A key aspect of high-level liquid-waste spray calcination is waste-feed atomization by using air atomizing nozzles. Atomization substantially increases the heat transfer area of the waste solution, which enhances rapid drying. Experience from the spray-calciner operations has demonstrated that nozzle flow conditions that produce 70-..mu.. median-volume-diameter or smaller spray droplets are required for small-scale spray calciners (drying capacity less than 80 L/h). For large-scale calciners (drying capacity greater than 300 L/h), nozzle flow conditions that produce 100-..mu.. median-volume-diameter or smaller spray droplets are required. Mass flow ratios of 0.2 to 0.4, depending on nozzle size, are required for proper operation of internal-mix atomizing nozzles. Both internal-mix and external-mix nozzles have been tested at PNL. Due to the lower airflow requirements and fewer large droplets produced, the internal-mix nozzle has been chosen for primary development in the spray calciner program at PNL. Several nozzle air-cap materials for internal-mix nozzles have been tested for wear resistance. Results show that nozzle air caps of stainless steel and Cer-vit (a machineable glass ceramic) are suceptible to rapid wear by abrasive slurries, whereas air caps of alumina and reaction-bonded silicon nitride show only slow wear. Longer-term testing is necessary to determine more accurately the actual frequency of nozzle replacement. Atomizing nozzle air caps of alumina are subject to fracture from thermal shock, whereas air caps of silicon nitride and Cer-vit are not. Fractured nozzles are held in place by the air-cap retaining ring and continue to atomize satisfactorily. Therefore, fractures caused by thermal shocking do not necessarily result in nozzle failure.

  20. Thermal-mechanical modeling of deep borehole disposal of high-level radioactive waste.

    SciTech Connect (OSTI)

    Arnold, Bill Walter; Hadgu, Teklu

    2010-12-01T23:59:59.000Z

    Disposal of high-level radioactive waste, including spent nuclear fuel, in deep (3 to 5 km) boreholes is a potential option for safely isolating these wastes from the surface and near-surface environment. Existing drilling technology permits reliable and cost-effective construction of such deep boreholes. Conditions favorable for deep borehole disposal in crystalline basement rocks, including low permeability, high salinity, and geochemically reducing conditions, exist at depth in many locations, particularly in geologically stable continental regions. Isolation of waste depends, in part, on the effectiveness of borehole seals and potential alteration of permeability in the disturbed host rock surrounding the borehole. Coupled thermal-mechanical-hydrologic processes induced by heat from the radioactive waste may impact the disturbed zone near the borehole and borehole wall stability. Numerical simulations of the coupled thermal-mechanical response in the host rock surrounding the borehole were conducted with three software codes or combinations of software codes. Software codes used in the simulations were FEHM, JAS3D, Aria, and Adagio. Simulations were conducted for disposal of spent nuclear fuel assemblies and for the higher heat output of vitrified waste from the reprocessing of fuel. Simulations were also conducted for both isotropic and anisotropic ambient horizontal stress in the host rock. Physical, thermal, and mechanical properties representative of granite host rock at a depth of 4 km were used in the models. Simulation results indicate peak temperature increases at the borehole wall of about 30 C and 180 C for disposal of fuel assemblies and vitrified waste, respectively. Peak temperatures near the borehole occur within about 10 years and decline rapidly within a few hundred years and with distance. The host rock near the borehole is placed under additional compression. Peak mechanical stress is increased by about 15 MPa (above the assumed ambient isotropic stress of 100 MPa) at the borehole wall for the disposal of fuel assemblies and by about 90 MPa for vitrified waste. Simulated peak volumetric strain at the borehole wall is about 420 and 2600 microstrain for the disposal of fuel assemblies and vitrified waste, respectively. Stress and volumetric strain decline rapidly with distance from the borehole and with time. Simulated peak stress at and parallel to the borehole wall for the disposal of vitrified waste with anisotropic ambient horizontal stress is about 440 MPa, which likely exceeds the compressive strength of granite if unconfined by fluid pressure within the borehole. The relatively small simulated displacements and volumetric strain near the borehole suggest that software codes using a nondeforming grid provide an adequate approximation of mechanical deformation in the coupled thermal-mechanical model. Additional modeling is planned to incorporate the effects of hydrologic processes coupled to thermal transport and mechanical deformation in the host rock near the heated borehole.

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

    SciTech Connect (OSTI)

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

    1989-09-01T23:59:59.000Z

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

  2. A literature review of coupled thermal-hydrologic-mechanical-chemical processes pertinent to the proposed high-level nuclear waste repository at Yucca Mountain

    SciTech Connect (OSTI)

    Manteufel, R.D.; Ahola, M.P.; Turner, D.R.; Chowdhury, A.H. [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

    1993-07-01T23:59:59.000Z

    A literature review has been conducted to determine the state of knowledge available in the modeling of coupled thermal (T), hydrologic (H), mechanical (M), and chemical (C) processes relevant to the design and/or performance of the proposed high-level waste (HLW) repository at Yucca Mountain, Nevada. The review focuses on identifying coupling mechanisms between individual processes and assessing their importance (i.e., if the coupling is either important, potentially important, or negligible). The significance of considering THMC-coupled processes lies in whether or not the processes impact the design and/or performance objectives of the repository. A review, such as reported here, is useful in identifying which coupled effects will be important, hence which coupled effects will need to be investigated by the US Nuclear Regulatory Commission in order to assess the assumptions, data, analyses, and conclusions in the design and performance assessment of a geologic reposit``. Although this work stems from regulatory interest in the design of the geologic repository, it should be emphasized that the repository design implicitly considers all of the repository performance objectives, including those associated with the time after permanent closure. The scope of this review is considered beyond previous assessments in that it attempts with the current state-of-knowledge) to determine which couplings are important, and identify which computer codes are currently available to model coupled processes.

  3. 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.

    1996-03-01T23:59:59.000Z

    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.

  4. Alternative concepts for treatment and disposal of Hanford site high-level waste in tanks

    SciTech Connect (OSTI)

    Claghorn, R.D.; Powell, W.J.

    1994-12-01T23:59:59.000Z

    Some innovative approaches have recently been proposed that may have significant schedule, cost, or environmental advantages which could improve the current HLW program strategy. Three general categories of alternative concepts are now under consideration: (1) process/product alternatives, (2) facility layout options, and (3) contracting strategies. This report compares the alternate approaches to the current program baseline to illustrate their potential significance and to identify the risks associated with each approach.

  5. Technology of high-level nuclear waste disposal. Advances in the science and engineering of the management of high-level nuclear wastes. Volume 1

    SciTech Connect (OSTI)

    Hofmann, P.L.; Breslin, J.J. (eds.)

    1981-01-01T23:59:59.000Z

    The papers in this volume cover the following subjects: waste isolation and the natural geohydrologic system; repository perturbations of the natural system; radionuclide migration through the natural system; and repository design technology. Individual papers are abstracted.

  6. Low-temperature lithium diffusion in simulated high-level boroaluminosilicate nuclear waste glasses

    SciTech Connect (OSTI)

    Neeway, James J.; Kerisit, Sebastien N.; Gin, Stephane; Wang, Zhaoying; Zhu, Zihua; Ryan, Joseph V.

    2014-12-01T23:59:59.000Z

    Ion exchange is recognized as an integral, if underrepresented, mechanism influencing glass corrosion. However, due to the formation of various alteration layers in the presence of water, it is difficult to conclusively deconvolute the mechanisms of ion exchange from other processes occurring simultaneously during corrosion. In this work, an operationally inert non-aqueous solution was used as an alkali source material to isolate ion exchange and study the solid-state diffusion of lithium. Specifically, the experiments involved contacting glass coupons relevant to the immobilization of high-level nuclear waste, SON68 and CJ-6, which contained Li in natural isotope abundance, with a non-aqueous solution of 6LiCl dissolved in dimethyl sulfoxide at 90 °C for various time periods. The depth profiles of major elements in the glass coupons were measured using time-of-flight secondary ion mass spectrometry (ToF-SIMS). Lithium interdiffusion coefficients, DLi, were then calculated based on the measured depth profiles. The results indicate that the penetration of 6Li is rapid in both glasses with the simplified CJ-6 glass (D6Li ? 4.0-8.0 × 10-21 m2/s) exhibiting faster exchange than the more complex SON68 glass (DLi ? 2.0-4.0 × 10-21 m2/s). Additionally, sodium ions present in the glass were observed to participate in ion exchange reactions; however, different diffusion coefficients were necessary to fit the diffusion profiles of the two alkali ions. Implications of the diffusion coefficients obtained in the absence of alteration layers to the long-term performance of nuclear waste glasses in a geological repository system are also discussed.

  7. Sedimentation behavior of noble metal particles in simulated high-level waste borosilicate glasses

    SciTech Connect (OSTI)

    Nakajima, M.; Ohyama, K.; Morikawa, Y.; Miyauchi, A.; Yamashita, T. [Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319-1109 (Japan); Komamine, S.; Ochi, E. [Japan Nuclear Fuel Limited, Bussan-Bldg. Bekkan, 1-1-5 Nishi-Shinbashi Minato-ku, Tokyo 105-0003 (Japan)

    2013-07-01T23:59:59.000Z

    Solubility of noble metal elements (NME) in the melted borosilicate glass is much smaller than its normal concentration of the high level liquid waste. Thus most of NME show small particles in the melted glass and tend to sediment in the bottom region of the vitrification melter due to their higher density than that of glass. Experiments of the sedimentation of NME particles in the melted glass were carried out under static condition. Three conditions of initial NME concentration (1.1, 3.0, 6.1 wt % with an equivalent for each oxide) in the simulated glass were set and held at 1100 C. degrees up to 2880 hours. The specimen with 1.1 wt % initial NME concentration indicated zone settling, and the settling rate of the interface is constant: 2.4 mm/h. This sedimentation behavior is the type of rapid settling. Following the rapid settling, the settling rate goes gradually slower; this is the type of compressive settling. The specimens with 3.0 wt % and 6.1 wt % initial NME concentration showed compression settling from the beginning. From the settling curve of the interface, the maximum concentration of NME in sediment was estimated to be around 23- 26 wt %. Growth of NME particles was observed by holding at 1100 C. degrees for up to 2880 hours. The viscosity becomes higher as NME concentration increases and the dependence on shear rate becomes simultaneously stronger. The effect of the particle growth to viscosity appears to be not significant.

  8. Oxidative Alkaline leaching of Americium from simulated high-level nuclear waste sludges

    SciTech Connect (OSTI)

    Reed, Wendy A.; Garnov, Alexander Yu.; Rao, Linfeng; Nash, Kenneth L.; Bond, Andrew H.

    2004-01-23T23:59:59.000Z

    Oxidative alkaline leaching has been proposed to pre-treat the high-level nuclear waste sludges to remove some of the problematic (e.g., Cr) and/or non-radioactive (e.g., Na, Al) constituents before vitrification. It is critical to understand the behavior of actinides, americium and plutonium in particular, in oxidative alkaline leaching. We have studied the leaching behavior of americium from four different sludge simulants (BiPO{sub 4}, BiPO{sub 4 modified}, Redox, PUREX) using potassium permanganate and potassium persulfate in alkaline solutions. Up to 60% of americium sorbed onto the simulants is leached from the sludges by alkaline persulfate and permanganate. The percentage of americium leached increases with [NaOH] (between 1.0 and 5.0 M). The initial rate of americium leaching by potassium persulfate increases in the order BiPO{sub 4} sludge < Redox sludge < PUREX sludge. The data are most consistent with oxidation of Am{sup 3+} in the sludge to either AmO{sub 2}{sup +} or AmO{sub 2}{sup 2+} in solution. Though neither of these species is expected to exhibit long-term stability in solution, the potential for mobilization of americium from sludge samples would have to be accommodated in the design of any oxidative leaching process for real sludge samples.

  9. Alternate approaches to verifying the structural adequacy of the Defense High Level Waste Shipping Cask

    SciTech Connect (OSTI)

    Zimmer, A.; Koploy, M.

    1991-12-01T23:59:59.000Z

    In the early 1980s, the US Department of Energy/Defense Programs (DOE/DP) initiated a project to develop a safe and efficient transportation system for defense high level waste (DHLW). A long-standing objective of the DHLW transportation project is to develop a truck cask that represents the leading edge of cask technology as well as one that fully complies with all applicable DOE, Nuclear Regulatory Commission (NRC), and Department of Transportation (DOT) regulations. General Atomics (GA) designed the DHLW Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Laboratories (SNL). The analytical techniques include two approaches, inelastic analysis and elastic analysis. This topical report presents the results of the two analytical approaches and the model testing results. The purpose of this work is to show that there are two viable analytical alternatives to verify the structural adequacy of a Type B package and to obtain an NRC license. It addition, this data will help to support the future acceptance by the NRC of inelastic analysis as a tool in packaging design and licensing.

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

    SciTech Connect (OSTI)

    Not Available

    1983-06-01T23:59:59.000Z

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

  11. Role of Congress in the High Level Radioactive Waste Odyssey: The Wisdom and Will of the Congress - 13096

    SciTech Connect (OSTI)

    Vieth, Donald L. [DOE/NVOO Project Manager for Yucca Mountain, 1982 thru 1987, 1154 Cheltenham Place, Maineville, OH 45039 (United States)] [DOE/NVOO Project Manager for Yucca Mountain, 1982 thru 1987, 1154 Cheltenham Place, Maineville, OH 45039 (United States); Voegele, Michael D. [Nye County Nuclear Waste Repository Project Office, 7404 Oak Grove Ave, Las Vegas, NV 89117 (United States)] [Nye County Nuclear Waste Repository Project Office, 7404 Oak Grove Ave, Las Vegas, NV 89117 (United States)

    2013-07-01T23:59:59.000Z

    Congress has had a dual role with regard to high level radioactive waste, being involved in both its creation and its disposal. A significant amount of time has passed between the creation of the nation's first high level radioactive waste and the present day. The pace of addressing its remediation has been highly irregular. Congress has had to consider the technical, regulatory, and political issues and all have had specific difficulties. It is a true odyssey framed by an imperative and accountability, by a sense of urgency, by an ability or inability to finish the job and by consequences. Congress had set a politically acceptable course by 1982. However, President Obama intervened in the process after he took office in January 2009. Through the efforts of his Administration, by the end of 2012, the US government has no program to dispose of high level radioactive waste and no reasonable prospect of a repository for high level radioactive waste. It is not obvious how the US government program will be reestablished or who will assume responsibility for leadership. The ultimate criteria for judging the consequences are 1) the outcome of the ongoing NRC's Nuclear Waste Confidence Rulemaking and 2) the concomitant permissibility of nuclear energy supplying electricity from operating reactors in the US. (authors)

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

    SciTech Connect (OSTI)

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

    2003-02-27T23:59:59.000Z

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

  13. Potential role of ABC-assisted repositories in U.S. plutonium and high-level waste disposition

    SciTech Connect (OSTI)

    Berwald, David; Favale, Anthony; Myers, Timothy; McDaniel, Jerry [Grumman Aerospace Corporation, Bethpage New York 11714 (United States); Bechtel Corporation, 50 Beal St., San Francisco, California 94105 (United States)

    1995-09-15T23:59:59.000Z

    This paper characterizes the issues involving deep geologic disposal of LWR spent fuel rods, then presents results of an investigation to quantify the potential role of Accelerator-Based Conversion (ABC) in an integrated national nuclear materials and high level waste disposition strategy. The investigation used the deep geological repository envisioned for Yucca Mt., Nevada as a baseline and considered complementary roles for integrated ABC transmutation systems. The results indicate that although a U.S. geologic waste repository will continue to be required, waste partitioning and accelerator transmutation of plutonium, the minor actinides, and selected long-lived fission products can result in the following substantial benefits: plutonium burndown to near zero levels, a dramatic reduction of the long term hazard associated with geologic repositories, an ability to place several-fold more high level nuclear waste in a single repository, electricity sales to compensate for capital and operating costs.

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

    SciTech Connect (OSTI)

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

    2007-07-01T23:59:59.000Z

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

  15. Technology of high-level nuclear waste disposal. Advances in the science and engineering of the management of high-level nuclear wastes. Volume 2

    SciTech Connect (OSTI)

    Hofmann, P.L. (ed.)

    1982-01-01T23:59:59.000Z

    The twenty papers in this volume are divided into three parts: site exploration and characterization; repository development and design; and waste package development and design. These papers represent the status of technology that existed in 1981 and 1982. Individual papers were processed for inclusion in the Energy Data Base.

  16. 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.

    1998-05-01T23:59:59.000Z

    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.

  17. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 4

    SciTech Connect (OSTI)

    Not Available

    1994-04-01T23:59:59.000Z

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 4) presents the standards and requirements for the following sections: Radiation Protection and Operations.

  18. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 2

    SciTech Connect (OSTI)

    Not Available

    1994-04-01T23:59:59.000Z

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Document (S/RID) is contained in multiple volumes. This document (Volume 2) presents the standards and requirements for the following sections: Quality Assurance, Training and Qualification, Emergency Planning and Preparedness, and Construction.

  19. Application of Quantitative NDE Techniques to High Level Waste Storage Tanks

    SciTech Connect (OSTI)

    Thompson, R. B.; Rehbein, D. K.; Bastiaans, G.; Terry, M.; Alers, R.

    2002-02-25T23:59:59.000Z

    As various issues make the continued usage of high-level waste storage tanks attractive, there is an increasing need to sharpen the assessment of their structural integrity. One aspect of a structural integrity program, nondestructive evaluation, is the focus of this paper. In September 2000, a program to support the sites was initiated jointly by Tanks Focus Area and Characterization, Monitoring, and Sensor Technologies Crosscutting Program of the Office of Environmental Management, Department of Energy (DOE). The vehicle was the Center for Nondestructive Evaluation, one of the National Science Foundation's Industry/University Cooperative Research Centers that is operated in close collaboration with the Ames Laboratory, USDOE. The support activities that have been provided by the center will be reviewed. Included are the organization of a series of annual workshops to allow the sites to share experiences and develop coordinated approaches to common problems, the development of an electronic source of relevant information, and assistance of the sites on particular technical problems. Directions and early results on some of these technical assistance projects are emphasized. Included are the discussion of theoretical analysis of ultrasonic wave propagation in curved plates to support the interpretation of tandem synthetic aperture focusing data to detect flaws in the knuckle region of double shell tanks; the evaluation of guided ultrasonic waves, excited by couplant free, electromagnetic acoustic transducers, to rapidly screen for inner wall corrosion in tanks; the use of spread spectrum techniques to gain information about the structural integrity of concrete domes; and the use of magnetic techniques to identify the alloys used in the construction of tanks.

  20. RIVER PROTECTION PROJECT MISSION ANALYSIS WASTE BLENDING STUDY

    SciTech Connect (OSTI)

    SHUFORD DH; STEGEN G

    2010-04-19T23:59:59.000Z

    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.

  1. Use of depleted uranium metal as cask shielding in high-level waste storage, transport, and disposal systems

    SciTech Connect (OSTI)

    Yoshimura, H.R.; Ludwigsen, J.S.; McAllaster, M.E. [and others

    1996-09-01T23:59:59.000Z

    The US DOE has amassed over 555,000 metric tons of depleted uranium from its uranium enrichment operations. Rather than dispose of this depleted uranium as waste, this study explores a beneficial use of depleted uranium as metal shielding in casks designed to contain canisters of vitrified high-level waste. Two high-level waste storage, transport, and disposal shielded cask systems are analyzed. The first system employs a shielded storage and disposal cask having a separate reusable transportation overpack. The second system employs a shielded combined storage, transport, and disposal cask. Conceptual cask designs that hold 1, 3, 4 and 7 high-level waste canisters are described for both systems. In all cases, cask design feasibility was established and analyses indicate that these casks meet applicable thermal, structural, shielding, and contact-handled requirements. Depleted uranium metal casting, fabrication, environmental, and radiation compatibility considerations are discussed and found to pose no serious implementation problems. About one-fourth of the depleted uranium inventory would be used to produce the casks required to store and dispose of the nearly 15,400 high-level waste canisters that would be produced. This study estimates the total-system cost for the preferred 7-canister storage and disposal configuration having a separate transportation overpack would be $6.3 billion. When credits are taken for depleted uranium disposal cost, a cost that would be avoided if depleted uranium were used as cask shielding material rather than disposed of as waste, total system net costs are between $3.8 billion and $5.5 billion.

  2. 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

    2011-09-01T23:59:59.000Z

    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.

  3. ALPHN: A computer program for calculating ([alpha], n) neutron production in canisters of high-level waste

    SciTech Connect (OSTI)

    Salmon, R.; Hermann, O.W.

    1992-10-01T23:59:59.000Z

    The rate of neutron production from ([alpha], n) reactions in canisters of immobilized high-level waste containing borosilicate glass or glass-ceramic compositions is significant and must be considered when estimating neutron shielding requirements. The personal computer program ALPHA calculates the ([alpha], n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the ([alpha], n) neutron production of each actinide in neutrons per second and the total for the canister. The ([alpha], n) neutron production rates are source terms only; that is, they are production rates within the glass and do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister. In a typical application, these cases might represent the same canister of vitrified high-level waste at eight different decay times. Run time for a typical problem containing 20 chemical species, 24 actinides, and 8 decay times was 35 s on an IBM AT personal computer. Results of an example based on an expected canister composition at the Defense Waste Processing Facility are shown.

  4. ALPHN: A computer program for calculating ({alpha}, n) neutron production in canisters of high-level waste

    SciTech Connect (OSTI)

    Salmon, R.; Hermann, O.W.

    1992-10-01T23:59:59.000Z

    The rate of neutron production from ({alpha}, n) reactions in canisters of immobilized high-level waste containing borosilicate glass or glass-ceramic compositions is significant and must be considered when estimating neutron shielding requirements. The personal computer program ALPHA calculates the ({alpha}, n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the ({alpha}, n) neutron production of each actinide in neutrons per second and the total for the canister. The ({alpha}, n) neutron production rates are source terms only; that is, they are production rates within the glass and do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister. In a typical application, these cases might represent the same canister of vitrified high-level waste at eight different decay times. Run time for a typical problem containing 20 chemical species, 24 actinides, and 8 decay times was 35 s on an IBM AT personal computer. Results of an example based on an expected canister composition at the Defense Waste Processing Facility are shown.

  5. Design and operating features of the high-level waste vitrification system for the West Valley demonstration project

    SciTech Connect (OSTI)

    Siemens, D.H.; Beary, M.M.; Barnes, S.M.; Berger, D.N.; Brouns, R.A.; Chapman, C.C.; Jones, R.M.; Peters, R.D.; Peterson, M.E.

    1986-03-01T23:59:59.000Z

    A liquid-fed joule-heated ceramic melter system is the reference process for immobilization of the high-level liquid waste in the US and several foreign countries. This system has been under development for over ten years at Pacific Northwest Laboratory and other national laboratories operated for the US Department of Energy. Pacific Northwest Laboratory contributed to this research through its Nuclear Waste Treatment Program and used applicable data to design and test melters and related systems using remote handling of simulated radioactive wastes. This report describes the equipment designed in support of the high-level waste vitrification program at West Valley, New York. Pacific Northwest Laboratory worked closely with West Valley Nuclear Services Company to design a liquid-fed ceramic melter, a liquid waste preparation and feed tank and pump, an off-gas treatment scrubber, and an enclosed turntable for positioning the waste canisters. Details of these designs are presented including the rationale for the design features and the alternatives considered.

  6. A postmortem assessment of environmental compliance of a high-level radioactive waste repository, Hanford Site, Washington

    E-Print Network [OSTI]

    Petrini, Rudolf Harald Wilhelm

    1988-01-01T23:59:59.000Z

    the engineered barrier and the accessible environment. The concept of geochemical retarda'tion has been analyzed by Domenico et al. (1988) from a regulatory point of view and the following discussion is a summary of their work. As discussed previously, a...A POSTMORTEM ASSESSMENT OF ENVIRONMENTAL COMPLIANCE OF A HIGH-LEVEL RADIOACTIVE WASTE REPOSITORY, HANFORD SITE, WASHINGTON A Thesis by RUDOLF HARALD WILHELM PETRINI Submitted to the Graduate College of Texas A & M University in partial...

  7. 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.

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

    SciTech Connect (OSTI)

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

    2014-05-31T23:59:59.000Z

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

  9. RESORCINOL-FORMALDEHYDE ION EXCHANGE RESIN CHEMISTRY FOR HIGH LEVEL WASTE TREATMENT

    SciTech Connect (OSTI)

    Nash, C.; Duignan, M.

    2010-01-14T23:59:59.000Z

    A principal goal at the Savannah River Site is to safely dispose of the large volume of liquid nuclear waste held in many storage tanks. In-tank ion exchange technology is being considered for cesium removal using a polymer resin made of resorcinol formaldehyde that has been engineered into microspheres. The waste under study is generally lower in potassium and organic components than Hanford waste; therefore, the resin performance was evaluated with actual dissolved salt waste. The ion exchange performance and resin chemistry results are discussed.

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

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

    -5 . No offsite transportation would occur. At INEEL - The estimated number of latent cancer fatalities in the population within 50 miles of INTEC related to waste processing...

  11. Thermal and Radiolytic Gas Generation in Hanford High-Level Waste

    SciTech Connect (OSTI)

    Bryan, Samuel A.; Pederson, Larry R.; King, C. M.

    2000-01-31T23:59:59.000Z

    The Hanford Site has 177 underground storage tanks containing radioactive wastes that are complex mixes of radioactive and chemical products. Some of these wastes are known to generate and retain large quantities of flammable gases consisting of hydrogen, nitrous oxide, nitrogen, and ammonia. Because these gases are flammable and have the potential for rapid release, the gas generation rate for each tank must be determined to establish the flammability hazard (Johnson et al. 1997). An understanding of gas generation is important to operation of the waste tanks for several reasons. First, knowledge of the overall rate of generation is needed to verify that any given tank has sufficient ventilation to ensure that flammable gases are maintained at a safe level within the dome space. Understanding the mechanisms for production of the various gases is important so that future waste operations do not create conditions that promote the production of hydrogen, ammonia, and nitrous oxide. Studying the generation of gases also provides important data for the composition of the gas mixture, which in turn is needed to assess the flammability characteristics. Finally, information about generation of gases, including the influence of various chemical constituents, temperature, and dose, would aid in assessing the future behavior of the waste during interim storage, implementation of controls, and final waste treatment. This paper summarizes the current knowledge of gas generation pathways and discusses models used in predicting gas generation rates from actual Hanford radioactive wastes. A comparison is made between measured gas generation rates and rates by the predictive models.

  12. Assessment of chemical vulnerabilities in the Hanford high-level waste tanks

    SciTech Connect (OSTI)

    Meacham, J.E. [and others

    1996-02-15T23:59:59.000Z

    The purpose of this report is to summarize results of relevant data (tank farm and laboratory) and analysis related to potential chemical vulnerabilities of the Hanford Site waste tanks. Potential chemical safety vulnerabilities examined include spontaneous runaway reactions, condensed phase waste combustibility, and tank headspace flammability. The major conclusions of the report are the following: Spontaneous runaway reactions are not credible; condensed phase combustion is not likely; and periodic releases of flammable gas can be mitigated by interim stabilization.

  13. Building the institutional capacity for managing commercial high-level radioactive waste

    SciTech Connect (OSTI)

    None

    1982-05-01T23:59:59.000Z

    In July 1981, the Office of Nuclear Waste Management of the Department of Energy contracted with the National Academy of Public Administration for a study of institutional issues associated with the commercial radioactive waste management program. The two major sets of issues which the Academy was asked to investigate were (1) intergovernmental relationships, how federal, state, local and Indian tribal council governments relate to each other in the planning and implementation of a waste management program, and (2) interagency relationships, how the federal agencies with major responsibilities in this public policy arena interact with each other. The objective of the study was to apply the perspectives of public administration to a difficult and controversial question - how to devise and execute an effective waste management program workable within the constraints of the federal system. To carry out this task, the Academy appointed a panel composed of individuals whose background and experience would provide the several types of knowledge essential to the effort. The findings of this panel are presented along with the executive summary. The report consists of a discussion of the search for a radioactive waste management strategy, and an analysis of the two major groups of institutional issues: (1) intergovernmental, the relationship between the three major levels of government; and (2) interagency, the relationships between the major federal agencies having responsibility for the waste management program.

  14. CHEMICAL ANALYSIS OF SIMULATED HIGH LEVEL WASTE GLASSES TO SUPPORT SULFATE SOLUBILITY MODELING

    SciTech Connect (OSTI)

    Fox, K.; Marra, J.

    2014-08-14T23:59:59.000Z

    The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms both within the DOE complex and to some extent at U.K. sites. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerated cleanup missions. Much of the previous work on improving sulfate retention in waste glasses has been done on an empirical basis, making it difficult to apply the findings to future waste compositions despite the large number of glass systems studied. A more fundamental, rather than empirical, model of sulfate solubility in glass, under development at Sheffield Hallam University (SHU), could provide a solution to the issues of sulfate solubility. The model uses the normalized cation field strength index as a function of glass composition to predict sulfate capacity, and has shown early success for some glass systems. The objective of the current scope is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DOE waste vitrification efforts, allowing for enhanced waste loadings and waste throughput. A series of targeted glass compositions was selected to resolve data gaps in the current model. SHU fabricated these glasses and sent samples to the Savannah River National Laboratory (SRNL) for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for simulated waste glasses fabricated SHU in support of sulfate solubility model development. A review of the measured compositions revealed that there are issues with the B{sub 2}O{sub 3} and Fe{sub 2}O{sub 3} concentrations missing their targeted values by a significant amount for several of the study glasses. SHU is reviewing the fabrication of these glasses and the chemicals used in batching them to identify the source of these issues. The measured sulfate concentrations were all below their targeted values. This is expected, as the targeted concentrations likely exceeded the solubility limit for sulfate in these glass compositions. Some volatilization of sulfate may also have occurred during fabrication of the glasses. Measurements of the other oxides in the study glasses were reasonably close to their targeted values

  15. Conditioning matrices from high level waste resulting from pyrochemical processing in fluorine salt

    SciTech Connect (OSTI)

    Grandjean, Agnes; Advocat, Thierry; Bousquet, Nicolas [SCDV - Service de Confinement des Dechets et Vitrification - Laboratoire d'Etudes de Base sur les Verres, CEA Valrho, Centre de Marcoule, 30207 Bagnols sur Ceze (France); Jegou, Christophe [SECM - Service d'Etude du Confinement et Materiaux - Laboratoire des Materiaux et Procedes Actifs - CEA Valrho, Centre de Marcoule, 30207 Bagnols sur Ceze (France)

    2007-07-01T23:59:59.000Z

    Separating the actinides from the fission products through reductive extraction by aluminium in a LiF/AlF{sub 3} medium is a process investigated for pyrometallurgical reprocessing of spent fuel. The process involves separation by reductive salt-metal extraction. After dissolving the fuel or the transmutation target in a salt bath, the noble metal fission products are first extracted by contacting them with a slightly reducing metal. After extracting the metal fission products, then the actinides are selectively separated from the remaining fission products. In this hypothesis, all the unrecoverable fission products would be conditioned as fluorides. Therefore, this process will generate first a metallic waste containing the 'reducible' fission products (Pd, Mo, Ru, Rh, Tc, etc.) and a fluorine waste containing alkali-metal, alkaline-earth and rare earth fission products. Immobilization of these wastes in classical borosilicate glasses is not feasible due to the very low solubility of noble metals, and of fluoride in these hosts. Alternative candidates have therefore been developed including silicate glass/ceramic system for fluoride fission products and metallic ones for noble metal fission products. These waste-forms were evaluated for their confinement properties like homogeneity, waste loading, volatility during the elaboration process, chemical durability, etc. using appropriate techniques. (authors)

  16. COMBINED RETENTION OF MOLYBDENUM AND SULFUR IN SIMULATED HIGH LEVEL WASTE GLASS

    SciTech Connect (OSTI)

    Fox, K.

    2009-10-16T23:59:59.000Z

    This study was undertaken to investigate the effect of elevated sulfate and molybdenum concentrations in nuclear waste glasses. A matrix of 24 glasses was developed and the glasses were tested for acceptability based on visual observations, canister centerline-cooled heat treatments, and chemical composition analysis. Results from the chemical analysis of the rinse water from each sample were used to confirm the presence of SO{sup 2-}{sub 4} and MoO{sub 3} on the surface of glasses as well as other components which might form water soluble compounds with the excess sulfur and molybdenum. A simple, linear model was developed to show acceptable concentrations of SO{sub 4}{sup 2-} and MoO{sub 3} in an example waste glass composition. This model was constructed for scoping studies only and is not ready for implementation in support of actual waste vitrification. Several other factors must be considered in determining the limits of sulfate and molybdenum concentrations in the waste vitrification process, including but not limited to, impacts on refractory and melter component corrosion, effects on the melter off-gas system, and impacts on the chemical durability and crystallization of the glass product.

  17. Comments on a paper tilted `The sea transport of vitrified high-level radioactive wastes: Unresolved safety issues`

    SciTech Connect (OSTI)

    Sprung, J.L.; McConnell, P.E.; Nigrey, P.J.; Ammerman, D.J. [and others

    1997-05-01T23:59:59.000Z

    The cited paper estimates the consequences that might occur should a purpose-built ship transporting Vitrified High Level Waste (VHLW) be involved in a severe collision that causes the VHLW canisters in one Type-B package to spill onto the floor of a major ocean fishing region. Release of radioactivity from VHLW glass logs, failure of elastomer cask seals, failure of VHLW canisters due to stress corrosion cracking (SCC), and the probabilities of the hypothesized accident scenario, of catastrophic cask failure, and of cask recovery from the sea are all discussed.

  18. Development of integraded mechanistically-based degradation-mode models for performance assessment of high-level waste containers

    SciTech Connect (OSTI)

    Farmer, J. C., LLNL

    1998-06-01T23:59:59.000Z

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological repository at Yucca Mountain, Nevada is a two-tayer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A516 Gr 55 or Monel 400. At the present time, Alloy C- 22 and A516 Gr 55 are favored.

  19. 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.

    2013-11-13T23:59:59.000Z

    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.

  20. Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results

    SciTech Connect (OSTI)

    Rechard, R.P. [ed.

    1995-03-01T23:59:59.000Z

    This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM.

  1. Functions and requirements document for interim store solidified high-level and transuranic waste

    SciTech Connect (OSTI)

    Smith-Fewell, M.A., Westinghouse Hanford

    1996-05-17T23:59:59.000Z

    The functions, requirements, interfaces, and architectures contained within the Functions and Requirements (F{ampersand}R) Document are based on the information currently contained within the TWRS Functions and Requirements database. The database also documents the set of technically defensible functions and requirements associated with the solidified waste interim storage mission.The F{ampersand}R Document provides a snapshot in time of the technical baseline for the project. The F{ampersand}R document is the product of functional analysis, requirements allocation and architectural structure definition. The technical baseline described in this document is traceable to the TWRS function 4.2.4.1, Interim Store Solidified Waste, and its related requirements, architecture, and interfaces.

  2. Probability, consequences, and mitigation for lightning strikes of Hanford high level waste tanks

    SciTech Connect (OSTI)

    Zach, J.J.

    1996-06-05T23:59:59.000Z

    The purpose of this report is to summarize selected lightning issues concerning the Hanford Waste Tanks. These issues include the probability of a lightning discharge striking the area immediately adjacent to a tank including a riser, the consequences of significant energy deposition from a lightning strike in a tank, and mitigating actions that have been or are being taken. The major conclusion of this report is that the probability of a lightning strike deposition sufficient energy in a tank to cause an effect on employees or the public is unlikely;but there are insufficient, quantitative data on the tanks and waste to prove that. Protection, such as grounding of risers and air terminals on existing light poles, is recommended.

  3. Probability, consequences, and mitigation for lightning strikes to Hanford site high-level waste tanks

    SciTech Connect (OSTI)

    Zach, J.J.

    1996-08-01T23:59:59.000Z

    The purpose of this report is to summarize selected lightning issues concerning the Hanford Waste Tanks. These issues include the probability of lightning discharge striking the area immediately adjacent to a tank including a riser, the consequences of significant energy deposition from a lightning strike in a tank, and mitigating actions that have been or are being taken. The major conclusion of this report is that the probability of a lightning strike depositing sufficient energy in a tank to cause an effect on employees or the public is unlikely;but there are insufficient, quantitative data on the tanks and waste to prove that. Protection, such as grounding of risers and air terminals on existing light poles, is recommended.

  4. Materials performance in a high-level radioactive waste vitrification system

    SciTech Connect (OSTI)

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

    1996-06-17T23:59:59.000Z

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

  5. Probabilistic safety assessment for Hanford high-level waste tank 241-SY-101

    SciTech Connect (OSTI)

    MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W. [Los Alamos National Lab., NM (United States)] [Los Alamos National Lab., NM (United States); Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J. [PLG, Inc., Newport Beach, CA (United States)] [PLG, Inc., Newport Beach, CA (United States)

    1994-05-01T23:59:59.000Z

    Los Alamos National Laboratory (Los Alamos) is performing a comprehensive probabilistic safety assessment (PSA), which will include consideration of external events for the 18 tank farms at the Hanford Site. This effort is sponsored by the Department of Energy (DOE/EM, EM-36). Even though the methodology described herein will be applied to the entire tank farm, this report focuses only on the risk from the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases ({open_quotes}burps{close_quotes}) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed first because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is being conducted in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. At the Hanford Site there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/salt cake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is about 60 million gal., which contains approximately 120 million Ci of radioactivity.

  6. Laboratory Report on Performance Evaluation of Key Constituents during Pre-Treatment of High Level Waste Direct Feed

    SciTech Connect (OSTI)

    Huber, Heinz J.

    2013-06-24T23:59:59.000Z

    The analytical capabilities of the 222-S Laboratory are tested against the requirements for an optional start up scenario of the Waste Treatment and Immobilization Plant on the Hanford Site. In this case, washed and in-tank leached sludge would be sent directly to the High Level Melter, bypassing Pretreatment. The sludge samples would need to be analyzed for certain key constituents in terms identifying melter-related issues and adjustment needs. The analyses on original tank waste as well as on washed and leached material were performed using five sludge samples from tanks 241-AY-102, 241-AZ-102, 241-AN-106, 241-AW-105, and 241-SY-102. Additionally, solid phase characterization was applied to determine the changes in mineralogy throughout the pre-treatment steps.

  7. Preliminary total-system analysis of a potential high-level nuclear waste repository at Yucca Mountain

    SciTech Connect (OSTI)

    Eslinger, P.W.; Doremus, L.A.; Engel, D.W.; Miley, T.B.; Murphy, M.T.; Nichols, W.E.; White, M.D. [Pacific Northwest Lab., Richland, WA (United States); Langford, D.W.; Ouderkirk, S.J. [Westinghouse Hanford Co., Richland, WA (United States)

    1993-01-01T23:59:59.000Z

    The placement of high-level radioactive wastes in mined repositories deep underground is considered a disposal method that would effectively isolate these wastes from the environment for long periods of time. This report describes modeling performed at PNL for Yucca Mountain between May and November 1991 addressing the performance of the entire repository system related to regulatory criteria established by the EPA in 40 CFR Part 191. The geologic stratigraphy and material properties used in this study were chosen in cooperation with performance assessment modelers at Sandia National Laboratories (SNL). Sandia modeled a similar problem using different computer codes and a different modeling philosophy. Pacific Northwest Laboratory performed a few model runs with very complex models, and SNL performed many runs with much simpler (abstracted) models.

  8. Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 2: Appendices

    SciTech Connect (OSTI)

    Rechard, R.P. [ed.

    1993-12-01T23:59:59.000Z

    This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency`s Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

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

    SciTech Connect (OSTI)

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

    2012-03-07T23:59:59.000Z

    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. Noble Metals and Spinel Settling in High Level Waste Glass Melters

    SciTech Connect (OSTI)

    Sundaram, S. K.; Perez, Joseph M.

    2000-09-30T23:59:59.000Z

    In the continuing effort to support the Defense Waste Processing Facility (DWPF), the noble metals issue is addressed. There is an additional concern about the amount of noble metals expected to be present in the future batches that will be considered for vitrification in the DWPF. Several laboratory, as well as melter-scale, studies have been completed by various organizations (mainly PNNL, SRTC, and WVDP in the USA). This letter report statuses the noble metals issue and focuses at the settling of noble metals in melters.

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

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion |Energyon ArmedWaste andAccess to OUO Access toEnergy 5 BTOofthe UnitedandContinentaland

  12. Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository

    SciTech Connect (OSTI)

    K.G. Mon; F. Hua

    2005-04-12T23:59:59.000Z

    This paper reviews the state-of-the-art understanding of the degradation processes by the Yucca Mountain Project (YMP) 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 first 10,000-years after repository closure. This paper provides an overview of the degradation of the waste packages and drip shields in the repository after permanent closure of the facility. The degradation modes 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 the degradation of Alloy 22 and titanium alloys are also discussed. Further, for titanium alloys, the effects of fluorides, bromides, and galvanic coupling to less noble metals are considered. It is concluded that the materials and design adopted will provide sufficient safety margins for at least 10,000-years after repository closure.

  13. INTEC High-Level Waste Studies Universal Solvent Extraction Feasibility Study

    SciTech Connect (OSTI)

    J. Banaee; C. M. Barnes; T. Battisti (ANL-W) [ANL-W; S. Herrmann (ANL-W) [ANL-W; S. J. Losinski; S. McBride (ANL-W) [ANL-W

    2000-09-01T23:59:59.000Z

    This report summarizes a feasibility study that has been conducted on the Universal Solvent Extraction (UNEX) Process for treatment and disposal of 4.3 million liters of INEEL sodium-bearing waste located at the Idaho Nuclear Technology and Engineering Center. This feasibility study covers two scenarios of treatment. The first, the UNEX Process, partitions the Cs/Sr from the SBW and creates remote-handled LLW and contact-handled TRU waste forms. Phase one of this study, covered in the 30% review documents, dealt with defining the processes and defining the major unit operations. The second phase of the project, contained in the 60% review, expanded on the application of the UNEX processes and included facility requirements and definitions. Two facility options were investigated for the UNEX process, resulting in a 2 x 2 matrix of process/facility scenarios as follows: Option A, UNEX at Greenfield Facility, Option B, Modified UNEX at Greenfield Facility, Option C, UNEX at NWCF, th is document, covers life-cycle costs for all options presented along with results and conclusions determined from the study.

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

    SciTech Connect (OSTI)

    S.M. Frank

    2011-09-01T23:59:59.000Z

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

  15. HLW Glass Waste Loadings

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG | Department of EnergyGeothermalGoingGuidelines forofHCHEFA andHI

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

    SciTech Connect (OSTI)

    Vinson, D.W.; Nutt, W.M.; Bullen, D.B. [Iowa State Univ. of Science and Technology, Ames, IA (United States)

    1995-06-01T23:59:59.000Z

    Oxidation and atmospheric corrosion data suggest that addition of Cr provides the greatest improvement in oxidation resistance. Cr-bearing cast irons are resistant to chloride environments and solutions containing strongly oxidizing constituents. Weathering steels, including high content and at least 0.04% Cu, appear to provide adequate resistance to oxidation under temperate conditions. However, data from long-term, high-temperature oxidation studies on weathering steels were not available. From the literature, it appears that the low alloy steels, plain carbon steels, cast steels, and cast irons con-ode at similar rates in an aqueous environment. Alloys containing more than 12% Cr or 36% Ni corrode at a lower rate than plain carbon steels, but pitting may be worse. Short term tests indicate that an alloy of 9Cr-1Mo may result in increased corrosion resistance, however long term data are not available. Austenitic cast irons show the best corrosion resistance. A ranking of total corrosion performance of the materials from most corrosion resistant to least corrosion resistant is: Austenitic Cast Iron; 12% Cr = 36% Ni = 9Cr-1Mo; Carbon Steel = Low Alloy Steels; and Cast Iron. Since the materials to be employed in the Advanced Conceptual Design (ACD) waste package are considered to be corrosion allowance materials, the austenitic cast irons, high Cr steels, high Ni steels and the high Cr-Mo steels should not be considered as candidates for the outer containment barrier. Based upon the oxidation and corrosion data available for carbon steels, low alloy steels, and cast irons, a suitable list of candidate materials for a corrosion allowance outer barrier for an ACD waste package could include, A516, 2.25%Cr -- 1%Mo Steel, and A27.

  17. Selection of candidate container materials for the conceptual waste package design for a potential high level nuclear waste repository at Yucca Mountain

    SciTech Connect (OSTI)

    Van Konynenburg, R.A.; Halsey, W.G.; McCright, R.D.; Clarke, W.L. Jr. [Lawrence Livermore National Lab., CA (United States)] [Lawrence Livermore National Lab., CA (United States); Gdowski, G.E. [KMI, Inc., Albuquerque, NM (United States)] [KMI, Inc., Albuquerque, NM (United States)

    1993-02-01T23:59:59.000Z

    Preliminary selection criteria have been developed, peer-reviewed, and applied to a field of 41 candidate materials to choose three alloys for further consideration during the advanced conceptual design phase of waste package development for a potential high level nuclear waste repository at Yucca Mountain, Nevada. These three alloys are titanium grade 12, Alloy C-4, and Alloy 825. These selections are specific to the particular conceptual design outlined in the Site Characterization Plan. Other design concepts that may be considered in the advanced conceptual design phase may favor other materials choices.

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

    SciTech Connect (OSTI)

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

    2012-11-13T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1994-04-01T23:59:59.000Z

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

  20. Seismic design and evaluation guidelines for the Department of Energy high-level waste storage tanks and appurtenances

    SciTech Connect (OSTI)

    Bandyopadhyay, K.; Cornell, A.; Costantino, C.; Kennedy, R.; Miller, C.; Veletsos, A.

    1993-01-01T23:59:59.000Z

    This document provides guidelines for the design and evaluation of underground high-level waste storage tanks due to seismic loads. Attempts were made to reflect the knowledge acquired in the last two decades in the areas of defining the ground motion and calculating hydrodynamic loads and dynamic soil pressures for underground tank structures. The application of the analysis approach is illustrated with an example. The guidelines are developed for specific design of underground storage tanks, namely double-shell structures. However, the methodology discussed is applicable for other types of tank structures as well. The application of these and of suitably adjusted versions of these concepts to other structural types will be addressed in a future version of this document.

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

    SciTech Connect (OSTI)

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

    2013-07-01T23:59:59.000Z

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

  2. EIS-0081: Long-Term Management of Liquid High-Level Radioactive Waste Stored at Western New York Nuclear Service Center, West Valley, New York

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy Office of Terminal Waste Disposal and Remedial Action prepared this statement to analyze the environmental and socioeconomic impacts resulting from the Department’s proposed action to construct and operate facilities necessary to solidify the liquid high level wastes currently stored in underground tanks at Wes t Valley, New York.

  3. Environmental evaluation of alternatives for long-term management of Defense high-level radioactive wastes at the Idaho Chemical Processing Plant

    SciTech Connect (OSTI)

    Not Available

    1982-09-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) is considering the selection of a strategy for the long-term management of the defense high-level wastes at the Idaho Chemical Processing Plant (ICPP). This report describes the environmental impacts of alternative strategies. These alternative strategies include leaving the calcine in its present form at the Idaho National Engineering Laboratory (INEL), or retrieving and modifying the calcine to a more durable waste form and disposing of it either at the INEL or in an offsite repository. This report addresses only the alternatives for a program to manage the high-level waste generated at the ICPP. 24 figures, 60 tables.

  4. US DOE-AECL cooperative program for development of high-level radioactive waste container fabrication, closure, and inspection techniques

    SciTech Connect (OSTI)

    Russell, E.W.

    1990-06-01T23:59:59.000Z

    The US Department of Energy (DOE) and Atomic Energy of Canada Limited (AECL) plan to initiate a cooperative research program on development of manufacturing processes for high-level radioactive waste containers. This joint program will benefit both countries in the development of processes for the fabrication, final closure in a hot-cell, and certification of the containers. Program activity objectives can be summarized as follows: to support the selection of suitable container fabrication, final closure, and inspection techniques for the candidate materials and container designs that are under development or are being considered in the US and Canadian repository programs; and to investigate these techniques for alternate materials and/or container designs, to be determined in future optimization studies relating to long-term performance of the waste packages. The program participants will carry out this work in a conditional phased approach, and the scope of work for subsequent years will evolve subject to developments in earlier years. The overall term of this cooperative program is planned to run roughly three years. 5 refs., 2 tabs.

  5. EM-21 HIGHER WASTE LOADING GLASSES FOR ENHANCED DOE HIGH-LEVEL WASTE MELTER THROUGHPUT STUDIES - 10194

    SciTech Connect (OSTI)

    Raszewski, F.; Peeler, D.; Edwards, T.

    2009-11-18T23:59:59.000Z

    Supplemental validation data has been generated that will be used to determine the applicability of the current Defense Waste Processing Facility (DWPF) liquidus temperature (T{sub L}) model to expanded DWPF glass regions of interest based on higher waste loadings. For those study glasses which had very close compositional overlap with the model development and/or model validation ranges (except TiO{sub 2} and MgO concentrations), there was very little difference in the predicted and measured TL values, even though the TiO{sub 2} contents were above the 2 wt% upper limit. The results indicate that the current T{sub L} model is applicable in these compositional regions. As the compositional overlap between the model validation ranges diverged from the target glass compositions, the T{sub L} data suggest that the model under-predicted the measured values. These discrepancies imply that there are individual oxides or their combinations that were outside of the model development and/or validation range over which the model was previously assessed. These oxides include B{sub 2}O{sub 3}, SiO{sub 2}, MnO, TiO{sub 2} and/or their combinations. More data is required to fill in these anticipated DWPF compositional regions so that the model coefficients could be refit to account for these differences.

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

    SciTech Connect (OSTI)

    Veronica J Rutledge; Vince Maio

    2013-10-01T23:59:59.000Z

    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.

  7. 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.

    2013-05-31T23:59:59.000Z

    The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the 56 million gallons of radioactive waste stored in 177 underground tanks at the Hanford Site. The WTP includes a pretreatment facility to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions for vitrification and disposal. The LAW will be converted to glass for final disposal at the Integrated Disposal Facility (IDF). 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.

  8. Fundamental properties of monolithic bentonite buffer material formed by cold isostatic pressing for high-level radioactive waste repository

    SciTech Connect (OSTI)

    Kawakami, S.; Yamanaka, Y.; Kato, K.; Asano, H.; Ueda, H.

    1999-07-01T23:59:59.000Z

    The methods of fabrication, handling, and emplacement of engineered barriers used in a deep geological repository for high level radioactive waste should be planned as simply as possible from the engineering and economic viewpoints. Therefore, a new concept of a monolithic buffer material around a waste package have been proposed instead of the conventional concept with the use of small blocks, which would decrease the cost for buffer material. The monolithic buffer material is composed of two parts of highly compacted bentonite, a cup type body and a cover. As the forming method of the monolithic buffer material, compaction by the cold isostatic pressing process (CIP) has been employed. In this study, monolithic bentonite bodies with the diameter of about 333 mm and the height of about 455 mm (corresponding to the approx. 1/5 scale for the Japanese reference concept) were made by the CIP of bentonite powder. The dry densities: {rho}d of the bodies as a whole were measured and the small samples were cut from several locations to investigate the density distribution. The swelling pressure and hydraulic conductivity as function of the monolithic body density for CIP-formed specimens were also measured. High density ({rho}d: 1.4--2.0 Mg/m{sup 3}) and homogeneous monolithic bodies were formed by the CIP. The measured results of the swelling pressure (3--15 MPa) and hydraulic conductivity (0.5--1.4 x 10{sup {minus}13} m/s) of the specimens were almost the same as those for the uniaxial compacted bentonite in the literature. It is shown that the vacuum hoist system is an applicable handling method for emplacement of the monolithic bentonite.

  9. Self-irradiation damage of a curium-doped titanate ceramic containing sodium-rich high level nuclear waste

    SciTech Connect (OSTI)

    Miyazaki, T. (Second Dept. of Nuclear Business, Ibaraki Center, Chiyoda Maintenance Ltd., Asahi, Kashima, Ibaraki 314-14 (JP)); White, T.J. (Electron Microscope Centre, Univ. of Queensland, St. Lucia, Brisbane 4067 (AU)); Mitamura, H.; Matsumoto, S.; Nukaga, K.; Togashi, Y.; Sagawa, T.; Tashiro, S. (Dept. of Environmental Safety Research, Japan Atomic Energy Research Inst., Tokai, Naka, Ibaraki 319-11 (JP)); Levins, D.M. (Environmental Science Program, Australian Nuclear Science and Technology Organisation, Lucas Heights Research Lab., Lucas Heights, New South Wales (AU)); Kikuchi, A. (Dept. of Reactor Fuel Examination, Japan Atomic Energy Research Inst., Tokai, Naka, Ibaraki 319-11 (JP))

    1990-11-01T23:59:59.000Z

    This paper reports on the polyphase titanate ceramic containing sodium-rich simulated high-level nuclear waste doped with 0.69 wt% of {sup 244}Cm to accelerate long-term self-irradiation due to {alpha} decays. {alpha} autoradiography showed that {alpha} emissions were almost uniformly distributed throughout the curium-doped samples on a {gt} 20{mu}m scale although micropore surfaces and titanium oxide agglomerates were free of {alpha} emitting nuclides. The phase assemblage of the curium-doped titanate ceramic included freudenbergite and loveringite in addition to the more abundant oxide phases: hollandite, perovskite, and zirconolite. Accumulation of {alpha} decays was accompanied by a gradual decrease in density. The increment of density was {minus}1% after an equivalent age of 5000 yr. Leach tests showed a slight rend toward higher total release of curium with equivalent age. The release of soluble nonradioactive elements (e.g., Na, Cs, Sr, and Ca) in the oldest specimens (equivalent age, 2000 yr) varied from specimen to specimen but, on average, were higher than specimens that had suffered a lower radiation dose.

  10. Application of single ion activity coefficients to determine solvent extraction mechanism for components of high level nuclear waste

    SciTech Connect (OSTI)

    Nunez, L.; Vandegrift, G.F.

    1995-12-31T23:59:59.000Z

    The TRUEX solvent extraction process is being developed to remove and concentrate transuranic (TRU) elements from high-level and TRU radioactive wastes currently stored at US Department of Energy sites. Phosphoric acid is one of the chemical species of concern at the Hanford site where bismuth phosphate was used to recover plutonium. The mechanism of phosphoric acid extraction with TRUEX-NPH solvent at 25{degrees}C was determined by phosphoric acid distribution ratios, which were measured by using phosphoric acid radiotracer and a variety of aqueous phases containing different concentrations of nitric acid and nitrate ions. A model was developed for predicting phosphoric acid distribution ratios as a function of the thermodynamic activities of nitrate ion and hydrogen ion. The Generic TRUEX Model (GTM) was used to calculate these activities based on the Bromley method. The derived model supports CMPO and TBP extraction of a phosphoric acid-nitric acid complex and a CMPO-phosphoric acid complex in TRUEX-NPH solvent.

  11. Seismic design and evaluation guidelines for the Department of Energy High-Level Waste Storage Tanks and Appurtenances

    SciTech Connect (OSTI)

    Bandyopadhyay, K.; Cornell, A.; Costantino, C.; Kennedy, R.; Miller, C.; Veletsos, A.

    1995-10-01T23:59:59.000Z

    This document provides seismic design and evaluation guidelines for underground high-level waste storage tanks. The guidelines reflect the knowledge acquired in the last two decades in defining seismic ground motion and calculating hydrodynamic loads, dynamic soil pressures and other loads for underground tank structures, piping and equipment. The application of the guidelines is illustrated with examples. The guidelines are developed for a specific design of underground storage tanks, namely double-shell structures. However, the methodology discussed is applicable for other types of tank structures as well. The application of these and of suitably adjusted versions of these concepts to other structural types will be addressed in a future version of this document. The original version of this document was published in January 1993. Since then, additional studies have been performed in several areas and the results are included in this revision. Comments received from the users are also addressed. Fundamental concepts supporting the basic seismic criteria contained in the original version have since then been incorporated and published in DOE-STD-1020-94 and its technical basis documents. This information has been deleted in the current revision.

  12. Transportation of Spent Nuclear Fuel and High Level Waste to Yucca Mountain: The Next Step in Nevada

    SciTech Connect (OSTI)

    Sweeney, Robin L,; Lechel, David J.

    2003-02-25T23:59:59.000Z

    In the U.S. Department of Energy's ''Final Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada,'' the Department states that certain broad transportation-related decisions can be made. These include the choice of a mode of transportation nationally (mostly legal-weight truck or mostly rail) and in Nevada (mostly rail, mostly legal-weight truck, or mostly heavy-haul truck with use of an associated intermodal transfer station), as well as the choice among alternative rail corridors or heavy-haul truck routes with use of an associated intermodal transfer station in Nevada. Although a rail line does not service the Yucca Mountain site, the Department has identified mostly rail as its preferred mode of transportation, both nationally and in the State of Nevada. If mostly rail is selected for Nevada, the Department would then identify a preference for one of the rail corridors in consultation with affected stakeholders, particularly the State of Nevada. DOE would then select the rail corridor and initiate a process to select a specific rail alignment within the corridor for the construction of a rail line. Five proposed rail corridors were analyzed in the Final Environmental Impact Statement. The assessment considered the impacts of constructing a branch rail line in the five 400-meter (0.25mile) wide corridors. Each corridor connects the Yucca Mountain site with an existing mainline railroad in Nevada.

  13. National survey of crystalline rocks and recommendations of regions to be explored for high-level radioactive waste repository sites

    SciTech Connect (OSTI)

    Smedes, H.W.

    1983-04-01T23:59:59.000Z

    A reconnaissance of the geological literature on large regions of exposed crystalline rocks in the United States provides the basis for evaluating if any of those regions warrant further exploration toward identifying potential sites for development of a high-level radioactive waste repository. The reconnaissance does not serve as a detailed evaluation of regions or of any smaller subunits within the regions. Site performance criteria were selected and applied insofar as a national data base exists, and guidelines were adopted that relate the data to those criteria. The criteria include consideration of size, vertical movements, faulting, earthquakes, seismically induced ground motion, Quaternary volcanic rocks, mineral deposits, high-temperature convective ground-water systems, hydraulic gradients, and erosion. Brief summaries of each major region of exposed crystalline rock, and national maps of relevant data provided the means for applying the guidelines and for recommending regions for further study. It is concluded that there is a reasonable likelihood that geologically suitable repository sites exist in each of the major regions of crystalline rocks. The recommendation is made that further studies first be conducted of the Lake Superior, Northern Appalachian and Adirondack, and the Southern Appalachian Regions. It is believed that those regions could be explored more effectively and suitable sites probably could be found, characterized, verified, and licensed more readily there than in the other regions.

  14. Adsorption of Ruthenium, Rhodium and Palladium from Simulated High-Level Liquid Waste by Highly Functional Xerogel - 13286

    SciTech Connect (OSTI)

    Onishi, Takashi [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan)] [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan); Koyama, Shin-ichi [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan)] [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan); Mimura, Hitoshi [Dept. of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University Aramaki-Aza-Aoba 6-6-01-2,Aoba-ku, Sendai-shi, Miyagi-ken, 980-8579 (Japan)] [Dept. of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University Aramaki-Aza-Aoba 6-6-01-2,Aoba-ku, Sendai-shi, Miyagi-ken, 980-8579 (Japan)

    2013-07-01T23:59:59.000Z

    Fission products are generated by fission reactions in nuclear fuel. Platinum group (Pt-G) elements, such as palladium (Pd), rhodium (Rh) and ruthenium (Ru), are also produced. Generally, Pt-G elements play important roles in chemical and electrical industries. Highly functional xerogels have been developed for recovery of these useful Pt-G elements from high - level radioactive liquid waste (HLLW). An adsorption experiment from simulated HLLW was done by the column method to study the selective adsorption of Pt-G elements, and it was found that not only Pd, Rh and Ru, but also nickel, zirconium and tellurium were adsorbed. All other elements were not adsorbed. Adsorbed Pd was recovered by washing the xerogel-packed column with thiourea solution and thiourea - nitric acid mixed solution in an elution experiment. Thiourea can be a poison for automotive exhaust emission system catalysts, so it is necessary to consider its removal. Thermal decomposition and an acid digestion treatment were conducted to remove sulfur in the recovered Pd fraction. The relative content of sulfur to Pd was decreased from 858 to 0.02 after the treatment. These results will contribute to design of the Pt-G element separation system. (authors)

  15. Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 1, Methodology and results

    SciTech Connect (OSTI)

    Rechard, R.P. [ed.

    1993-12-01T23:59:59.000Z

    This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste. Although numerous caveats must be placed on the results, the general findings were as follows: Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

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

    SciTech Connect (OSTI)

    S. Frank

    2010-09-01T23:59:59.000Z

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

  17. Comprehensive data base of high-level nuclear waste glasses: September 1987 status report: Volume 1, Discussion and glass durability data

    SciTech Connect (OSTI)

    Kindle, C.H.; Kreiter, M.R.

    1987-12-01T23:59:59.000Z

    The Materials Characterization Center (MCC) at Pacific Northwest Laboratory is assembling a comprehensive data base (CDB) of experimental data collected for high-level nuclear waste package components. Data collected throughout the world are included in the data base; current emphasis is on waste glasses and their properties. The goal is to provide a data base of properties and compositions and an analysis of dominant property trends as a function of composition. This data base is a resource that nuclear waste producers, disposers, and regulators can use to compare properties of a particular high-level nuclear waste glass product with the properties of other glasses of similar compositions. Researchers may use the data base to guide experimental tests to fill gaps in the available knowledge or to refine empirical models. The data are incorporated into a computerized data base that will allow the data to be extracted based on, for example, glass composition or test duration. 3 figs.

  18. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 7. Revision 1

    SciTech Connect (OSTI)

    Burt, D.L.

    1994-04-01T23:59:59.000Z

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 7) presents the standards and requirements for the following sections: Occupational Safety and Health, and Environmental Protection.

  19. EIS-0074: Long-Term Management of Defense High-Level Radioactive Wastes Idaho Chemical Processing Plant, Idaho National Engineering Lab, Idaho

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy prepared this statement to analyze the environmental implications of the proposed selection of a strategy for long- term management of the high- level radioactive wastes generated as part of the national defense effort at the Department's Idaho Chemical Processing Plant a t the Idaho National Engineering Laboratory.

  20. EIS-0023: Long-Term Management of Defense High-Level Radioactive Wastes (Research and Development Program for Immobilization), Savannah River Plant, Aiken, South Carolina

    Broader source: Energy.gov [DOE]

    This environmental impact statement (EIS) analyzes the environmental implications of the proposed continuation of a large Federal research and development (R&D) program directed toward the immobilization of the high-level radioactive wastes resulting from chemical separations operations for defense radionuclides production at the DOE Savannah River Plant (SRP) near Aiken, South Carolina.

  1. [alpha]-Decay damage effects in curium-doped titanate ceramic containing sodium-free high-level nuclear waste

    SciTech Connect (OSTI)

    Mitamura, Hisayoshi; Matsumoto, Seiichiro; Tsuboi, Takashi; Hashimoto, Masaaki; Togashi, Yoshihiro; Kanazawa, Hiroyuki (Japan Atomic Energy Research Inst., Ibaraki (Japan)); Stewart, M.W.A.; Vance, E.R.; Hart, K.P.; Ball, C.J. (Australian Nuclear Science and Technology Organization, Lucas Heights, New South Wales (Australia). Lucas Heights Research Labs.); White, T.J.

    1994-09-01T23:59:59.000Z

    A polyphase titanate ceramic incorporating sodium-free simulated high-level nuclear waste was doped with 0.91 wt% of [sup 224]Cm to accelerate the effects of long-term self-irradiation arising from [alpha] decays. The ceramic included three main constituent minerals: hollandite, perovskite, and zirconolite, with some minor phases. Although hollandite showed the broadening of its X-ray diffraction lines and small lattice parameter changes during damage in growth, the unit cell was substantially unaltered. Perovskite and zirconolite, which are the primary hosts of curium, showed 2.7% and 2.6% expansions, respectively, of their unit cell volumes after a dose of 12 [times] 10[sup 17] [alpha] decays[center dot]g[sup [minus]1]. Volume swelling due to damage in growth caused an exponential (almost linear) decrease in density, which reached 1.7% after a dose of 12.4 [times] 10[sup 17] [alpha] decays[center dot]g[sup [minus]1]. Leach tests on samples that had incurred doses of 2.0 [times] 10[sup 17] and 4.5 [times] 10[sup 17] [alpha] decays[center dot]g[sup [minus]1] showed that the rates of dissolution of cesium and barium were similar to analogous leach rates from the equivalent cold ceramic, while strontium and calcium leach rates were 2--15 times higher. Although the cerium, molybdenum, strontium, and calcium leach rates in the present material were similar to those in the curium-doped sodium-bearing titanate ceramic reported previously, the cesium leach rate was 3--8 times lower.

  2. Branch technical position on the use of expert elicitation in the high-level radioactive waste program

    SciTech Connect (OSTI)

    Kotra, J.P.; Lee, M.P.; Eisenberg, N.A. [Nuclear Regulatory Commission, Washington, DC (United States); DeWispelare, A.R. [Center for Nuclear Waste Regulatory Analyses, San Antonio, TX (United States)

    1996-11-01T23:59:59.000Z

    Should the site be found suitable, DOE will apply to the US Nuclear Regulatory Commission for permission to construct and then operate a proposed geologic repository for the disposal of spent nuclear fuel and other high-level radioactive waste at Yucca Mountain. In deciding whether to grant or deny DOE`s license application for a geologic repository, NRC will closely examine the facts and expert judgment set forth in any potential DOE license application. NRC expects that subjective judgments of individual experts and, in some cases, groups of experts, will be used by DOE to interpret data obtained during site characterization and to address the many technical issues and inherent uncertainties associated with predicting the performance of a repository system for thousands of years. NRC has traditionally accepted, for review, expert judgment to evaluate and interpret the factual bases of license applications and is expected to give appropriate consideration to the judgments of DOE`s experts regarding the geologic repository. Such consideration, however, envisions DOE using expert judgments to complement and supplement other sources of scientific and technical information, such as data collection, analyses, and experimentation. In this document, the NRC staff has set forth technical positions that: (1) provide general guidelines on those circumstances that may warrant the use of a formal process for obtaining the judgments of more than one expert (i.e., expert elicitation); and (2) describe acceptable procedures for conducting expert elicitation when formally elicited judgments are used to support a demonstration of compliance with NRC`s geologic disposal regulation, currently set forth in 10 CFR Part 60. 76 refs.

  3. Application of curium measurements for safeguarding at reprocessing plants. Study 1: High-level liquid waste and Study 2: Spent fuel assemblies and leached hulls

    SciTech Connect (OSTI)

    Rinard, P.M.; Menlove, H.O.

    1996-03-01T23:59:59.000Z

    In large-scale reprocessing plants for spent fuel assemblies, the quantity of plutonium in the waste streams each year is large enough to be important for nuclear safeguards. The wastes are drums of leached hulls and cylinders of vitrified high-level liquid waste. The plutonium amounts in these wastes cannot be measured directly by a nondestructive assay (NDA) technique because the gamma rays emitted by plutonium are obscured by gamma rays from fission products, and the neutrons from spontaneous fissions are obscured by those from curium. The most practical NDA signal from the waste is the neutron emission from curium. A diversion of waste for its plutonium would also take a detectable amount of curium, so if the amount of curium in a waste stream is reduced, it can be inferred that there is also a reduced amount of plutonium. This report studies the feasibility of tracking the curium through a reprocessing plant with neutron measurements at key locations: spent fuel assemblies prior to shearing, the accountability tank after dissolution, drums of leached hulls after dissolution, and canisters of vitrified high-level waste after separation. Existing pertinent measurement techniques are reviewed, improvements are suggested, and new measurements are proposed. The authors integrate these curium measurements into a safeguards system.

  4. Technology and apparatus for solidification of radioactive wastes from nuclear fuel cycle by high temperature adsorption of metals on inorganic matrices

    SciTech Connect (OSTI)

    Nardova, A.K.; Philipov, E.A.; Kudriavtsev, Y.G.; Dzekun, E.G.; Parfanovitch, B.N. [Russian Research Inst. of Chemical Technology, Moscow (Russian Federation)

    1993-12-31T23:59:59.000Z

    This study deals with the investigation of high-level waste (HLW) solidification by high-temperature adsorption of radionuclides on porous inorganic matrices. An appropriate drum-type apparatus using magnetic gear drive was designed and tested. The report contains the test results of the solidification process of high-level radioactive raffinate from the first regeneration extraction cycle of irradiated fuel elements from nuclear power plants. Industrial-scale tests of the HLW solidification process (technology and equipment) are planned.

  5. High Performance Zero-Bleed CLSM/Grout Mixes for High-Level Waste Tank Closures Strategic Research and Development - FY99 Report

    SciTech Connect (OSTI)

    Langton, C.A.

    2000-08-11T23:59:59.000Z

    The overall objective of this program, SRD-99-08, was to design and test suitable materials, which can be used to close high-level waste tanks at SRS. Fill materials can be designed to perform several functions including chemical stabilization and/or physical encapsulation of incidental waste so that the potential for transport of contaminants into the environment is reduced. Also they are needed to physically stabilize the void volume in the tanks to prevent/minimize future subsidence and inadvertent intrusion. The intent of this work was to develop a zero-bleed soil CLSM (ZBS-CLSM) and a zero-bleed concrete mix (ZBC) which meet the unique placement and stabilization/encapsulation requirements for high-level waste tank closures. These mixes in addition to the zero-bleed CLSM mixes formulated for closure of Tanks 17-F and 20-F provide design engineers with a suite of options for specifying materials for future tank closures.

  6. Draft Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada

    SciTech Connect (OSTI)

    N /A

    1999-08-13T23:59:59.000Z

    The Proposed Action addressed in this EIS is to construct, operate and monitor, and eventually close a geologic repository at Yucca Mountain in southern Nevada for the disposal of spent nuclear fuel and high-level radioactive waste currently in storage at 72 commercial and 5 DOE sites across the United States. The EIS evaluates (1) projected impacts on the Yucca Mountain environment of the construction, operation and monitoring, and eventual closure of the geologic repository; (2) the potential long-term impacts of repository disposal of spent nuclear fuel and high-level radioactive waste; (3) the potential impacts of transporting these materials nationally and in the State of Nevada; and (4) the potential impacts of not proceeding with the Proposed Action.

  7. Site characterization plan conceptual design report for a high-level nuclear waste repository in salt, vertical emplacement mode: Volume 1

    SciTech Connect (OSTI)

    Not Available

    1987-12-01T23:59:59.000Z

    This Conceptual Design Report describes the conceptual design of a high-level nuclear waste repository in salt at a proposed site in Deaf Smith County, Texas. Waste receipt, processing, packing, and other surface facility operations are described. Operations in the shafts underground are described, including waste hoisting, transfer, and vertical emplacement. This report specifically addresses the vertical emplacement mode, the reference design for the repository. Waste retrieval capability is described. The report includes a description of the layout of the surface, shafts, and underground. Major equipment items are identified. The report includes plans for decommissioning and sealing of the facility. The report discusses how the repository will satisfy performance objectives. Chapters are included on basis for design, design analyses, and data requirements for completion of future design efforts. 105 figs., 52 tabs.

  8. AKUFVE studies on extraction behavior of neptunium from simulated HLW solutions by 30% TBP

    SciTech Connect (OSTI)

    Chitnis, R.R.; Wattal, P.K.; Murali, M.S.; Nair, G.C.; Mathur, J.N. [Bhabha Atomic Research Centre, Bombay (India)

    1998-07-01T23:59:59.000Z

    The extraction behavior of neptunium from 3 M nitric acid as well as simulated pressurized heavy water reactor high level radioactive waste (PHWR-HLW) solution by 30% TBP/dodecane was studied using AKUFVE. Np(IV)/Np(V) was oxidized to Np(VI) using oxidizing agents, such as K{sub 2}Cr{sub 2}O{sub 7}, VO{sub 2}{sup +} and NaNO{sub 2}. Stripping of neptunium from the loaded TBP phase was studied using reducing agents like hydrogen peroxide, ascorbic acid, hydroxylamine hydrochloride and hydrazine sulfate. Results of these extraction and stripping studies have been discussed in this paper.

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

    Broader source: Energy.gov [DOE]

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

  10. Molybdenum in Nuclear Waste Glasses -Incorporation and Redox state R.J. Short, R.J. Hand, N.C. Hyatt,

    E-Print Network [OSTI]

    Sheffield, University of

    Molybdenum in Nuclear Waste Glasses - Incorporation and Redox state R.J. Short, R.J. Hand, N form in simulated high level nuclear waste (HLW) glass melts have been studied. It was found less attention has been paid to the effects of redox on nuclear waste glasses. One particular element

  11. Degradation mode survey galvanic corrosion of candidate metallic materials for high-level radioactive waste disposal containers

    SciTech Connect (OSTI)

    Roy, A.K.; Jones, D.A.; McCright, R.D.

    1996-08-01T23:59:59.000Z

    This report presents the results of a literature review on galvanic corrosion between candidate metals and alloys currently being considered for the waste package containers.

  12. A report on high-level nuclear waste transportation: Prepared pursuant to assembly concurrent resolution No. 8 of the 1987 Nevada Legislature

    SciTech Connect (OSTI)

    NONE

    1988-12-01T23:59:59.000Z

    This report has been prepared by the staff of the State of Nevada Agency for Nuclear Projects/Nuclear Waste Project Office (NWPO) in response to Assembly Concurrent Resolution No. 8 (ACR 8), passed by the Nevada State Legislature in 1987. ACR 8 directed the NWPO, in cooperation with affected local governments and the Legislative committee on High-Level Radioactive Waste, to prepare this report which scrutinizes the US Department of Energy`s (DOE) plans for transportation of high-level radioactive waste to the proposed yucca Mountain repository, which reviews the regulatory structure under which shipments to a repository would be made and which presents NWPO`s plans for addressing high-level radioactive waste transportation issues. The report is divided into three major sections. Section 1.0 provides a review of DOE`s statutory requirements, its repository transportation program and plans, the major policy, programmatic, technical and institutional issues and specific areas of concern for the State of Nevada. Section 2.0 contains a description of the current federal, state and tribal transportation regulatory environment within which nuclear waste is shipped and a discussion of regulatory issues which must be resolved in order for the State to minimize risks and adverse impacts to its citizens. Section 3.0 contains the NWPO plan for the study and management of repository-related transportation. The plan addresses four areas, including policy and program management, regulatory studies, technical reviews and studies and institutional relationships. A fourth section provides recommendations for consideration by State and local officials which would assist the State in meeting the objectives of the plan.

  13. Outlooks of HLW Partitioning Technologies Usage for Recovering of Platinum Metals from Spent Fuel

    SciTech Connect (OSTI)

    Pokhitonov, Y. A.; Estimantovskiy, V.; Romanovski, v.; Zatsev, B.; Todd, T.

    2003-02-24T23:59:59.000Z

    The existing practice of management of high level waste (HLW) generated by NPPs, call for a task of selective separation of the most dangerous long-lived radionuclides with the purpose of their subsequent immobilization and disposal. HLW partitioning allows to reduce substantially the cost of vitrified product storage owing to isolation of the most dangerous radionuclides, such as transplutonium elements (TPE) into separate fractions of small volumes, intended for ultimate storage. By now numerous investigations on partitioning of HLW of various composition have been carried out in many countries and a lot of processes permitting to recover cesium, strontium, TPE and rare earth elements (REE) have been already tested. Apart from enumerated radionuclides, a fair quantity of palladium and rhodium presents in spent fuel, but the problem of these elements recovery has not yet been decided at the operating radiochemical plants. A negative effect of platinum group metals (PGM) occurrence is determined by the formation of separate metal phase, which not only worsens the conditions of glass-melting but also shortens considerably the service life of the equipment. At the same time, the exhaustion of PGMs natural resources may finally lead to such a growth of their costs that the spent nuclear fuel would became a substituting source of these elements industrial production. Allowing above mentioned, it is of interest to develop the technique for ''reactor'' palladium and rhodium recovery process which would be compatible with HLW partitioning and could be realized using the same facilities. In the report the data on platinum metals distribution in spent fuel reprocessing products and the several flowsheets for palladium separation from HLW are presented.

  14. EIS-0113: Disposal of Hanford Defense High-Level, Transuranic and Tank Waste, Hanford Site, Richland, Washington

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy developed this EIS to examine the potential environmental impacts of final disposal options for legacy and future radioactive defense wastes stored at the Hanford Site.

  15. FURTHER DEVELOPMENT OF MODIFIED MONOSODIUM TITANATE, AN IMPROVED SORBENT FOR PRETREATMENT OF HIGH LEVEL NUCLEAR WASTE AT THE SAVANNAH RIVER SITE

    SciTech Connect (OSTI)

    Taylor-Pashow, K.; Hobbs, D.; Fondeur, F.; Fink, S.

    2011-01-12T23:59:59.000Z

    High-level nuclear waste produced from fuel reprocessing operations at the Savannah River Site (SRS) requires pretreatment to remove Cs-137, Sr-90, and alpha-emitting radionuclides (i.e., actinides) prior to disposal onsite as low level waste. Separation processes planned at SRS include caustic side solvent extraction, for Cs-137 removal, and sorption of Sr-90 and alpha-emitting radionuclides onto monosodium titanate (MST). The predominant alpha-emitting radionuclides in the highly alkaline waste solutions include plutonium isotopes Pu-238, Pu-239, and Pu-240. This paper describes recent results from the development of an improved titanate material that exhibits increased removal kinetics and effective capacity for Sr-90 and alpha-emitting radionuclides compared to the baseline MST material.

  16. High Level Waste Tank Farm Replacement Project for the Idaho Chemical Processing Plant at the Idaho National Engineering Laboratory. Environmental Assessment

    SciTech Connect (OSTI)

    Not Available

    1993-06-01T23:59:59.000Z

    The Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA-0831, for the construction and operation of the High-Level Waste Tank Farm Replacement (HLWTFR) Project for the Idaho Chemical Processing Plant located at the Idaho National Engineering Laboratory (INEL). The HLWTFR Project as originally proposed by the DOE and as analyzed in this EA included: (1) replacement of five high-level liquid waste storage tanks with four new tanks and (2) the upgrading of existing tank relief piping and high-level liquid waste transfer systems. As a result of the April 1992 decision to discontinue the reprocessing of spent nuclear fuel at INEL, DOE believes that it is unlikely that the tank replacement aspect of the project will be needed in the near term. Therefore, DOE is not proposing to proceed with the replacement of the tanks as described in this-EA. The DOE`s instant decision involves only the proposed upgrades aspect of the project described in this EA. The upgrades are needed to comply with Resource Conservation and Recovery Act, the Idaho Hazardous Waste Management Act requirements, and the Department`s obligations pursuant to the Federal Facilities Compliance Agreement and Consent Order among the Environmental Protection Agency, DOE, and the State of Idaho. The environmental impacts of the proposed upgrades are adequately covered and are bounded by the analysis in this EA. If DOE later proposes to proceed with the tank replacement aspect of the project as described in the EA or as modified, it will undertake appropriate further review pursuant to the National Environmental Policy Act.

  17. Application of Analytical Heat Transfer Models of Multi-layered Natural and Engineered Barriers in Potential High-Level Nuclear Waste Repositories - 12435

    SciTech Connect (OSTI)

    Greenberg, Harris R.; Blink, James A.; Fratoni, Massimiliano; Sutton, Mark [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Ross, Amber D. [University of the Sciences in Philadelphia, Philadelphia, PA 19104 (United States)

    2012-07-01T23:59:59.000Z

    A combination of transient heat transfer analytical solutions for a finite line source, a series of point sources, and a series of parallel infinite line sources were combined with a quasi-steady-state multi-layered cylindrical solution to simulate the temperature response of a deep geologic radioactive waste repository with multi-layered natural and engineered barriers. This evaluation was performed to provide information to scientists and decision makers to compare candidate geologic media for a repository (crystalline rock [granite], clay, salt, and deep borehole), and to provide input for the future evaluation of the trade-off between pre-emplacement surface storage time, waste package size, and repository footprint. This approach was selected in favor of the finite element solution typically used to analyze the temperature response because it allowed rapid comparison of a large number of alternative disposal options and design configurations. More than 100 combinations of waste form, geologic environment, repository design configuration, and surface storage times were analyzed and compared. The analytical solution approach used to analyze the repository temperature response allowed rapid comparison of a large number of alternative disposal options and design configurations. More than 100 combinations of waste form, geologic environment, repository design configuration, and surface storage times were analyzed and compared. This approach allowed investigation of the sensitivity of the results to combinations of parameters that show that there is much flexibility to be gained in terms of spent fuel management options by varying a few key parameters. This initial analysis used representative design concepts and thermal constraints based on international design concepts, and it also included waste forms representing future fuel cycles with high burnup fuels. Unlike repository designs with large open tunnels and pre-closure ventilation, all of the disposal concepts analyzed in this study used enclosed emplacement modes, where the waste packages were in direct contact with encapsulating engineered or natural materials. The deep borehole repository concept limits the size of the SNF waste packages and may require rod consolidation to fit within the drill casing diameter. A single assembly waste package, assuming rod consolidation, was evaluated in the current analysis. Similar size restrictions apply for the HLW canisters. At this time no thermal constraints have been defined for the deep borehole repository concept. Representative EBS materials and properties were evaluated. However, changes in EBS design concepts and materials can also have significant effects on the maximum waste package surface temperature. Increased thermal conductivity of the buffer layer can be achieved by using an engineered buffer consisting of a mixture of graphite, sand, and bentonite [14]. One of the advantages of the analytical model is that it highlights the sensitivity of the results to the parameters that define the repository layout, including spacing between axial and lateral neighboring waste packages and drifts. It is clear that repository layout adjustments can be made to reduce the calculated peak temperatures. The results also show that significant reductions in required surface storage times can be achieved if higher thermal constraints can be justified Additional studies are planned to evaluate the trade-offs between surface storage times, repository layout parameters, and variations in EBS design concepts. Model validation and uncertainties will also be addressed. It is expected that shorter surface storage times and more optimized repository design configurations may be achieved. (authors)

  18. Risk-based systems analysis of emerging high-level waste tank remediation technologies. Volume 2: Final report

    SciTech Connect (OSTI)

    Peters, B.B.; Cameron, R.J.; McCormack, W.D. [Enserch Environmental Corp., Richland, WA (United States)

    1994-08-01T23:59:59.000Z

    The objective of DOE`s Radioactive Waste Tank Remediation Technology Focus Area is to identify and develop new technologies that will reduce the risk and/or cost of remediating DOE underground waste storage tanks and tank contents. There are, however, many more technology investment opportunities than the current budget can support. Current technology development selection methods evaluate new technologies in isolation from other components of an overall tank waste remediation system. This report describes a System Analysis Model developed under the US Department of Energy (DOE) Office of Technology Development (OTD) Underground Storage Tank-Integrated Demonstration (UST-ID) program. The report identifies the project objectives and provides a description of the model. Development of the first ``demonstration`` version of this model and a trial application have been completed and the results are presented. This model will continue to evolve as it undergoes additional user review and testing.

  19. Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers. Final report

    SciTech Connect (OSTI)

    Vinson, D.W.; Bullen, D.B. [Iowa State Univ. of Science and Technology, Ames, IA (United States)

    1995-09-22T23:59:59.000Z

    One of the most significant factors impacting the performance of waste package container materials under repository relevant conditions is the thermal environment. This environment will be affected by the areal power density of the repository, which is dictated by facility design, and the dominant heat transfer mechanism at the site. The near-field environment will evolve as radioactive decay decreases the thermal output of each waste package. Recent calculations (Buscheck and Nitao, 1994) have addressed the importance of thermal loading conditions on waste package performance at the Yucca Mountain site. If a relatively low repository thermal loading design is employed, the temperature and relative humidity near the waste package may significantly affect the degradation of corrosion allowance barriers due to moist air oxidation and radiolytically enhanced corrosion. The purpose this report is to present a literature review of the potential degradation modes for moderately corrosion resistant nickel copper and nickel based candidate materials that may be applicable as alternate barriers for the ACD systems in the Yucca Mountain environment. This report presents a review of the corrosion of nickel-copper alloys, summaries of experimental evaluations of oxidation and atmospheric corrosion in nickel-copper alloys, views of experimental studies of aqueous corrosion in nickel copper alloys, a brief review of galvanic corrosion effects and a summary of stress corrosion cracking in these alloys.

  20. Final Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada

    SciTech Connect (OSTI)

    N /A

    2002-10-25T23:59:59.000Z

    The purpose of this environmental impact statement (EIS) is to provide information on potential environmental impacts that could result from a Proposed Action to construct, operate and monitor, and eventually close a geologic repository for the disposal of spent nuclear fuel and high-level radioactive waste at the Yucca Mountain site in Nye County, Nevada. The EIS also provides information on potential environmental impacts from an alternative referred to as the No-Action Alternative, under which there would be no development of a geologic repository at Yucca Mountain.

  1. The U.S. Nuclear Waste Technical Review Board evaluates the technical and scientific validity of ac-

    E-Print Network [OSTI]

    Chapter 1 Overview The U.S. Nuclear Waste Technical Review Board evaluates the technical in Nevada for its suit- ability as a location for a repository for high-level ra- dioactive waste (HLW) and spent nuclear fuel (SNF). The U.S. Department of Energy (DOE) began studying Yucca Mountain

  2. Some logistical considerations in designing a system of deep boreholes for disposal of high-level radioactive waste.

    SciTech Connect (OSTI)

    Gray, Genetha Anne; Brady, Patrick Vane [Sandia National Laboratories, Albuquerque, NM] [Sandia National Laboratories, Albuquerque, NM; Arnold, Bill Walter [Sandia National Laboratories, Albuquerque, NM] [Sandia National Laboratories, Albuquerque, NM

    2012-09-01T23:59:59.000Z

    Deep boreholes could be a relatively inexpensive, safe, and rapidly deployable strategy for disposing Americas nuclear waste. To study this approach, Sandia invested in a three year LDRD project entitled %E2%80%9CRadionuclide Transport from Deep Boreholes.%E2%80%9D In the first two years, the borehole reference design and backfill analysis were completed and the supporting modeling of borehole temperature and fluid transport profiles were done. In the third year, some of the logistics of implementing a deep borehole waste disposal system were considered. This report describes what was learned in the third year of the study and draws some conclusions about the potential bottlenecks of system implementation.

  3. Summary report of first and foreign high-level waste repository concepts; Technical report, working draft 001

    SciTech Connect (OSTI)

    Hanke, P.M.

    1987-11-04T23:59:59.000Z

    Reference repository concepts designs adopted by domestic and foreign waste disposal programs are reviewed. Designs fall into three basic categories: deep borehole from the surface; disposal in boreholes drilled from underground excavations; and disposal in horizontal tunnels or drifts. The repository concepts developed in Sweden, Switzerland, Finland, Canada, France, Japan, United Kingdom, Belgium, Italy, Holland, Denmark, West Germany and the United States are described. 140 refs., 315 figs., 19 tabs.

  4. Design and performance of a full-scale spray calciner for nonradioactive high-level-waste-vitrification studies

    SciTech Connect (OSTI)

    Miller, F.A.

    1981-06-01T23:59:59.000Z

    In the spray calcination process, liquid waste is spray-dried in a heated-wall spray dryer (termed a spray calciner), and then it may be combined in solid form with a glass-forming frit. This mixture is then melted in a continuous ceramic melter or in an in-can melter. Several sizes of spray calciners have been tested at PNL- laboratory scale, pilot scale and full scale. Summarized here is the experience gained during the operation of PNL's full-scale spray calciner, which has solidified approx. 38,000 L of simulated acid wastes and approx. 352,000 L of simulated neutralized wastes in 1830 h of processing time. Operating principles, operating experience, design aspects, and system descriptions of a full-scale spray calciner are discussed. Individual test run summaries are given in Appendix A. Appendices B and C are studies made by Bechtel Inc., under contract by PNL. These studies concern, respectively, feed systems for the spray calciner process and a spray calciner vibration analysis. Appendix D is a detailed structural analysis made at PNL of the spray calciner. These appendices are included in the report to provide a complete description of the spray calciner and to include all major studies made concerning PNL's full-scale spray calciner.

  5. APPLICATION OF A THIN FILM EVAPORATOR SYSTEM FOR MANAGEMENT OF LIQUID HIGH-LEVEL WASTES AT HANFORD

    SciTech Connect (OSTI)

    TEDESCHI AR; WILSON RA

    2010-01-14T23:59:59.000Z

    A modular, transportable evaporator system, using thin film evaporative technology, is planned for deployment at the Hanford radioactive waste storage tank complex. This technology, herein referred to as a wiped film evaporator (WFE), will be located at grade level above an underground storage tank to receive pumped liquids, concentrate the liquid stream from 1.1 specific gravity to approximately 1.4 and then return the concentrated solution back into the tank. Water is removed by evaporation at an internal heated drum surface exposed to high vacuum. The condensed water stream will be shipped to the site effluent treatment facility for final disposal. This operation provides significant risk mitigation to failure of the aging 242-A Evaporator facility; the only operating evaporative system at Hanford maximizing waste storage. This technology is being implemented through a development and deployment project by the tank farm operating contractor, Washington River Protection Solutions (WRPS), for the Office of River Protection/Department of Energy (ORP/DOE), through Columbia Energy & Environmental Services, Inc. (Columbia Energy). The project will finalize technology maturity and install a system at one of the double-shell tank farms. This paper discusses results of pre-project pilot-scale testing by Columbia Energy and ongoing technology maturation development scope through fiscal year 2012, including planned additional pilot-scale and full-scale simulant testing and operation with actual radioactive tank waste.

  6. Development of Dodecaniobate Keggin Chain Materials as Alternative Sorbents for SR and Actinide Removal from High-Level Nuclear Waste Solutions

    SciTech Connect (OSTI)

    Nyman, May; Bonhomme, Francois

    2004-03-28T23:59:59.000Z

    The current baseline sorbent (monosodium titanate) for Sr and actinide removal from Savannah River Site's high level wastes has excellent adsorption capabilities for Sr but poor performance for the actinides. We are currently investigating the development of alternative materials that sorb radionuclides based on chemical affinity and/or size selectivity. The polyoxometalates, negatively-charged metal oxo clusters, have known metal binding properties and are of interest for radionuclide sequestration. We have developed a class of Keggin-ion based materials, where the Keggin ions are linked in 1- dimensional chains separated by hydrated, charge-balancing cations. These Nb-based materials are stable in the highly basic nuclear waste solutions and show good selectivity for Sr and Pu. Synthesis, characterization and structure of these materials in their native forms and Sr-exchanged forms will be presented.

  7. High-level and transuranic radioactive wastes: Background information document for amendments to 40 CFR part 191

    SciTech Connect (OSTI)

    Not Available

    1993-11-01T23:59:59.000Z

    The report provides the necessary background information technical analyses, and justifications in support of the proposed amendments to 40 CFR Part 191. The scope of the report encompasses the conceptual framework for assessing radiation exposures and associated health risks. In general terms, this assessment examines the radioactive source term characterization, analysis of the movement of radionuclides from the repository through the appropriate environmental exposure pathways and doses received by members of the general public. The report used transuranic waste for individual dose and ground-water protection analysis.

  8. End of FY10 report - used fuel disposition technical bases and lessons learned : legal and regulatory framework for high-level waste disposition in the United States.

    SciTech Connect (OSTI)

    Weiner, Ruth F.; Blink, James A. (Lawrence Livermore National Laboratory, Livermore, CA); Rechard, Robert Paul; Perry, Frank (Los Alamos National Laboratory, Los Alamos, NM); Jenkins-Smith, Hank C. (University of Oklahoma, Norman, OK); Carter, Joe (Savannah River Nuclear Solutions, Aiken, SC); Nutt, Mark (Argonne National Laboratory, Argonne, IL); Cotton, Tom (Complex Systems Group, Washington DC)

    2010-09-01T23:59:59.000Z

    This report examines the current policy, legal, and regulatory framework pertaining to used nuclear fuel and high level waste management in the United States. The goal is to identify potential changes that if made could add flexibility and possibly improve the chances of successfully implementing technical aspects of a nuclear waste policy. Experience suggests that the regulatory framework should be established prior to initiating future repository development. Concerning specifics of the regulatory framework, reasonable expectation as the standard of proof was successfully implemented and could be retained in the future; yet, the current classification system for radioactive waste, including hazardous constituents, warrants reexamination. Whether or not consideration of multiple sites are considered simultaneously in the future, inclusion of mechanisms such as deliberate use of performance assessment to manage site characterization would be wise. Because of experience gained here and abroad, diversity of geologic media is not particularly necessary as a criterion in site selection guidelines for multiple sites. Stepwise development of the repository program that includes flexibility also warrants serious consideration. Furthermore, integration of the waste management system from storage, transportation, and disposition, should be examined and would be facilitated by integration of the legal and regulatory framework. Finally, in order to enhance acceptability of future repository development, the national policy should be cognizant of those policy and technical attributes that enhance initial acceptance, and those policy and technical attributes that maintain and broaden credibility.

  9. AN EVALUATION OF HYDROGEN INDUCED CRACKING SUSCEPTIBILITY OF TITANIUM ALLOYS IN US HIGH-LEVEL NUCLEAR WASTE REPOSITORY ENVIRONMENTS

    SciTech Connect (OSTI)

    G. De; K. Mon; G. Gordon; D. Shoesmith; F. Hua

    2006-02-21T23:59:59.000Z

    This paper evaluates hydrogen-induced cracking (HIC) susceptibility of titanium alloys in environments anticipated in the Yucca Mountain nuclear waste repository with particular emphasis on the. effect of the oxide passive film on the hydrogen absorption process of titanium alloys being evaluated. The titanium alloys considered in this review include Ti 2, 5 , 7, 9, 11, 12, 16, 17, 18, 24 and 29. In general, the concentration of hydrogen in a titanium alloy can increase due to absorption of atomic hydrogen produced from passive general corrosion of that alloy or galvanic coupling of it to a less noble metal. It is concluded that under the exposure conditions anticipated in the Yucca Mountain repository, the HIC of titanium drip shield will not occur because there will not be sufficient hydrogen in the metal even after 10,000 years of emplacement. Due to the conservatisms adopted in the current evaluation, this assessment is considered very conservative.

  10. THE IMPACT OF OZONE ON THE LOWER FLAMMABLE LIMIT OF HYDROGEN IN VESSELS CONTAINING SAVANNAH RIVER SITE HIGH LEVEL WASTE

    SciTech Connect (OSTI)

    Sherburne, Carol [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Remediation, LLC; Osterberg, Paul [Fauske and Associates, LLC, Burr Ridge, IL (United States); Johnson, Tom [Fauske and Associates, LLC, Burr Ridge, IL (United States); Frawely, Thomas [Fauske and Associates, LLC, Burr Ridge, IL (United States)

    2013-01-23T23:59:59.000Z

    The Savannah River Site, in conjunction with AREVA Federal services, has designed a process to treat dissolved radioactive waste solids with ozone. It is known that in this radioactive waste process, radionuclides radiolytically break down water into gaseous hydrogen and oxygen, which presents a well defined flammability hazard. Flammability limits have been established for both ozone and hydrogen separately; however, there is little information on mixtures of hydrogen and ozone. Therefore, testing was designed to provide critical flammability information necessary to support safety related considerations for the development of ozone treatment and potential scale-up to the commercial level. Since information was lacking on flammability issues at low levels of hydrogen and ozone, a testing program was developed to focus on filling this portion of the information gap. A 2-L vessel was used to conduct flammability tests at atmospheric pressure and temperature using a fuse wire ignition source at 1 percent ozone intervals spanning from no ozone to the Lower Flammable Limit (LFL) of ozone in the vessel, determined as 8.4%(v/v) ozone. An ozone generator and ozone detector were used to generate and measure the ozone concentration within the vessel in situ, since ozone decomposes rapidly on standing. The lower flammability limit of hydrogen in an ozone-oxygen mixture was found to decrease from the LFL of hydrogen in air, determined as 4.2 % (v/v) in this vessel. From the results of this testing, Savannah River was able to develop safety procedures and operating parameters to effectively minimize the formation of a flammable atmosphere.

  11. FULL SCALE TESTING TECHNOLOGY MATURATION OF A THIN FILM EVAPORATOR FOR HIGH-LEVEL LIQUID WASTE MANAGEMENT AT HANFORD - 12125

    SciTech Connect (OSTI)

    TEDESCHI AR; CORBETT JE; WILSON RA; LARKIN J

    2012-01-26T23:59:59.000Z

    Simulant testing of a full-scale thin-film evaporator system was conducted in 2011 for technology development at the Hanford tank farms. Test results met objectives of water removal rate, effluent quality, and operational evaluation. Dilute tank waste simulant, representing a typical double-shell tank supernatant liquid layer, was concentrated from a 1.1 specific gravity to approximately 1.5 using a 4.6 m{sup 2} (50 ft{sup 2}) heated transfer area Rototherm{reg_sign} evaporator from Artisan Industries. The condensed evaporator vapor stream was collected and sampled validating efficient separation of the water. An overall decontamination factor of 1.2E+06 was achieved demonstrating excellent retention of key radioactive species within the concentrated liquid stream. The evaporator system was supported by a modular steam supply, chiller, and control computer systems which would be typically implemented at the tank farms. Operation of these support systems demonstrated successful integration while identifying areas for efficiency improvement. Overall testing effort increased the maturation of this technology to support final deployment design and continued project implementation.

  12. Development of an NDA system for high-level waste from the Chernobyl new safe confinement construction site

    SciTech Connect (OSTI)

    Lee, Sang-yoon [Los Alamos National Laboratory; Browne, Michael C [Los Alamos National Laboratory; Rael, Carlos D [Los Alamos National Laboratory; Carroll, Colin J [SONALYST INC.; Sunshine, Alexander [NA-243; Novikov, Alexander [CHNPP; Lebedev, Evgeny [CHNPP

    2010-01-01T23:59:59.000Z

    In early 2009, preliminary excavation work has begun in preparation for the construction of the New Safe Confinement (NSC) at the Chernobyl Nuclear Power Plant (ChNPP) in Ukraine. The NSC is the structure that will replace the present containment structure and will confine the radioactive remains of the ChNPP Unit-4 reactor for the next 100 years. It is expected that special nuclear material (SNM) that was ejected from the Unit-4 reactor during the accident in 1986 could be uncovered and would therefore need to be safeguarded. ChNPP requested the assistance of the United States Department of Energy/National Nuclear Security Administration (NNSA) with developing a new non-destructive assay (NDA) system that is capable of assaying radioactive debris stored in 55-gallon drums. The design of the system has to be tailored to the unique circumstances and work processes at the NSC construction site and the ChNPP. This paper describes the Chernobyl Drum Assay System (CDAS), the solution devised by Los Alamos National Laboratory, Sonalysts Inc., and the ChNPP, under NNSA's International Safeguards and Engagement Program (INSEP). The neutron counter measures the spontaneous fission neutrons from the {sup 238}U, {sup 240}Pu, {sup 244}Cm in a waste drum and estimates the mass contents of the SNMs in the drum by using of isotopic compositions determined by fuel burnup. The preliminary evaluation on overall measurement uncertainty shows that the system meets design performance requirements imposed by the facility.

  13. A Transportation Risk Assessment Tool for Analyzing the Transport of Spent Nuclear Fuel and High-Level Radioactive Waste to the Proposed Yucca Mountain Repository

    SciTech Connect (OSTI)

    Ralph Best; T. Winnard; S. Ross; R. Best

    2001-08-17T23:59:59.000Z

    The Yucca Mountain Transportation Database was developed as a data management tool for assembling and integrating data from multiple sources to compile the potential transportation impacts presented in the Draft Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada (DEIS). The database uses the results from existing models and codes such as RADTRAN, RISKIND, INTERLINE, and HIGHWAY to estimate transportation-related impacts of transporting spent nuclear fuel and high-level radioactive waste from commercial reactors and U. S. Department of Energy (DOE) facilities to Yucca Mountain. The source tables in the database are compendiums of information from many diverse sources including: radionuclide quantities for each waste type; route and route characteristics for rail, legal-weight truck, heavy haul. truck, and barge transport options; state-specific accident and fatality rates for routes selected for analysis; packaging and shipment data by waste type; unit risk factors; the complex behavior of the packaged waste forms in severe transport accidents; and the effects of exposure to radiation or the isotopic specific effects of radionclides should they be released in severe transportation accidents. The database works together with the codes RADTRAN (Neuhauser, et al, 1994) and RISKlND (Yuan, et al, 1995) to calculate incident-free dose and accident risk. For the incident-free transportation scenario, the database uses RADTRAN and RISKIND-generated data to calculate doses to offlink populations, onlink populations, people at stops, crews, inspectors, workers at intermodal transfer stations, guards at overnight stops, and escorts, as well as non-radioactive pollution health effects. For accident scenarios, the database uses RADTRAN-generated data to calculate dose risks based on ingestion, inhalation, resuspension, immersion (cloudshine), and groundshine as well as non-radioactive traffic fatalities. The Yucca Mountain EIS Transportation Database was developed using Microsoft Access 97{trademark} software and the Microsoft Windows NT{trademark} operating system. The database consists of tables for storing data, forms for selecting data for querying, and queries for retrieving the data in a predefined format. Database queries retrieve records based on input parameters and are used to calculate incident-free and accident doses using unit risk factors obtained from RADTRAN results. The next section briefly provides some background that led to the development of the database approach used in preparing the Yucca Mountain DEIS. Subsequent sections provide additional details on the database structure and types of impacts calculated using the database.

  14. Walk the Line: The Development of Route Selection Standards for Spent Nuclear Fuel and High-level Radioactive Waste in the United States - 13519

    SciTech Connect (OSTI)

    Dilger, Fred [Black Mountain Research, Henderson, NV 81012 (United States)] [Black Mountain Research, Henderson, NV 81012 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects, Carson City, NV 80906 (United States)] [State of Nevada Agency for Nuclear Projects, Carson City, NV 80906 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge, CA 91330 (United States)] [Department of Sociology, California State University, Northridge, CA 91330 (United States)

    2013-07-01T23:59:59.000Z

    Although storage facilities for spent nuclear fuel (SNF) and high-level radioactive waste (HLRW) are widely dispersed throughout the United States, these materials are also relatively concentrated in terms of geographic area. That is, the impacts of storage occur in a very small geographic space. Once shipments begin to a national repository or centralized interim storage facility, the impacts of SNF and HLRW will become more geographically distributed, more publicly visible, and almost certainly more contentious. The selection of shipping routes will likely be a major source of controversy. This paper describes the development of procedures, regulations, and standards for the selection of routes used to ship spent nuclear fuel and high-level radioactive waste in the United States. The paper begins by reviewing the circumstances around the development of HM-164 routing guidelines. The paper discusses the significance of New York City versus the Department of Transportation and application of HM-164. The paper describes the methods used to implement those regulations. The paper will also describe the current HM-164 designated routes and will provide a summary data analysis of their characteristics. This analysis will reveal the relatively small spatial scale of the effects of HM 164. The paper will then describe subsequent developments that have affected route selection for these materials. These developments include the use of 'representative routes' found in the Department of Energy (DOE) 2008 Supplemental Environmental Impact Statement for the formerly proposed Yucca Mountain geologic repository. The paper will describe recommendations related to route selection found in the National Academy of Sciences 2006 report Going the Distance, as well as recommendations found in the 2012 Final Report of the Blue Ribbon Commission on America's Nuclear Future. The paper will examine recently promulgated federal regulations (HM-232) for selection of rail routes for hazardous materials transport. The paper concludes that while the HM 164 regime is sufficient for certain applications, it does not provide an adequate basis for a national plan to ship spent nuclear fuel and high-level radioactive waste to centralized storage and disposal facilities over a period of 30 to 50 years. (authors)

  15. Uncertainty and sensitivity analysis in the 2008 performance assessment for the proposed repository for high-level radioactive waste at Yucca Mountain, Nevada.

    SciTech Connect (OSTI)

    Helton, Jon Craig; Sallaberry, Cedric M.; Hansen, Clifford W.

    2010-05-01T23:59:59.000Z

    Extensive work has been carried out by the U.S. Department of Energy (DOE) in the development of a proposed geologic repository at Yucca Mountain (YM), Nevada, for the disposal of high-level radioactive waste. As part of this development, an extensive performance assessment (PA) for the YM repository was completed in 2008 [1] and supported a license application by the DOE to the U.S. Nuclear Regulatory Commission (NRC) for the construction of the YM repository [2]. This presentation provides an overview of the conceptual and computational structure of the indicated PA (hereafter referred to as the 2008 YM PA) and the roles that uncertainty analysis and sensitivity analysis play in this structure.

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-SeriesFlickr FlickrGuidedCH2M HILLAdministration | National|GradientHigh.

  17. TRU decontamination of high-level Purex waste by solvent extraction using a mixed octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine oxide/TBP/NPH (TRUEX) solvent

    SciTech Connect (OSTI)

    Horwitz, E.P.; Kalina, D.G.; Diamond, H.; Kaplan, L.; Vandegrift, G.F.; Leonard, R.A.; Steindler, M.J.; Schulz, W.W.

    1984-01-01T23:59:59.000Z

    The TRUEX (transuranium extraction) process was tested on a simulated high-level dissolved sludge waste (DSW). A batch counter-current extraction mode was used for seven extraction and three scrub stages. One additional extraction stage and two scrub stages and all strip stages were performed by batch extraction. The TRUEX solvent consisted of 0.20 M octyl(phenyl)-N,N-diisobutylcarbamoyl-methylphosphine oxide-1.4 M TBP in Conoco (C/sub 12/-C/sub 14/). The feed solution was 1.0 M in HNO/sub 3/, 0.3 M in H/sub 2/C/sub 2/O/sub 4/ and contained mixed (stable) fission products, U, Np, Pu, and Am, and a number of inert constituents, e.g., Fe and Al. The test showed that the process is capable of reducing the TRU concentration in the DSW by a factor of 4 x 10/sup 4/ (to <100 nCi/g of disposed form) and reducing the quantity of TRU waste by two orders of magnitude.

  18. Vitrified waste option study report

    SciTech Connect (OSTI)

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

    1998-02-01T23:59:59.000Z

    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.

  19. Standard practice for prediction of the long-term behavior of materials, including waste forms, used in engineered barrier systems (EBS) for geological disposal of high-level radioactive waste

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2007-01-01T23:59:59.000Z

    1.1 This practice describes test methods and data analyses used to develop models for the prediction of the long-term behavior of materials, such as engineered barrier system (EBS) materials and waste forms, used in the geologic disposal of spent nuclear fuel (SNF) and other high-level nuclear waste in a geologic repository. The alteration behavior of waste form and EBS materials is important because it affects the retention of radionuclides by the disposal system. The waste form and EBS materials provide a barrier to release either directly (as in the case of waste forms in which the radionuclides are initially immobilized), or indirectly (as in the case of containment materials that restrict the ingress of groundwater or the egress of radionuclides that are released as the waste forms and EBS materials degrade). 1.1.1 Steps involved in making such predictions include problem definition, testing, modeling, and model confirmation. 1.1.2 The predictions are based on models derived from theoretical considerat...

  20. Proceedings of the 1993 international conference on nuclear waste management and environmental remediation. Volume 2: High level radioactive waste and spent fuel management

    SciTech Connect (OSTI)

    Ahlstroem, P.E.; Chapman, C.C.; Kohout, R.; Marek, J. [eds.

    1993-12-31T23:59:59.000Z

    This conference was held in 1993 in Prague, Czech Republic to provide a forum for exchange of state-of-the-art information on radioactive waste management. Volume 2 contains 109 papers divided into the following sections: recent developments in environmental remediation technologies; decommissioning of nuclear power reactors; environmental restoration site characterization and monitoring; decontamination and decommissioning of other nuclear facilities; prediction of contaminant migration and related doses; treatment of wastes from decontamination and decommissioning operations; management of complex environmental cleanup projects; experiences in actual cleanup actions; decontamination and decommissioning demolition technologies; remediation of obsolete sites from uranium mining and milling; ecological impacts from radioactive environmental contamination; national environmental management regulations--issues and assessments; significant issues and strategies in environmental management; acceptance criteria for very low-level radioactive wastes; processes for public involvement in environmental activities and decisions; recent experiences in public participation activities; established and emerging environmental management organizations; and economic considerations in environmental management. Individual papers have been processed separately for inclusion in the appropriate data bases.

  1. A TRANSPORTATION RISK ASSESSMENT TOOL FOR ANALYZING THE TRANSPORT OF SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE TO THE PROPOSED YUCCA MOUNTAIN REPOSITORY

    SciTech Connect (OSTI)

    NA

    2001-02-15T23:59:59.000Z

    The Yucca Mountain Draft Environmental Impact Statement (DEIS) analysis addressed the potential for transporting spent nuclear fuel and high-level radioactive waste from 77 origins for 34 types of spent fuel and high-level radioactive waste, 49,914 legal weight truck shipments, and 10,911 rail shipments. The analysis evaluated transportation over 59,250 unique shipment links for travel outside Nevada (shipment segments in urban, suburban or rural zones by state), and 22,611 links in Nevada. In addition, the analysis modeled the behavior of 41 isotopes, 1091 source terms, and used 8850 food transfer factors (distinct factors by isotope for each state). The analysis also used mode-specific accident rates for legal weight truck, rail, and heavy haul truck by state, and barge by waterway. This complex mix of data and information required an innovative approach to assess the transportation impacts. The approach employed a Microsoft{reg_sign} Access database tool that incorporated data from many sources, including unit risk factors calculated using the RADTRAN IV transportation risk assessment computer program. Using Microsoft{reg_sign} Access, the analysts organized data (such as state-specific accident and fatality rates) into tables and developed queries to obtain the overall transportation impacts. Queries are instructions to the database describing how to use data contained in the database tables. While a query might be applied to thousands of table entries, there is only one sequence of queries that is used to calculate a particular transportation impact. For example, the incident-free dose to off-link populations in a state is calculated by a query that uses route segment lengths for each route in a state that could be used by shipments, populations for each segment, number of shipments on each segment, and an incident-free unit risk factor calculated using RADTRAN IV. In addition to providing a method for using large volumes of data in the calculations, the queries provide a straight-forward means used to verify results. Another advantage of using the MS Access database was the ability to develop query hierarchies using nested queries. Calculations were broken into a series of steps, each step represented by a query. For example, the first query might calculate the number of shipment kilometers traveled through urban, rural and suburban zones for all states. Subsequent queries could join the shipment kilometers query results with another table containing unit risk factors calculated using RADTRAN IV to produce radiological impacts. Through the use of queries, impacts by origin, mode, fuel type or many other parameters can be obtained. The paper will show both the flexibility of the assessment tool and the ease it provides for verifying results.

  2. CLOSURE OF HLW TANKS PHASE 2 FULL SCALE COOLING COILS GROUT FILL DEMONSTATIONS

    SciTech Connect (OSTI)

    Hansen, E; Alex Cozzi, A

    2008-06-19T23:59:59.000Z

    This report documents the Savannah River National Laboratory (SRNL) support for the Tank Closure and Technology Development (TCTD) group's strategy for closing high level radioactive waste (HLW) tanks at the Savannah River Site (SRS). Specifically, this task addresses the ability to successfully fill intact cooling coils, presently within the HLW tanks, with grout that satisfies the fresh and cured grout requirements [1] under simulated field conditions. The overall task was divided into two phases. The first phase was the development of a grout formulation that satisfies the processing requirements for filling the HLW tank cooling coils [5]. The second phase of the task, which is documented in this report, was the filling of full scale cooling coils under simulated field conditions using the grout formulation developed in the first phase. SRS Type I tank cooling coil assembly design drawings and pressure drop calculations were provided by the Liquid Waste (LW) customer to be used as the basis for configuring the test assemblies. The current concept for closing tanks equipped with internal cooling coils is to pump grout into the coils to inhibit pathways for infiltrating water. Access to the cooling coil assemblies is through the existing supply/return manifold headers located on top of the Type I tanks. The objectives for the second phase of the testing, as stated in the Task Technical and Quality Assurance plan (TTQAP) [2], were to: (1) Perform a demonstration test to assess cooling coil grout performance in simulated field conditions, and (2) Measure relevant properties of samples prepared under simulated field conditions. SRNL led the actual work of designing, fabricating and filling two full-scale cooling coil assemblies which were performed at Clemson Engineering Technologies Laboratory (CETL) using the South Carolina University Research and Education Foundation (SCUREF) program. A statement of work (SOW) was issued to CETL [6] to perform this work.

  3. EVOLUTION OF CHEMICAL CONDITIONS AND ESTIMATED SOLUBILITY CONTROLS ON RADIONUCLIDES IN THE RESIDUAL WASTE LAYER DURING POST-CLOSURE AGING OF HIGH-LEVEL WASTE TANKS

    SciTech Connect (OSTI)

    Denham, M.; Millings, M.

    2012-08-28T23:59:59.000Z

    This document provides information specific to H-Area waste tanks that enables a flow and transport model with limited chemical capabilities to account for varying waste release from the tanks through time. The basis for varying waste release is solubilities of radionuclides that change as pore fluids passing through the waste change in composition. Pore fluid compositions in various stages were generated by simulations of tank grout degradation. The first part of the document describes simulations of the degradation of the reducing grout in post-closure tanks. These simulations assume flow is predominantly through a water saturated porous medium. The infiltrating fluid that reacts with the grout is assumed to be fluid that has passed through the closure cap and into the tank. The results are three stages of degradation referred to as Reduced Region II, Oxidized Region II, and Oxidized Region III. A reaction path model was used so that the transitions between each stage are noted by numbers of pore volumes of infiltrating fluid reacted. The number of pore volumes to each transition can then be converted to time within a flow and transport model. The bottoms of some tanks in H-Area are below the water table requiring a different conceptual model for grout degradation. For these simulations the reacting fluid was assumed to be 10% infiltrate through the closure cap and 90% groundwater. These simulations produce an additional four pore fluid compositions referred to as Conditions A through D and were intended to simulate varying degrees of groundwater influence. The most probable degradation path for the submerged tanks is Condition C to Condition D to Oxidized Region III and eventually to Condition A. Solubilities for Condition A are estimated in the text for use in sensitivity analyses if needed. However, the grout degradation simulations did not include sufficient pore volumes of infiltrating fluid for the grout to evolve to Condition A. Solubility controls for use in a flow and transport model were estimated for 27 elements in each of the chemical stages generated in the grout simulations plus local groundwater. The grout simulations were run with the initial infiltrating fluid in equilibrium with atmospheric oxygen to account for degradation of the reduction capacity of the grout. However, a lower Eh was used in pore fluids in the oxidizing conditions used to estimate solubilities to be more consistent with measured Eh values and natural systems. Solubilities of plutonium are affected by this decision, but those of other elements are not. In addition, the baseline for H-Area tanks is that they will be washed with oxalic acid prior to being filled with grout. Hence, oxalate was included in the pore fluids by assuming equilibrium with calcium oxalate. Solubility estimates were done by equilibrating a solubility controlling phase for each element with the pore fluid compositions using The Geochemist’s Workbench®. Condition B pore fluids are similar to Condition D. Therefore, solubilities for Condition B were not estimated, but assumed to be the same as in Condition D. In general solubility controlling phases were selected to bias solubilities to higher values. Several elements had no solubility controls and solubility estimates for other elements were omitted because the elements had short half-lives or were present in residual waste in very low amounts. For these it is recommended that release from the tank be instantaneous when the tank liner is breached. There is considerable uncertainty in this approach to enabling a flow and transport model to account for variable waste release. Yet, it is also flexible and requires much less computing time than a fully coupled reactive transport model. This allows some of the uncertainty to be addressed by multiple flow and transport sensitivity cases. Some of the uncertainties are addressed within this document. These include uncertainty in infiltrate composition, grout mineralogy, and disposition of certain components during the simulations. Uncertainty in the solubility estima

  4. Update to Assessment of Direct Disposal in Unsaturated Tuff of Spent Nuclear Fuel and High-Level Waste Owned by U.S. Department of Energy

    SciTech Connect (OSTI)

    P. D. Wheatley (INEEL POC); R. P. Rechard (SNL)

    1998-09-01T23:59:59.000Z

    The overall purpose of this study is to provide information and guidance to the Office of Environmental Management of the U.S. Department of Energy (DOE) about the level of characterization necessary to dispose of DOE-owned spent nuclear fuel (SNF). The disposal option modeled was codisposal of DOE SNF with defense high-level waste (DHLW). A specific goal was to demonstrate the influence of DOE SNF, expected to be minor, in a predominately commercial repository using modeling conditions similar to those currently assumed by the Yucca Mountain Project (YMP). A performance assessment (PA) was chosen as the method of analysis. The performance metric for this analysis (referred to as the 1997 PA) was dose to an individual; the time period of interest was 100,000 yr. Results indicated that cumulative releases of 99Tc and 237Np (primary contributors to human dose) from commercial SNF exceed those of DOE SNF both on a per MTHM and per package basis. Thus, if commercial SNF can meet regulatory performance criteria for dose to an individual, then the DOE SNF can also meet the criteria. This result is due in large part to lower burnup of the DOE SNF (less time for irradiation) and to the DOE SNF's small percentage of the total activity (1.5%) and mass (3.8%) of waste in the potential repository. Consistent with the analyses performed for the YMP, the 1997 PA assumed all cladding as failed, which also contributed to the relatively poor performance of commercial SNF compared to DOE SNF.

  5. Property/composition relationships for Hanford high-level waste glasses melting at 1150{degrees}C volume 2: Chapters 12-16 and appendices A-K

    SciTech Connect (OSTI)

    Hrma, P.R.; Piepel, G.F.

    1994-12-01T23:59:59.000Z

    A Composition Variation Study (CVS) is being performed within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) project in support of a future high-level nuclear waste vitrification plant at the Hanford site in Washington. From 1989 to 1994, over 120 nonradioactive glasses were melted and properties measured in five statistically-designed experimental phases. Glass composition is represented by the 10 components SiO{sub 2}, B{sub 2}O{sub 3}, ZrO{sub 2}, Na{sub 2}O, Li{sub 2}O, CaO, MgO, and Others (all remaining components). The properties measured include viscosity ({eta}), electrical conductivity ({epsilon}), glass transition temperature (T{sub g}), thermal expansion of solid glass ({alpha}{sub s}) and molten glass ({alpha}{sub m}), crystallinity (quenched and canister centerline cooled glasses), liquidus temperature (T{sub L}), durability based on normalized elemental releases from the Materials Characterization Center-1 28-day dissolution test (MCC-1, r{sub mi}) and the 7-day Product Consistency Test (PCT, r{sub pi}), and solution pHs from MCC-1 and PCT. Amorphous phase separation was also evaluated. Empirical first- and second-order mixture models were fit using the CVS data to relate the various properties to glass composition. Equations for calculating the uncertainty associated with property values predicted by the models were also developed. The models were validated using both internal and external data. Other modeling approaches (e.g., non-bridging oxygen, free energy of hydration, phase-equilibria T{sub L}) were investigated for specific properties. A preliminary Qualified Composition Region was developed to identify glass compositions with high confidence of being processable in a melter and meeting waste form acceptance criteria.

  6. Property/composition relationships for Hanford high-level waste glasses melting at 115{degrees}C volume 1: Chapters 1-11

    SciTech Connect (OSTI)

    Hrma, P.R.; Piepel, G.F.

    1994-12-01T23:59:59.000Z

    A Composition Variation study (CVS) is being performed within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) project in support of a future high-level nuclear waste vitrification plant at the Hanford site in Washington. From 1989 to 1994, over 120 nonradioactive glasses were melted and properties measured in five statistically-designed experimental phases. Glass composition is represented by the 10 components SiO{sub 2}, B{sub 2}O{sub 3}, Al{sub 2}O{sub 3}, Fe{sub 2}O{sub 3}, ZrO{sub 2}, Na{sub 2}O, Li{sub 2}O, CaO, MgO, and Others (all remaining components). The properties measured include viscosity ({eta}), electrical conductivity ({epsilon}), glass transition temperature (T{sub g} ), thermal expansion of solid glass ({alpha}{sub s}) and molten glass ({alpha}{sub m}), crystallinity (quenched and canister centerline cooled glasses), liquidus temperature (T{sub L}), durability based on normalized elemental releases from the Materials Characterization Center-1 28-day dissolution test (MCC-1, r{sub mi}) and the 7-day Product Consistency Test (PCT, r{sub pi}), and solution pHs from MCC-1 and PCT. Amorphous phase separation was also evaluated. Empirical first- and second-order mixture models were fit using the CVS data to relate the various properties to glass composition. Equations for calculating the uncertainty associated with property values predicted by the models were also developed. The models were validated using both internal and external data. Other modeling approaches (e.g., non-bridging oxygen, free energy of hydration, phase-equilibria T{sub L}) were investigated for specific properties. A preliminary Qualified Composition Region was developed to identify glass compositions with high confidence of being processable in a melter and meeting waste form acceptance criteria.

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

    SciTech Connect (OSTI)

    Vince Maio

    2014-08-01T23:59:59.000Z

    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.

  8. Illustration of sampling-based approaches to the calculation of expected dose in performance assessments for the proposed high level radioactive waste repository at Yucca Mountain, Nevada.

    SciTech Connect (OSTI)

    Helton, Jon Craig (Arizona State University, Tempe, AZ); Sallaberry, Cedric J. PhD. (.; .)

    2007-04-01T23:59:59.000Z

    A deep geologic repository for high level radioactive waste is under development by the U.S. Department of Energy at Yucca Mountain (YM), Nevada. As mandated in the Energy Policy Act of 1992, the U.S. Environmental Protection Agency (EPA) has promulgated public health and safety standards (i.e., 40 CFR Part 197) for the YM repository, and the U.S. Nuclear Regulatory Commission has promulgated licensing standards (i.e., 10 CFR Parts 2, 19, 20, etc.) consistent with 40 CFR Part 197 that the DOE must establish are met in order for the YM repository to be licensed for operation. Important requirements in 40 CFR Part 197 and 10 CFR Parts 2, 19, 20, etc. relate to the determination of expected (i.e., mean) dose to a reasonably maximally exposed individual (RMEI) and the incorporation of uncertainty into this determination. This presentation describes and illustrates how general and typically nonquantitive statements in 40 CFR Part 197 and 10 CFR Parts 2, 19, 20, etc. can be given a formal mathematical structure that facilitates both the calculation of expected dose to the RMEI and the appropriate separation in this calculation of aleatory uncertainty (i.e., randomness in the properties of future occurrences such as igneous and seismic events) and epistemic uncertainty (i.e., lack of knowledge about quantities that are poorly known but assumed to have constant values in the calculation of expected dose to the RMEI).

  9. Thermo-Hydrological-Mechanical Analysis of a Clay Barrier for Radioactive Waste Isolation: Probabilistic Calibration and Advanced Modeling

    E-Print Network [OSTI]

    Dontha, Lakshman

    2012-07-16T23:59:59.000Z

    Sanchez Co-Chairs of Committee, Zenon Medina-Cetina Committee Member, Frederick Chester Head of Department, John Niedzwecki May 2012 Major Subject: Civil Engineering iii ABSTRACT Thermo... Committee, Dr. Marcelo Sanchez Dr. Zenon Medina-Cetina The engineered barrier system is a basic element in the design of repository to isolate high level radioactive waste (HLW...

  10. Identification of potential transuranic waste tanks at the Hanford Site

    SciTech Connect (OSTI)

    Colburn, R.P.

    1995-05-05T23:59:59.000Z

    The purpose of this document is to identify potential transuranic (TRU) material among the Hanford Site tank wastes for possible disposal at the Waste Isolation Pilot Plant (WIPP) as an alternative to disposal in the high-level waste (HLW) repository. Identification of such material is the initial task in a trade study suggested in WHC-EP-0786, Tank Waste Remediation System Decisions and Risk Assessment (Johnson 1994). The scope of this document is limited to the identification of those tanks that might be segregated from the HLW for disposal as TRU, and the bases for that selection. It is assumed that the tank waste will be washed to remove soluble inert material for disposal as low-level waste (LLW), and the washed residual solids will be vitrified for disposal. The actual recommendation of a disposal strategy for these materials will require a detailed cost/benefit analysis and is beyond the scope of this document.

  11. HLW system plan - revision 2

    SciTech Connect (OSTI)

    Not Available

    1994-01-14T23:59:59.000Z

    The projected ability of the Tank Farm to support DWPF startup and continued operation has diminished somewhat since revision 1 of this Plan. The 13 month delay in DWPF startup, which actually helps the Tank Farm condition in the near term, was more than offset by the 9 month delay in ITP startup, the delay in the Evaporator startups and the reduction to Waste Removal funding. This Plan does, however, describe a viable operating strategy for the success of the HLW System and Mission, albeit with less contingency and operating flexibility than in the past. HLWM has focused resources from within the division on five near term programs: The three evaporator restarts, DWPF melter heatup and completion of the ITP outage. The 1H Evaporator was restarted 12/28/93 after a 9 month shutdown for an extensive Conduct of Operations upgrade. The 2F and 2H Evaporators are scheduled to restart 3/94 and 4/94, respectively. The RHLWE startup remains 11/17/97.

  12. Characteristics of potential repository wastes. Volume 1

    SciTech Connect (OSTI)

    Not Available

    1992-07-01T23:59:59.000Z

    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.

  13. 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.

    1993-02-01T23:59:59.000Z

    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.

  14. EIS-0250-S1: Final Supplemental Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada

    Broader source: Energy.gov [DOE]

    The Proposed Action defined in the Yucca Mountain FEIS is to construct, operate, monitor, and eventually close a geologic repository at Yucca Mountain to dispose of spent nuclear fuel and high-level radioactive waste. The Proposed Action includes transportation of these materials from commercial and DOE sites to the repository.

  15. Final Systems Development Report for the Clark County Socioeconomic Impact Assessment of the Proposed High-Level Nuclear Waste Repository at Yucca Mountain, NV

    SciTech Connect (OSTI)

    NONE

    1992-06-18T23:59:59.000Z

    The Systems Development Report represents the third major step in the Clark County Socioeconomic Impact Assessment of the Proposed High-Level Nuclear Waste Repository at Yucca Mound Nevada. The first of these steps was to forge a Research Design that would serve as a guide for the overall research process. The second step was the construction of the Base Case, the purpose of which was to describe existing conditions in Clark County in the specified analytic areas of Economic-Demographic/Fiscal, Emergency Planning and Management, Transportation and Sociocultural analysis. The base case description will serve as a basis for assessing changes in these topic areas that might result from the Yucca Mountain project. These changes will be assessed by analyzing conditions with and without repository development in the county. Prior to performing such assessments, however, the snapshot type of data found in the base case must be operationalized or systematized to allow for more dynamic data utilization. In other words, a data system that can be used to analyze the consequences of the introduction of different variables (or variable values) in the Clark County context must be constructed. Such a system must be capable of being updated through subsequent data collection and monitoring efforts to both provide a rolling base case and supply information necessary to construct trend analyses. For example, during the Impact Assessment phase of the study process, the without repository analysis is accomplished by analyzing growth for the county given existing conditions and likely trends. These data are then compared to the with Yucca Mountain project conditions anticipated for the county. Similarly, once the emergency planning management and response needs associated with the repository are described, these needs will be juxtaposed against existing (and various future) capacity(ies) in order to determine the nature and magnitude of impacts in this analytic area. Analogous tasks will be performed for the other analytic areas detailed in the Base Case and outlined below.

  16. Final base case community analysis: Indian Springs, Nevada for the Clark County socioeconomic impact assessment of the proposed high- level nuclear waste repository at Yucca Mountain, Nevada

    SciTech Connect (OSTI)

    NONE

    1992-06-18T23:59:59.000Z

    This document provides a base case description of the rural Clark County community of Indian Springs in anticipation of change associated with the proposed high-level nuclear waste repository at Yucca Mountain. As the community closest to the proposed site, Indian Springs may be seen by site characterization workers, as well as workers associated with later repository phases, as a logical place to live. This report develops and updates information relating to a broad spectrum of socioeconomic variables, thereby providing a `snapshot` or `base case` look at Indian Springs in early 1992. With this as a background, future repository-related developments may be analytically separated from changes brought about by other factors, thus allowing for the assessment of the magnitude of local changes associated with the proposed repository. Given the size of the community, changes that may be considered small in an absolute sense may have relatively large impacts at the local level. Indian Springs is, in many respects, a unique community and a community of contrasts. An unincorporated town, it is a small yet important enclave of workers on large federal projects and home to employees of small- scale businesses and services. It is a rural community, but it is also close to the urbanized Las Vega Valley. It is a desert community, but has good water resources. It is on flat terrain, but it is located within 20 miles of the tallest mountains in Nevada. It is a town in which various interest groups diverge on issues of local importance, but in a sense of community remains an important feature of life. Finally, it has a sociodemographic history of both surface transience and underlying stability. If local land becomes available, Indian Springs has some room for growth but must first consider the historical effects of growth on the town and its desired direction for the future.

  17. Geology of the Yucca Mountain Region, Chapter in Stuckless, J.S., ED., Yucca Mountain, Nevada - A Proposed Geologic Repository for High-Level Radioactive Waste

    SciTech Connect (OSTI)

    J.S. Stuckless; D. O'Leary

    2006-09-25T23:59:59.000Z

    Yucca Mountain has been proposed as the site for the Nation's first geologic repository for high-level radioactive waste. This chapter provides the geologic framework for the Yucca Mountain region. The regional geologic units range in age from late Precambrian through Holocene, and these are described briefly. Yucca Mountain is composed dominantly of pyroclastic units that range in age from 11.4 to 15.2 Ma. The proposed repository would be constructed within the Topopah Spring Tuff, which is the lower of two major zoned and welded ash-flow tuffs within the Paintbrush Group. The two welded tuffs are separated by the partly to nonwelded Pah Canyon Tuff and Yucca Mountain Tuff, which together figure prominently in the hydrology of the unsaturated zone. The Quaternary deposits are primarily alluvial sediments with minor basaltic cinder cones and flows. Both have been studied extensively because of their importance in predicting the long-term performance of the proposed repository. Basaltic volcanism began about 10 Ma and continued as recently as about 80 ka with the eruption of cones and flows at Lathrop Wells, approximately 10 km south-southwest of Yucca Mountain. Geologic structure in the Yucca Mountain region is complex. During the latest Paleozoic and Mesozoic, strong compressional forces caused tight folding and thrust faulting. The present regional setting is one of extension, and normal faulting has been active from the Miocene through to the present. There are three major local tectonic domains: (1) Basin and Range, (2) Walker Lane, and (3) Inyo-Mono. Each domain has an effect on the stability of Yucca Mountain.

  18. An Overview of Project Planning for Hot-Isostatic Pressure Treatment of High-Level Waste Calcine for the Idaho Cleanup Project - 12289

    SciTech Connect (OSTI)

    Nenni, Joseph A.; Thompson, Theron J. [CH2M-WG Idaho, LLC, Idaho Cleanup Project, Idaho Falls, Idaho 83403 (United States)

    2012-07-01T23:59:59.000Z

    The Calcine Disposition Project is responsible for retrieval, treatment by hot-isostatic pressure, packaging, and disposal of highly radioactive calcine stored at the Idaho Nuclear Technology and Engineering Center at the Idaho National Laboratory Site in southeast Idaho. In the 2009 Amended Record of Decision: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement the Department of Energy documented the selection of hot-isostatic pressure as the technology to treat the calcine. The Record of Decision specifies that the treatment results in a volume-reduced, monolithic waste form suitable for transport outside of Idaho by a target date of December 31, 2035. That target date is specified in the 1995 Idaho Settlement Agreement to treat and prepare the calcine for transport out of Idaho in exchange for allowing storage of Navy spent nuclear fuel at the INL Site. The project is completing the design of the calcine-treatment process and facility to comply with Record of Decision, Settlement Agreement, Idaho Department of Environmental Quality, and Department of Energy requirements. A systems engineering approach is being used to define the project mission and requirements, manage risks, and establish the safety basis for decision making in compliance with DOE O 413.3B, 'Program and Project Management for the Acquisition of Capital Assets'. The approach draws heavily on 'design-for-quality' tools to systematically add quality, predict design reliability, and manage variation in the earliest possible stages of design when it is most efficient. Use of these tools provides a standardized basis for interfacing systems to interact across system boundaries and promotes system integration on a facility-wide basis. A mass and energy model was developed to assist in the design of process equipment, determine material-flow parameters, and estimate process emissions. Data generated from failure modes and effects analysis and reliability, availability, maintainability, and inspectability analysis were incorporated into a time and motion model to validate and verify the capability to complete treatment of the calcine within the required schedule. The Calcine Disposition Project systems engineering approach, including use of industry-proven design-for-quality tools and quantitative assessment techniques, has strengthened the project's design capability to meet its intended mission in a safe, cost-effective, and timely manner. Use of these tools has been particularly helpful to the project in early design planning to manage variation; improve requirements and high-consequence risk management; and more effectively apply alternative, interface, failure mode, RAMI, and time and motion analyses at the earliest possible stages of design when their application is most efficient and cost effective. The project is using these tools to design and develop HIP treatment of highly radioactive calcine to produce a volume-reduced, monolithic waste form with immobilization of hazardous and radioactive constituents. (authors)

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

    SciTech Connect (OSTI)

    Deffenbaugh, M.L.

    1997-09-04T23:59:59.000Z

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

  20. 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

    2011-01-05T23:59:59.000Z

    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.

  1. The Development of an Effective Transportation Risk Assessment Model for Analyzing the Transport of Spent Fuel and High-Level Radioactive Waste to the Proposed Yucca Mountain Repository

    SciTech Connect (OSTI)

    McSweeney; Thomas; Winnard; Ross; Steven B.; Best; Ralph E.

    2001-02-06T23:59:59.000Z

    Past approaches for assessing the impacts of transporting spent fuel and high-level radioactive waste have not been effectively implemented or have used relatively simple approaches. The Yucca Mountain Draft Environmental Impact Statement (DEIS) analysis considers 83 origins, 34 fuel types, 49,914 legal weight truck shipments, 10,911 rail shipments, consisting of 59,250 shipment links outside Nevada (shipment kilometers and population density pairs through urban, suburban or rural zones by state), and 22,611 shipment links in Nevada. There was additional complexity within the analysis. The analysis modeled the behavior of 41 isotopes, 1091 source terms, and used 8850 food transfer factors (distinct factors by isotope for each state). The model also considered different accident rates for legal weight truck, rail, and heavy haul truck by state, and barge by waterway. To capture the all of the complexities of the transportation analysis, a Microsoft{reg_sign} Access database was created. In the Microsoft{reg_sign} Access approach the data is placed in individual tables and equations are developed in queries to obtain the overall impacts. While the query might be applied to thousands of table entries, there is only one equation for a particular impact. This greatly simplifies the validation effort. Furthermore, in Access, data in tables can be linked automatically using query joins. Another advantage built into MS Access is nested queries, or the ability to develop query hierarchies. It is possible to separate the calculation into a series of steps, each step represented by a query. For example, the first query might calculate the number of shipment kilometers traveled through urban, rural and suburban zones for all states. Subsequent queries could join the shipment kilometers query results with another table containing the state and mode specific accident rate to produce accidents by state. One of the biggest advantages of the nested queries is in validation. Temporarily restricting the query to one origin, one shipment, or one state and validating that the query calculation is returning the expected result allows simple validation. The paper will show the flexibility of the assessment tool to consider a wide variety of impacts. Through the use of pre-designed queries, impacts by origin, mode, fuel type or many other parameters can be obtained.

  2. EIS-0250: Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada

    Broader source: Energy.gov [DOE]

    This EIS analyzes DOE's proposed action to construct, operate, monitor, and eventually close a geologic repository at Yucca Mountain  for the disposal of spent nuclear fuel and high-level...

  3. Combined Waste Form Cost Trade Study

    SciTech Connect (OSTI)

    Dirk Gombert; Steve Piet; Timothy Trickel; Joe Carter; John Vienna; Bill Ebert; Gretchen Matthern

    2008-11-01T23:59:59.000Z

    A new generation of aqueous nuclear fuel reprocessing, now in development under the auspices of the DOE Office of Nuclear Energy (NE), separates fuel into several fractions, thereby partitioning the wastes into groups of common chemistry. This technology advance enables development of waste management strategies that were not conceivable with simple PUREX reprocessing. Conventional wisdom suggests minimizing high level waste (HLW) volume is desirable, but logical extrapolation of this concept suggests that at some point the cost of reducing volume further will reach a point of diminishing return and may cease to be cost-effective. This report summarizes an evaluation considering three groupings of wastes in terms of cost-benefit for the reprocessing system. Internationally, the typical waste form for HLW from the PUREX process is borosilicate glass containing waste elements as oxides. Unfortunately several fission products (primarily Mo and the noble metals Ru, Rh, Pd) have limited solubility in glass, yielding relatively low waste loading, producing more glass, and greater disposal costs. Advanced separations allow matching the waste form to waste stream chemistry, allowing the disposal system to achieve more optimum waste loading with improved performance. Metals can be segregated from oxides and each can be stabilized in forms to minimize the HLW volume for repository disposal. Thus, a more efficient waste management system making the most effective use of advanced waste forms and disposal design for each waste is enabled by advanced separations and how the waste streams are combined. This trade-study was designed to juxtapose a combined waste form baseline waste treatment scheme with two options and to evaluate the cost-benefit using available data from the conceptual design studies supported by DOE-NE.

  4. Accelerated Weathering of High-Level and Plutonium-bearing Lanthanide...

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

    Lanthanide Borosilicate Waste Glasses under Hydraulically Unsaturated Accelerated Weathering of High-Level and Plutonium-bearing Lanthanide Borosilicate Waste...

  5. EIS-0250-S2: Supplemental EIS for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada- Nevada Rail Transportation Corridor

    Broader source: Energy.gov [DOE]

    This SEIS is to evaluate the potential environmental impacts of constructing and operating a railroad for shipments of spent nuclear fuel and high-level radioactive waste from an existing rail line in Nevada to a geologic repository at Yucca Mountain. The purpose of the evaluation is to assist the Department in deciding whether to construct and operate a railroad in Nevada, and if so, in which corridor and along which specific alignment within the selected corridor.

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

    SciTech Connect (OSTI)

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

    2012-01-12T23:59:59.000Z

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

  7. Cementitious waste option scoping study report

    SciTech Connect (OSTI)

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

    1998-02-01T23:59:59.000Z

    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.

  8. Evaluation and compilation of DOE waste package test data: Biannual report, August 1986-January 1987

    SciTech Connect (OSTI)

    Interrante, C.; Escalante, E.; Fraker, A.; Harrison, S.; Shull, R.; Linzer, M.; Ricker, R.; Ruspi, J.

    1987-10-01T23:59:59.000Z

    This report summarizes results of the National Bureau of Standards (NBS) evaluations of Department of Energy (DOE) activities 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. Technical discussions are given for the corrosion of metals proposed for the canister, particularly carbon and stainless steels, and copper. In the section on tuff, the current level of understanding of several canister materials is questioned. Within the Basalt Waste Isolation Project (BWIP) section, discussions are given on problems concerning groundwater, materials for use in the metallic overpack, and diffusion through the packing. For the proposed salt site, questions are raised on the work on both ASTM A216 Steel and Ti-Code 12. NBS work related to the vitrification of HLW borosilicate glass at the West Valley Demonstration Project (WVDP) and the Defense Waste Processing Facility (DWPF) is covered. NBS reviews of selected DOE technical reports and a summary of current waste-package activities of the Materials Characterization Center (MCC) is presented. Using a database management system, a computerized database for storage and retrieval of reviews and evaluations of HLW data has been developed and is described. 17 refs., 2 figs., 2 tabs.

  9. An Assessment of the Stability and the Potential for In-Situ Synthesis of Regulated Organic Compounds in High Level Radioactive Waste Stored at Hanford, Richland, Washington

    SciTech Connect (OSTI)

    Wiemers, K.D.; Babad, H.; Hallen, R.T.; Jackson, L.P.; Lerchen, M.E.

    1999-01-04T23:59:59.000Z

    The stability assessment examined 269 non-detected regulated compounds, first seeking literature references of the stability of the compounds, then evaluating each compound based upon the presence of functional groups using professional judgment. Compounds that could potentially survive for significant periods in the tanks (>1 year) were designated as stable. Most of the functional groups associated with the regulated organic compounds were considered unstable under tank waste conditions. The general exceptions with respect to functional group stability are some simple substituted aromatic and polycyclic aromatic compounds that resist oxidation and the multiple substituted aliphatic and aromatic halides that hydrolyze or dehydrohalogenate slowly under tank waste conditions. One-hundred and eighty-one (181) regulated, organic compounds were determined as likely unstable in the tank waste environment.

  10. Workshop on the source term for radionuclide migration from high-level waste or spent nuclear fuel under realistic repository conditions: proceedings

    SciTech Connect (OSTI)

    Hunter, T.O.; Muller, A.B. (eds.)

    1985-07-01T23:59:59.000Z

    Sixteen papers were presented at the workshop. The fourteen full-length papers included in the proceedings were processed separately. Only abstracts were included for the following two papers: Data Requirements Based on Performance Assessment Analyses of Conceptual Waste Packages in Salt Repositories, and The Potential Effects of Radiation on the Source Term in a Salt Repository. (LM)

  11. Social impacts of hazardous and nuclear facilities and events: Implications for Nevada and the Yucca Mountain high-level nuclear waste repository; [Final report

    SciTech Connect (OSTI)

    Freudenburg, W.R. [Wisconsin Univ., Madison, WI (United States); Carter, L.F.; Willard, W. [Washington State Univ., Pullman, WA (United States); Lodwick, D.G. [Miami Univ., Oxford, OH (United States); Hardert, R.A. [Arizona State Univ., Tempe, AZ (United States); Levine, A.G. [State Univ. of New York, Buffalo, NY (United States). Dept. of Sociology; Kroll-Smith, S. [New Orleans Univ., LA (United States); Couch, S.R. [Pennsylvania State Univ., University Park, PA (United States); Edelstein, M.R. [Ramapo College, Mahwah, NJ (United States)

    1992-05-01T23:59:59.000Z

    Social impacts of a nuclear waste repository are described. Various case studies are cited such as Rocky Flats Plant, the Feed Materials Production Center, and Love Canal. The social impacts of toxic contamination, mitigating environmental stigma and loss of trust are also discussed.

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

    SciTech Connect (OSTI)

    Swanson, J.L.

    1993-09-01T23:59:59.000Z

    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.

  13. Secondary Waste Cast Stone Waste Form Qualification Testing Plan

    SciTech Connect (OSTI)

    Westsik, Joseph H.; Serne, R. Jeffrey

    2012-09-26T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, WA (United States); Rodriguez, C. A. [Pacific Northwest National Laboratory, Richland, WA (United States); Matyas, J. [Pacific Northwest National Laboratory, Richland, WA (United States); Owen, A. T. [Pacific Northwest National Laboratory, Richland, WA (United States); Jansik, D. P. [Pacific Northwest National Laboratory, Richland, WA (United States); Lang, J. B. [Pacific Northwest National Laboratory, Richland, WA (United States)

    2012-11-12T23:59:59.000Z

    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.

  15. XRD, Electron Microscopy and Vibrational Spectroscopy Characterization of Simulated SB6 HLW Glasses - 13028

    SciTech Connect (OSTI)

    Stefanovsky, S.V. [SIA Radon, 7th Rostovskii lane 2/14, Moscow 119121 (Russian Federation) [SIA Radon, 7th Rostovskii lane 2/14, Moscow 119121 (Russian Federation); Institute of Physical Chemistry and Electrochemistry RAS, Leninskii av. 31, Moscow 119991 (Russian Federation); Nikonov, B.S.; Omelianenko, B.I. [Institute of Geology of Ore Deposits, Petrography, Mineralogy and Geochemistry RAS, Staromonetniy lane 35, Moscow 100117 (Russian Federation)] [Institute of Geology of Ore Deposits, Petrography, Mineralogy and Geochemistry RAS, Staromonetniy lane 35, Moscow 100117 (Russian Federation); Choi, A.; Marra, J.C. [Savannah River National Laboratory, Building 773A, Aiken 29808 (United States)] [Savannah River National Laboratory, Building 773A, Aiken 29808 (United States)

    2013-07-01T23:59:59.000Z

    Sample glasses have been made using SB6 high level waste (HLW) simulant (high in both Al and Fe) with 12 different frit compositions at a constant waste loading of 36 wt.%. As follows from X-ray diffraction (XRD) and optical and scanning electron microscopy (SEM) data, all the samples are composed of primarily glass and minor concentration of spinel phases which form both isometric grains and fine cubic (?1 ?m) crystals. Infrared spectroscopy (IR) spectra of all the glasses within the range of 400-1600 cm{sup -1} consist of the bands due to stretching and bending modes in silicon-oxygen, boron-oxygen, aluminum-oxygen and iron-oxygen structural groups. Raman spectra showed that for the spectra of all the glasses within the range of 850-1200 cm{sup -1} the best fit is achieved by suggestion of overlapping of three major components with maxima at 911-936 cm{sup -1}, 988-996 cm{sup -1} and 1020-1045 cm{sup -1}. The structural network is primarily composed of metasilicate chains and rings with embedded AlO{sub 4} and FeO{sub 4} tetrahedra. Major BO{sub 4} tetrahedra and BO{sub 3} triangles form complex borate units and are present as separate constituents. (authors)

  16. Melter Throughput Enhancements for High-Iron HLW

    SciTech Connect (OSTI)

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

    2012-12-26T23:59:59.000Z

    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.

  17. 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)

    2012-12-13T23:59:59.000Z

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

  18. 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.

    2012-04-02T23:59:59.000Z

    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.

  19. PILOT-SCALE TEST RESULTS OF A THIN FILM EVAPORATOR SYSTEM FOR MANAGEMENT OF LIQUID HIGH-LEVEL WASTES AT THE HANFORD SITE WASHINGTON USA -11364

    SciTech Connect (OSTI)

    CORBETT JE; TEDESCH AR; WILSON RA; BECK TH; LARKIN J

    2011-02-14T23:59:59.000Z

    A modular, transportable evaporator system, using thin film evaporative technology, is planned for deployment at the Hanford radioactive waste storage tank complex. This technology, herein referred to as a wiped film evaporator (WFE), will be located at grade level above an underground storage tank to receive pumped liquids, concentrate the liquid stream from 1.1 specific gravity to approximately 1.4 and then return the concentrated solution back into the tank. Water is removed by evaporation at an internal heated drum surface exposed to high vacuum. The condensed water stream will be shipped to the site effluent treatment facility for final disposal. This operation provides significant risk mitigation to failure of the aging 242-A Evaporator facility; the only operating evaporative system at Hanford maximizing waste storage. This technology is being implemented through a development and deployment project by the tank farm operating contractor, Washington River Protection Solutions (WRPS), for the Office of River Protection/Department of Energy (ORPIDOE), through Columbia Energy and Environmental Services, Inc. (Columbia Energy). The project will finalize technology maturity and install a system at one of the double-shell tank farms. This paper summarizes results of a pilot-scale test program conducted during calendar year 2010 as part of the ongoing technology maturation development scope for the WFE.

  20. THE EFFECT OF THE PRESENCE OF OZONE ON THE LOWER FLAMMABILITY LIMIT OF HYDROGEN IN VESSELS CONTAINING SAVANNAH RIVER SITE HIGH LEVEL WASTE

    SciTech Connect (OSTI)

    Sherburne, C.

    2012-01-12T23:59:59.000Z

    The Enhanced Chemical Cleaning (ECC) process uses ozone to effect the oxidation of metal oxalates produced during the dissolution of sludge in the Savannah River Site (SRS) waste tanks. The ozone reacts with the metal oxalates to form metal oxide and hydroxide precipitants, and the CO{sub 2}, O{sub 2}, H{sub 2}O and any unreacted O{sub 3} gases are discharged into the vapor space. In addition to the non-radioactive metals in the waste, however, the SRS radioactive waste also contains a variety of radionuclides, hence, hydrogen gas is also present in the vapor space of the ECC system. Because hydrogen is flammable, the impact of this resultant gas stream on the Lower Flammability Limit (LFL) of hydrogen must be understood for all possible operating scenarios of both normal and off-normal situations, with particular emphasis at the elevated temperatures and pressures of the typical ECC operating conditions. Oxygen is a known accelerant in combustion reactions, but while there are data associated with the behavior of hydrogen/oxygen environments, recent, relevant studies addressing the effect of ozone on the flammability limit of hydrogen proved scarce. Further, discussions with industry experts verified the absence of data in this area and indicated that laboratory testing, specific to defined operating parameters, was needed to comprehensively address the issue. Testing was thus designed and commissioned to provide the data necessary to support safety related considerations for the ECC process. A test matrix was developed to envelope the bounding conditions considered credible during ECC processing. Each test consists of combining a gas stream of high purity hydrogen with a gas stream comprised of a specified mixture of ozone and oxygen in a temperature and pressure regulated chamber such that the relative compositions of the two streams are controlled. The gases are then stirred to obtain a homogeneous mixture and ignition attempted by applying 10J of energy to a fuse wire. A gas combination is considered flammable when a pressure rise of 7% of the initial absolute pressure is observed. The specified testing methodology is consistent with guidelines established in ASTM E-918-83 (2005) 'Standard Practices for Determining Limits of Flammability of Chemicals at Elevated Temperature and Pressure'.

  1. 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.

    1988-08-01T23:59:59.000Z

    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.

  2. Chemistry of application of calcination/dissolution to the Hanford tank waste inventory

    SciTech Connect (OSTI)

    Delegard, C.H.; Elcan, T.D.; Hey, B.E.

    1994-05-01T23:59:59.000Z

    Approximately 330,000 metric tons of sodium-rich radioactive waste originating from separation of plutonium from irradiated uranium fuel are stored in underground tanks at the Hanford Site in Washington State. Fractionation of the waste into low-level waste (LLW) and high-level waste (HLW) streams is envisioned via partial water dissolution and limited radionuclide extraction operations. Under optimum conditions, LLW would contain most of the chemical bulk while HLW would contain virtually all of the transuranic and fission product activity. Calcination at around 850 C, followed by water dissolution, has been proposed as an alternative initial treatment of Hanford Site waste to improve waste dissolution and the envisioned LLW/HLW split. Results of literature and laboratory studies are reported on the application of calcination/dissolution (C/D) to the fractionation of the Hanford Site tank waste inventory. Both simulated and genuine Hanford Site waste materials were used in the lab tests. To evaluation confirmed that C/D processing reduced the amount of several components from the waste. The C/D dissolutions of aluminum and chromium allow redistribution of these waste components from the HLW to the LLW fraction. Comparisons of simple water-washing with C/D processing of genuine Hanford Site waste are also reported based on material (radionuclide and chemical) distributions to solution and solid residue phases. The lab results show that C/D processing yielded superior dissolution of aluminum and chromium sludges compared to simple water dissolution. 57 refs., 26 figs., 18 tabs.

  3. Summary - WTP HLW Waste Vitrification Facility

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNGInternational EnergyCommittee onGASRainey STAR Center |Product EM wa in BuilW

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

    Office of Environmental Management (EM)

    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...

  5. An international initiative on long-term behavior of high-level...

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

    international initiative on long-term behavior of high-level nuclear waste glass. An international initiative on long-term behavior of high-level nuclear waste glass. Abstract:...

  6. THERMAL ANALYSIS OF WASTE GLASS MELTER FEEDS

    SciTech Connect (OSTI)

    KRUGER AA; HRMA PR; POKORNY R; PIERCE DA

    2011-10-21T23:59:59.000Z

    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. Direct cementitious waste option study report

    SciTech Connect (OSTI)

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

    1998-02-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Meacham, J.E.

    1996-10-03T23:59:59.000Z

    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.

  9. Survey of National Programs for Managing High-Level Radioactive

    E-Print Network [OSTI]

    Survey of National Programs for Managing High-Level Radioactive Waste and Spent Nuclear Fuel-Level Radioactive Waste and Spent Nuclear Fuel A Report to Congress and the Secretary of Energy October 2009 #12 Safety (Germany) Peter De Preter: National Agency for Radioactive Waste and Enriched Fissile Materials

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

    SciTech Connect (OSTI)

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

    2013-09-30T23:59:59.000Z

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

  11. High Level Waste ManagemenfDivision ..

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

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  12. High Level Waste Corporate Board Charter

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG | Department ofHTS Cable Projects HTSSeparationHelping to Finance

  13. 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.

    2012-10-19T23:59:59.000Z

    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.

  14. Secondary Waste Forms and Technetium Management

    Office of Environmental Management (EM)

    Secondary Waste Forms and Technetium Management Joseph H. Westsik, Jr. Pacific Northwest National Laboratory EM HLW Corporate Board Meeting November 18, 2010 What are Secondary...

  15. HLW-OVP-96 C

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-SeriesFlickr FlickrGuided Self-Assembly of GoldHAWCHIGS flux4-00n High Level6 C

  16. Waste Treatment Plant - 12508

    SciTech Connect (OSTI)

    Harp, Benton; Olds, Erik [US DOE (United States)

    2012-07-01T23:59:59.000Z

    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)

  17. 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.

    1994-12-01T23:59:59.000Z

    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.

  18. Tank waste remediation system retrieval and disposal mission initial updated baseline summary

    SciTech Connect (OSTI)

    Swita, W.R.

    1998-01-09T23:59:59.000Z

    This document provides a summary of the Tank Waste Remediation System (TWRS) Retrieval and Disposal Mission Initial Updated Baseline (scope, schedule, and cost), developed to demonstrate Readiness-to-Proceed (RTP) in support of the TWRS Phase 1B mission. This Updated Baseline is the proposed TWRS plan to execute and measure the mission work scope. This document and other supporting data demonstrate that the TWRS Project Hanford Management Contract (PHMC) team is prepared to fully support Phase 1B by executing the following scope, schedule, and cost baseline activities: Deliver the specified initial low-activity waste (LAW) and high-level waste (HLW) feed batches in a consistent, safe, and reliable manner to support private contractors` operations starting in June 2002; Deliver specified subsequent LAW and HLW feed batches during Phase 1B in a consistent, safe, and reliable manner; Provide for the interim storage of immobilized HLW (IHLW) products and the disposal of immobilized LAW (ILAW) products generated by the private contractors; Provide for disposal of byproduct wastes generated by the private contractors; and Provide the infrastructure to support construction and operations of the private contractors` facilities.

  19. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM - 2011

    SciTech Connect (OSTI)

    West, B.; Waltz, R.

    2012-06-21T23:59:59.000Z

    Aqueous radioactive wastes from Savannah River Site (SRS) separations and vitrification processes are contained in large underground carbon steel tanks. Inspections made during 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.

  20. E-Print Network 3.0 - actual hanford high-level Sample Search...

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

    three major facilities are planned: a pretreatment facility, a high-level... -shell tanks) that contain millions of liters of high-level liquid waste. The 400 Area is...

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

    SciTech Connect (OSTI)

    Ramsey, William Gene

    2013-08-15T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2011-09-12T23:59:59.000Z

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

  3. Evaluation and compilation of DOE waste package test data; Volume 8: Biannual report, August 1989--January 1990

    SciTech Connect (OSTI)

    Interrante, C.G. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of High-Level Waste Management; Fraker, A.C.; Escalante, E. [National Inst. of Standards and Technology (MSEL), Gaithersburg, MD (United States). Metallurgy Div.

    1993-06-01T23:59:59.000Z

    This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of some of the Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six-month period, August 1989--January 1990. This includes reviews of related materials research and plans, information on the Yucca Mountain, Nevada disposal site activities, and other information regarding supporting research and special assistance. Short discussions are given relating to the publications reviewed and complete reviews and evaluations are included. Reports of other work are included in the Appendices.

  4. 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.

    2013-11-13T23:59:59.000Z

    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.

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

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

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  6. Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility

    Office of Environmental Management (EM)

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

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

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  8. Geologyy of the Yucca Mountain Site Area, Southwestern Nevada, Chapter in Stuckless, J.S., ED., Yucca Mountain, Nevada - A Proposed Geologic Repository for High-Level Radioactive Waste (Volume 1)

    SciTech Connect (OSTI)

    W.R. Keefer; J.W. Whitney; D.C. Buesch

    2006-09-25T23:59:59.000Z

    Yucca Mountain in southwestern Nevada is a prominent, irregularly shaped upland formed by a thick apron of Miocene pyroclastic-flow and fallout tephra deposits, with minor lava flows, that was segmented by through-going, large-displacement normal faults into a series of north-trending, eastwardly tilted structural blocks. The principal volcanic-rock units are the Tiva Canyon and Topopah Spring Tuffs of the Paintbrush Group, which consist of volumetrically large eruptive sequences derived from compositionally distinct magma bodies in the nearby southwestern Nevada volcanic field, and are classic examples of a magmatic zonation characterized by an upper crystal-rich (> 10% crystal fragments) member, a more voluminous lower crystal-poor (< 5% crystal fragments) member, and an intervening thin transition zone. Rocks within the crystal-poor member of the Topopah Spring Tuff, lying some 280 m below the crest of Yucca Mountain, constitute the proposed host rock to be excavated for the storage of high-level radioactive wastes. Separation of the tuffaceous rock formations into subunits that allow for detailed mapping and structural interpretations is based on macroscopic features, most importantly the relative abundance of lithophysae and the degree of welding. The latter feature, varying from nonwelded through partly and moderately welded to densely welded, exerts a strong control on matrix porosities and other rock properties that provide essential criteria for distinguishing hydrogeologic and thermal-mechanical units, which are of major interest in evaluating the suitability of Yucca Mountain to host a safe and permanent geologic repository for waste storage. A thick and varied sequence of surficial deposits mantle large parts of the Yucca Mountain site area. Mapping of these deposits and associated soils in exposures and in the walls of trenches excavated across buried faults provides evidence for multiple surface-rupturing events along all of the major faults during Pleistocene and Holocene times; these paleoseismic studies form the basis for evaluating the potential for future earthquakes and fault displacements. Thermoluminescence and U-series analyses were used to date the surficial materials involved in the Quaternary faulting events. The rate of erosional downcutting of bedrock on the ridge crests and hillslopes of Yucca Mountain, being of particular concern with respect to the potential for breaching of the proposed underground storage facility, was studied by using rock varnish cation-ratio and {sup 10}Be and {sup 36}Cl cosmogenic dating methods to determine the length of time bedrock outcrops and hillslope boulder deposits were exposed to cosmic rays, which then served as a basis for calculating long-term erosion rates. The results indicate rates ranging from 0.04 to 0.27 cm/k.y., which represent the maximum downcutting along the summit of Yucca Mountain under all climatic conditions that existed there during most of Quaternary time. Associated studies include the stratigraphy of surficial deposits in Fortymile Wash, the major drainage course in the area, which record a complex history of four to five cut-and-fill cycles within the channel during middle to late Quaternary time. The last 2 to 4 m of incision probably occurred during the last pluvial climatic period, 22 to 18 ka, followed by aggradation to the present time.

  9. 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.

    2013-11-13T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Howden, G.F.

    1994-10-24T23:59:59.000Z

    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. Radioactive Waste Management (Minnesota)

    Broader source: Energy.gov [DOE]

    This section regulates the transportation and disposal of high-level radioactive waste in Minnesota, and establishes a Nuclear Waste Council to monitor the federal high-level radioactive waste...

  12. 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.

    2011-11-29T23:59:59.000Z

    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.

  13. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM 2009

    SciTech Connect (OSTI)

    West, B.; Waltz, R.

    2010-06-21T23:59:59.000Z

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

  14. 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.

    2011-09-20T23:59:59.000Z

    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

  15. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM 2010

    SciTech Connect (OSTI)

    West, B.; Waltz, R.

    2011-06-23T23:59:59.000Z

    Aqueous radioactive wastes from Savannah River Site (SRS) separations and vitrification processes are contained in large underground carbon steel tanks. Inspections made during 2010 to evaluate these vessels and other waste handling facilities along with evaluations based on data from previous inspections are the subject of this report. The 2010 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-2009-00138, HLW Tank Farm Inspection Plan for 2010, were completed. Ultrasonic measurements (UT) performed in 2010 met the requirements of C-ESG-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 30, 31 and 32 and the findings are documented in SRNL-STI-2010-00533, Tank Inspection NDE Results for Fiscal Year 2010, Waste Tanks 30, 31 and 32. A total of 5824 photographs were made and 1087 visual and video inspections were performed during 2010. Ten new leaksites at Tank 5 were identified in 2010. The locations of these leaksites are documented in C-ESR-G-00003, SRS High Level Waste Tank Leaksite Information, Rev.5. Ten leaksites at Tank 5 were documented during tank wall/annulus cleaning activities. None of these new leaksites resulted in a release to the environment. The leaksites were documented during wall cleaning activities and the waste nodules associated with the leaksites were washed away. Previously documented leaksites were reactivated at Tank 12 during waste removal activities.

  16. 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

    2011-12-29T23:59:59.000Z

    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.

  17. Proceedings of the US Department of Energy Office of Environmental Restoration and Waste Management

    SciTech Connect (OSTI)

    Not Available

    1990-09-01T23:59:59.000Z

    The fifth of a series of waste minimization (WMIN)/reduction workshops (Waste Reduction Workshop V) was held at the Little Tree Inn in Idaho Falls, Idaho, on July 24--26, 1990. The workshops are held under the auspices of the US Department of Energy's (DOE's) Office of Environmental Restoration and Waste Management (EM). The purpose of this workshop was to provide a forum for sharing site activities in WMIN/reduction planning. Topics covered were management commitment, organizational structure, goal setting, reporting requirements, data bases and tracking systems, pollution prevention, awareness and incentives, information exchange, process waste assessment (PWA) implementation, and recycling internal and external. The workshops assist DOE waste-generating sites in implementing WMIN/reduction programs, plans, and activities, thus providing for optimal waste reduction within the DOE complex. All wastes are considered within this discipline: liquid, solid, and airborne, within the categories of high-level waste (HLW), transuranic waste (TRU), low-level waste (LLW), hazardous waste, and mixed waste.

  18. Modeling Offgas Systems for the Hanford Waste Treatment Plant

    SciTech Connect (OSTI)

    Smith, Frank G., III

    2005-09-02T23:59:59.000Z

    To augment steady-state design calculations, dynamic models of three offgas systems that will be used in the Waste Treatment Plant now under construction at the Hanford Site were developed using Aspen Custom Modeler{trademark}. The offgas systems modeled were those for the High Level Waste (HLW) melters, Low Activity Waste (LAW) melters and HLW Pulse Jet Ventilation (PJV) system. The models do not include offgas chemistry but only consider the two major species in the offgas stream which are air and water vapor. This is sufficient to perform material and energy balance calculations that accurately show the dynamic behavior of gas pressure, temperature, humidity and flow throughout the systems. The models are structured to perform pressure drop calculations across the various unit operations using a combination of standard engineering calculations and empirical data based correlations for specific pieces of equipment. The models include process controllers, gas ducting, control valves, exhaust fans and the offgas treatment equipment. The models were successfully used to analyze a large number of operating scenarios including both normal and off-normal conditions.

  19. Organic tank safety project: Preliminary results of energetics and thermal behavior studies of model organic nitrate and/or nitrite mixtures and a simulated organic waste

    SciTech Connect (OSTI)

    Scheele, R.D.; Sell, R.L.; Sobolik, J.L.; Burger, L.L.

    1995-08-01T23:59:59.000Z

    As a result of years of production and recovery of nuclear defense materials and subsequent waste management at the Hanford Site, organic-bearing radioactive high-level wastes (HLW) are currently stored in large (up to 3. ML) single-shell storage tanks (SSTs). Because these wastes contain both fuels (organics) and the oxidants nitrate and nitrite, rapid energetic reactions at certain conditions could occur. In support of Westinghouse Hanford Company`s (WHC) efforts to ensure continued safe storage of these organic- and oxidant-bearing wastes and to define the conditions necessary for reactions to occur, we measured the thermal sensitivities and thermochemical and thermokinetic properties of mixtures of selected organics and sodium nitrate and/or nitrite and a simulated Hanford organic-bearing waste using thermoanalytical technologies. These thermoanalytical technologies are used by chemical reactivity hazards evaluation organizations within the chemical industry to assess chemical reaction hazards.

  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

    2011-01-13T23:59:59.000Z

    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. 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.

    1996-12-01T23:59:59.000Z

    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.

  2. Retention of Halogens in Waste Glass

    SciTech Connect (OSTI)

    Hrma, Pavel R.

    2010-05-01T23:59:59.000Z

    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.

  3. 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.

    2013-11-13T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Stefanko, D.; Herbert, J.

    2012-01-10T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Kruger, Albert A.

    2013-01-16T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Kruger, Albert A. [WTP Engineering Division, United States Department of Energy, Office of River Protection, Post Office Box 450, Richland, Washington 99352 (United States)] [WTP Engineering Division, United States Department of Energy, Office of River Protection, Post Office Box 450, Richland, Washington 99352 (United States)

    2013-07-01T23:59:59.000Z

    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. (author)

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

    SciTech Connect (OSTI)

    Jantzen, Carol M. [Savannah River National Lab., Aiken SC (United States); Lee, William E. [Imperial College, London (United Kingdom). Dept. of Materials; Ojovan, Michael I. [Univ. of Sheffield (United Kingdom). Dept. of Materials Science and Engineering

    2012-10-19T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    James Harvey; Michael Gula

    1998-12-07T23:59:59.000Z

    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.

  9. Waste Package Component Design Methodology Report

    SciTech Connect (OSTI)

    D.C. Mecham

    2004-07-12T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1993-10-12T23:59:59.000Z

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

  11. UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 Arlington is intended to update Congress and the Secretary of Energy on the U.S. Nuclear Waste Technical Review Board-level radioactive waste (HLW) is evolving. The letter is issued in accordance with provisions of the Nuclear Waste

  12. 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.

    2013-11-13T23:59:59.000Z

    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.

  13. Summary of national and international fuel cycle and radioactive waste management programs, 1984

    SciTech Connect (OSTI)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

    1984-07-01T23:59:59.000Z

    Worldwide activities related to nuclear fuel cycle and radioactive waste management programs are summarized. Several trends have developed in waste management strategy: All countries having to dispose of reprocessing wastes plan on conversion of the high-level waste (HLW) stream to a borosilicate glass and eventual emplacement of the glass logs, suitably packaged, in a deep geologic repository. Countries that must deal with plutonium-contaminated waste emphasize pluonium recovery, volume reduction and fixation in cement or bitumen in their treatment plans and expect to use deep geologic repositories for final disposal. Commercially available, classical engineering processing are being used worldwide to treat and immobilize low- and intermediate-level wastes (LLW, ILW); disposal to surface structures, shallow-land burial and deep-underground repositories, such as played-out mines, is being done widely with no obvious technical problems. Many countries have established extensive programs to prepare for construction and operation of geologic repositories. Geologic media being studied fall into three main classes: argillites (clay or shale); crystalline rock (granite, basalt, gneiss or gabbro); and evaporates (salt formations). Most nations plan to allow 30 years or longer between discharge of fuel from the reactor and emplacement of HLW or spent fuel is a repository to permit thermal and radioactive decay. Most repository designs are based on the mined-gallery concept, placing waste or spent fuel packages into shallow holes in the floor of the gallery. Many countries have established extensive and costly programs of site evaluation, repository development and safety assessment. Two other waste management problems are the subject of major R and D programs in several countries: stabilization of uranium mill tailing piles; and immobilization or disposal of contaminated nuclear facilities, namely reactors, fuel cycle plants and R and D laboratories.

  14. Tank waste remediation system operation and utilization plan,vol. I {ampersand} II

    SciTech Connect (OSTI)

    Kirkbride, R.A.

    1997-09-01T23:59:59.000Z

    The U.S. Department of Energy Richland Operations Office (RL) is in the first stages of contracting with private companies for the treatment and immobilization of tank wastes. The components of tank waste retrieval, treatment, and immobilization have been conceived in two phases (Figure 1.0-1). To meet RL's anticipated contractual requirements, the Project Hanford Management Contractor (PHMC) companies will be required to provide waste feeds to the private companies consistent with waste envelopes that define the feeds in terms of quantity, and concentration of both chemicals and radionuclides. The planning that supports delivery of the feed must be well thought out in four basic areas: (1) Low-activity waste (LAW)/high-level waste (HLW) feed staging plans. How is waste moved within the existing tanks to deliver waste that corresponds to the defined feed envelopes to support the Private Contractor's processing schedule and processing rate? (2) Single-shell tank (SST) retrieval sequence. How are Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) (Ecology et al. 1994) milestones for SST retrieval integrated into the Phase I processing to set the stage for Phase II processing to complete the mission? (3) Tank Waste Remediation System (TWRS) process flowsheet. How do materials flow from existing tank inventories through: (1) blending and pretreatment functions in the double-shell tanks (DSTs), (2) contractor processing facilities, and (3) stored waste forms (Figure 1.0-2); (4) Storage and disposal of the immobilized low-activity waste (ILAW) and immobilized high-level waste (IHLW) product. How is the ILAW and IHLW product received from the private companies, the ILAW disposed onsite, and the IHLW stored onsite until final disposal?

  15. SRNL PHASE 1 ASSESSMENT OF THE WTP WASTE QUALIFICATION PROGRAM

    SciTech Connect (OSTI)

    Peeler, D.; Hansen, E.; Herman, C.; Marra, S.; Wilmarth, B.

    2012-03-06T23:59:59.000Z

    The Hanford Tank Waste Treatment and Immobilization Plant (WTP) Project is currently transitioning its emphasis from an engineering design and construction phase toward facility completion, start-up and commissioning. With this transition, the WTP Project has initiated more detailed assessments of the requirements that must be met during the actual processing of the Hanford Site tank waste. One particular area of interest is the waste qualification program. In general, the waste qualification program involves testing and analysis to demonstrate compliance with waste acceptance criteria, determine waste processability, and demonstrate laboratory-scale unit operations to support WTP operations. The testing and analysis are driven by data quality objectives (DQO) requirements necessary for meeting waste acceptance criteria for transfer of high-level wastes from the tank farms to the WTP, and for ensuring waste processability including proper glass formulations during processing within the WTP complex. Given the successful implementation of similar waste qualification efforts at the Savannah River Site (SRS) which were based on critical technical support and guidance from the Savannah River National Laboratory (SRNL), WTP requested subject matter experts (SMEs) from SRNL to support a technology exchange with respect to waste qualification programs in which a critical review of the WTP program could be initiated and lessons learned could be shared. The technology exchange was held on July 18-20, 2011 in Richland, Washington, and was the initial step in a multi-phased approach to support development and implementation of a successful waste qualification program at the WTP. The 3-day workshop was hosted by WTP with representatives from the Tank Operations Contractor (TOC) and SRNL in attendance as well as representatives from the US DOE Office of River Protection (ORP) and the Defense Nuclear Facility Safety Board (DNFSB) Site Representative office. The purpose of the workshop was to share lessons learned and provide a technology exchange to support development of a technically defensible waste qualification program. The objective of this report is to provide a review, from SRNL's perspective, of the WTP waste qualification program as presented during the workshop. In addition to SRNL's perspective on the general approach to the waste qualification program, more detailed insight into the specific unit operations presented by WTP during the workshop is provided. This report also provides a general overview of the SRS qualification program which serves as a basis for a comparison between the two programs. Recommendations regarding specific steps are made based on the review and SRNL's lessons learned from qualification of SRS low-activity waste (LAW) and high-level waste (HLW) to support maturation of the waste qualification program leading to WTP implementation.

  16. TECHNICAL PEER REVIEW REPORT - YUCCA MOUNTAIN: WASTE PACKAGE CLOSURE CONTROL SYSTEM

    SciTech Connect (OSTI)

    NA

    2005-10-25T23:59:59.000Z

    The objective of the Waste Package Closure System (WPCS) project is to assist in the disposal of spent nuclear fuel (SNF) and associated high-level wastes (HLW) at the Yucca Mountain site in Nevada. Materials will be transferred from the casks into a waste package (WP), sealed, and placed into the underground facility. The SNF/HLW transfer and closure operations will be performed in an aboveground facility. The objective of the Control System is to bring together major components of the entire WPCS ensuring that unit operations correctly receive, and respond to, commands and requests for data. Integrated control systems will be provided to ensure that all operations can be performed remotely. Maintenance on equipment may be done using hands-on or remote methods, depending on complexity, exposure, and ease of access. Operating parameters and nondestructive examination results will be collected and stored as permanent electronic records. Minor weld repairs must be performed within the closure cell if the welds do not meet the inspection acceptance requirements. Any WP with extensive weld defects that require lids to be removed will be moved to the remediation facility for repair.

  17. 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

    2011-12-29T23:59:59.000Z

    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.

  18. Comparison of simulants to actual neutralized current acid waste: process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

    SciTech Connect (OSTI)

    Morrey, E.V.; Tingey, J.M.; Elliott, M.L.

    1996-10-01T23:59:59.000Z

    A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs were established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste was performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property ,models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions.

  19. Comparison of simulants to actual neutralized current acid waste: Process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

    SciTech Connect (OSTI)

    Morrey, E.V.; Tingey, J.M.

    1996-04-01T23:59:59.000Z

    A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs have been established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste is being performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions.

  20. Geological problems in radioactive waste isolation

    SciTech Connect (OSTI)

    Witherspoon, P.A. (ed.)

    1991-01-01T23:59:59.000Z

    The problem of isolating radioactive wastes from the biosphere presents specialists in the fields of earth sciences with some of the most complicated problems they have ever encountered. This is especially true for high level waste (HLW) which must be isolated in the underground and away from the biosphere for thousands of years. Essentially every country that is generating electricity in nuclear power plants is faced with the problem of isolating the radioactive wastes that are produced. The general consensus is that this can be accomplished by selecting an appropriate geologic setting and carefully designing the rock repository. Much new technology is being developed to solve the problems that have been raised and there is a continuing need to publish the results of new developments for the benefit of all concerned. The 28th International Geologic Congress that was held July 9--19, 1989 in Washington, DC provided an opportunity for earth scientists to gather for detailed discussions on these problems. Workshop W3B on the subject, Geological Problems in Radioactive Waste Isolation -- A World Wide Review'' was organized by Paul A Witherspoon and Ghislain de Marsily and convened July 15--16, 1989 Reports from 19 countries have been gathered for this publication. Individual papers have been cataloged separately.

  1. Process Options Description for Vitrification Flowsheet Model of INEEL Sodium Bearing Waste

    SciTech Connect (OSTI)

    Nichols, Todd Travis; Taylor, Dean Dalton; Lauerhass, Lance; Barnes, Charles Marshall

    2001-02-01T23:59:59.000Z

    The purpose of this document is to provide the technical information to Savannah River Site (SRS) personnel that is required for the development of a basic steady-state process simulation of the vitrification treatment train of sodium bearing waste (SBW) at Idaho National Engineering and nvironmental Laboratory (INEEL). INEEL considers simulation to have an important role in the integration/optimization of treatment process trains for the High Level Waste (HLW) Program. This project involves a joint Technical Task Plan (TTP ID77WT31, Subtask C) between SRS and INEEL. The work scope of simulation is different at the two sites. This document addresses only the treatment of SBW at INEEL. The simulation model(s) is to be built by SRS for INEEL in FY-2001.

  2. Incorporating Cold Cap Behavior in a Joule-heated Waste Glass Melter Model

    SciTech Connect (OSTI)

    Varija Agarwal; Donna Post Guillen

    2013-08-01T23:59:59.000Z

    In this paper, an overview of Joule-heated waste glass melters used in the vitrification of high level waste (HLW) is presented, with a focus on the cold cap region. This region, in which feed-to-glass conversion reactions occur, is critical in determining the melting properties of any given glass melter. An existing 1D computer model of the cold cap, implemented in MATLAB, is described in detail. This model is a standalone model that calculates cold cap properties based on boundary conditions at the top and bottom of the cold cap. Efforts to couple this cold cap model with a 3D STAR-CCM+ model of a Joule-heated melter are then described. The coupling is being implemented in ModelCenter, a software integration tool. The ultimate goal of this model is to guide the specification of melter parameters that optimize glass quality and production rate.

  3. 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

    2011-12-29T23:59:59.000Z

    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.

  4. CSER 94-004: Criticality safety of double-shell waste storage tanks

    SciTech Connect (OSTI)

    Rogers, C.A.

    1994-09-22T23:59:59.000Z

    This criticality safety evaluation covers double-shell waste storage tanks (DSTs), double-contained receiver tanks (DCRTs), vault tanks, and the 242-A Evaporator located in the High Level Waste (HLW) Tank Farms on the Hanford Site. Limits and controls are specified and the basis for ensuring criticality safety is discussed. A minimum limit of 1,000 is placed upon the solids/plutonium mass ratio in incoming waste. The average solids/Pu mass ratio over all waste in tank farms is estimated to be about 74,500, about 150 times larger than required to assure subcriticality in homogeneous waste. PFP waste in Tank-102-SY has an estimated solids/Pu mass ratio of 10,000. Subcriticality is assured whenever the plutonium concentration is less than 2.6 g. The median reported plutonium concentration for 200 samples of waste solids is about 0.01 g (0.038 g/gal). A surveillance program is proposed to increase the knowledge of the waste and provide added assurance of the high degree of subcriticality.

  5. State-of-the-art review of materials properties of nuclear waste forms.

    SciTech Connect (OSTI)

    Mendel, J. E.; Nelson, R. D.; Turcotte, R. P.; Gray, W. J.; Merz, M. D.; Roberts, F. P.; Weber, W. J.; Westsik, Jr., J. H.; Clark, D. E.

    1981-04-01T23:59:59.000Z

    The Materials Characterization Center (MCC) was established at the Pacific Northwest Laboratory to assemble a standardized nuclear waste materials data base for use in research, systems and facility design, safety analyses, and waste management decisions. This centralized data base will be provided through the means of a Nuclear Waste Materials Handbook. The first issue of the Handbook will be published in the fall of 1981 in looseleaf format so that it can be updated as additional information becomes available. To ensure utmost reliability, all materials data appearing in the Handbook will be obtained by standard procedures defined in the Handbook and approved by an independent Materials Review Board (MRB) comprised of materials experts from Department of Energy laboratories and from universities and industry. In the interim before publication of the Handbook there is need for a report summarizing the existing materials data on nuclear waste forms. This review summarizes materials property data for the nuclear waste forms that are being developed for immobilization of high-level radioactive waste. It is intended to be a good representation of the knowledge concerning the properties of HLW forms as of March 1981. The table of contents lists the following topics: introduction which covers waste-form categories, and important waste-form materials properties; physical properties; mechanical properties; chemical durability; vaporization; radiation effects; and thermal phase stability.

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

    SciTech Connect (OSTI)

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

    2011-09-27T23:59:59.000Z

    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. Conclusions from this work include the following. The CSSX process is capable of reducing {sup 137}Cs in high level radioactive waste by a factor of more than 40,000 using five extraction, two scrub, and five strip stages. Tests demonstrated extraction and strip section stage efficiencies of greater than 93% for the Tank 49H waste test and greater than 88% for the simulant waste test. During a test with HLW, researchers processed 39 liters of Tank 49H solution and the waste raffinate had an average decontamination factor (DF) of 6.78E+04, with a maximum of 1.08E+05. A simulant waste solution ({approx}34.5 liters) with an initial Cs concentration of 83.1 mg/L was processed and had an average DF greater than 5.9E+03, with a maximum DF of greater than 6.6E+03. The difference may be attributable to differences in contactor stage efficiencies. Test results showed the solvent can be stripped of cesium and recycled for {approx}25 solvent turnovers without the occurrence of any measurable solvent degradation or negative effects from minor components. Based on the performance of the 12-stage 2-cm apparatus with the Tank 49H HLW, the projected DF for MCU with seven extraction, two scrub, and seven strip stages operating at a nominal efficiency of 90% is {approx}388,000. At 95% stage efficiency, the DF in MCU would be {approx}3.2 million. Carryover of organic solvent in aqueous streams (and aqueous in organic streams) was less than 0.1% when processing Tank 49H HLW. The entrained solvent concentration measured in the decontaminated salt solution (DSS) was as much as {approx}140 mg/L, although that value may be overstated by as much as 50% due to modifier solubility in the DSS. The entrained solvent concentration was measured in the strip effluent (SE) and the results are pending. A steady-state concentration factor (CF) of 15.9 was achieved with Tank 49H HLW. Cesium distribution ratios [D(Cs)] were measured with non-radioactive Tank 49H waste simulant and actual Tank 49H waste. Below is a comparison of D(Cs) values of ESS and 2-cm tests. Batch Extraction-Strip-Scrub (ESS) tests yielded D(Cs) values for extraction of {approx}81-88 for tests with Tank 49H waste and waste simulant. The results from the 2-cm contactor tests were in agreement with values of 58-92 for the Tank 49H HLW test and 54-83 for the simulant waste test. These values are consistent with the reference D(Cs) for extraction of {approx}60. In tests with Tank 49H waste and waste simulant, batch ESS tests measured D(Cs) values for the two scrub stages as {approx}3.5-5.0 for the first scrub stage and {approx}1.0-3.0 for the second scrub stage. In the Tank 49H test, the D(Cs) values for the 2-cm test were far from the ESS values. A D(Cs) value of 161 was measured for the first scrub stage and 10.8 for the second scrub stage. The data suggest that the scrub stage is not operating as effectively as intended. For the simulant test, a D(Cs) value of 1.9 was measured for the first scrub stage; the sample from the second scrub stage was compromised. Measurements of the pH of all stage samples for the Tank 49H test showed that the pH for extraction and scrub stages was 14 and the pH for the strip stages was {approx}7. It is expected that the pH of the second scrub stage would be {approx}12-13. Batch ESS tests measured D(Cs) values for the strip stages to be {approx}0.002-0.010. A high value in Strip No.3 of a test with simulant solution has been attributed to issues associated with the limits of detection for the

  7. Spent Fuel and High-Level Waste Requirements (Maine)

    Broader source: Energy.gov [DOE]

    All proposed nuclear power generation facilities must be certified by the Public Utilities Commission under this statute prior to construction and operation. The facility may be certified if the...

  8. Hanford Site River Protection Project (RPP) High Level Waste Storage

    SciTech Connect (OSTI)

    KRISTOFZSKI, J.G.

    2000-01-31T23:59:59.000Z

    The CH2M HILL Hanford Group (CHG) conducts business to achieve the goals of the U.S. Department of Energy's (DOE) Office of River Protection at the Hanford Site. The CHG is organized to manage and perform work to safely store, retrieve, etc.

  9. 2008 DOE Spent Nuclear Fuel and High Level Waste Inventory

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

    Working with INL Community Outreach Visitor Information Calendar of Events ATR National Scientific User Facility Center for Advanced Energy Studies Light Water Reactor...

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-SeriesFlickr FlickrGuidedCH2M HILLAdministration | National|GradientHigh. H L

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists'Montana.ProgramJulietip sheetK-4In 2013 many| Department of4 Energy

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists'Montana.ProgramJulietip sheetK-4In 2013 many| Department HIGHImage ofEnergy

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG | Department ofHTS Cable Projects HTSSeparationHelping to Finance3 June 2008

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG | Department ofHTS Cable Projects HTSSeparationHelping to Finance3 June

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG | Department ofHTS Cable Projects HTSSeparationHelping to Finance3

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartmentEnergy DataRemediatedLandsEnergyDepartmentReport on

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG |September 15, 2010 PrintingNeed forNews »DepartmentToolsNorth

  18. 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

    2011-12-29T23:59:59.000Z

    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 impr

  19. Baseline Flowsheet Generation for the Treatment and Disposal of Idaho National Engineering and Environmental Laboratory Sodium Bearing Waste

    SciTech Connect (OSTI)

    Barnes, C.M.; Lauerhass, L.; Olson, A.L.; Taylor, D.D.; Valentine, J.H.; Lockie, K.A. (DOE- ID)

    2002-01-16T23:59:59.000Z

    The High-Level Waste (HLW) Program at the Idaho National Engineering and Environmental Laboratory (INEEL) must implement technologies and processes to treat and qualify radioactive wastes located at the Idaho Nuclear Technology and Engineering Center (INTEC) for permanent disposal. This paper describes the approach and accomplishments to date for completing development of a baseline vitrification treatment flowsheet for sodium-bearing waste (SBW), including development of a relational database used to manage the associated process assumptions. A process baseline has been developed that includes process requirements, basis and assumptions, process flow diagrams, a process description, and a mass balance. In the absence of actual process or experimental results, mass and energy balance data for certain process steps are based on assumptions. Identification, documentation, validation, and overall management of the flowsheet assumptions are critical to ensuring an integrated, focused program. The INEEL HLW Program initially used a roadmapping methodology, developed through the INEEL Environmental Management Integration Program, to identify, document, and assess the uncertainty and risk associated with the SBW flowsheet process assumptions. However, the mass balance assumptions, process configuration and requirements should be accessible to all program participants. This need resulted in the creation of a relational database that provides formal documentation and tracking of the programmatic uncertainties related to the SBW flowsheet.

  20. Baseline Flowsheet Generation for the Treatment and Disposal of Idaho National Engineering and Environmental Laboratory Sodium Bearing Waste

    SciTech Connect (OSTI)

    Barnes, Charles Marshall; Lauerhass, Lance; Olson, Arlin Leland; Taylor, Dean Dalton; Valentine, James Henry; Lockie, Keith Andrew

    2002-02-01T23:59:59.000Z

    The High-Level Waste (HLW) Program at the Idaho National Engineering and Environmental Laboratory (INEEL) must implement technologies and processes to treat and qualify radioactive wastes located at the Idaho Nuclear Technology and Engineering Center (INTEC) for permanent disposal. This paper describes the approach and accomplishments to date for completing development of a baseline vitrification treatment flowsheet for sodium-bearing waste (SBW), including development of a relational database used to manage the associated process assumptions. A process baseline has been developed that includes process requirements, basis and assumptions, process flow diagrams, a process description, and a mass balance. In the absence of actual process or experimental results, mass and energy balance data for certain process steps are based on assumptions. Identification, documentation, validation, and overall management of the flowsheet assumptions are critical to ensuring an integrated, focused program. The INEEL HLW Program initially used a roadmapping methodology, developed through the INEEL Environmental Management Integration Program, to identify, document, and assess the uncertainty and risk associated with the SBW flowsheet process assumptions. However, the mass balance assumptions, process configuration and requirements should be accessible to all program participants. This need resulted in the creation of a relational database that provides formal documentation and tracking of the programmatic uncertainties related to the SBW flowsheet.

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

    SciTech Connect (OSTI)

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

    1980-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Reimann, G.A.

    1991-10-01T23:59:59.000Z

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

  3. Service-oriented high level architecture

    E-Print Network [OSTI]

    Wang, Wenguang; Li, Qun; Wang, Weiping; Liu, Xichun

    2009-01-01T23:59:59.000Z

    Service-oriented High Level Architecture (SOHLA) refers to the high level architecture (HLA) enabled by Service-Oriented Architecture (SOA) and Web Services etc. techniques which supports distributed interoperating services. The detailed comparisons between HLA and SOA are made to illustrate the importance of their combination. Then several key enhancements and changes of HLA Evolved Web Service API are introduced in comparison with native APIs, such as Federation Development and Execution Process, communication mechanisms, data encoding, session handling, testing environment and performance analysis. Some approaches are summarized including Web-Enabling HLA at the communication layer, HLA interface specification layer, federate interface layer and application layer. Finally the problems of current research are discussed, and the future directions are pointed out.

  4. INSTALLATION OF BUBBLERS IN THE SAVANNAH RIVER SITED DEFENSE WASTE PROCESSING FACILITY MELTER

    SciTech Connect (OSTI)

    Smith, M.; Iverson, D.

    2010-12-08T23:59:59.000Z

    Savannah River Remediation (SRR) LLC assumed the liquid waste contract at the Savannah River Site (SRS) in the summer of 2009. The main contractual agreement was to close 22 High Level Waste (HLW) tanks in eight years. To achieve this aggressive commitment, faster waste processing throughout the SRS liquid waste facilities will be required. Part of the approach to achieve faster waste processing is to increase the canister production rate of the Defense Waste Processing Facility (DWPF) from approximately 200 canisters filled with radioactive waste glass per year to 400 canisters per year. To reach this rate for melter throughput, four bubblers were installed in the DWPF Melter in the late summer of 2010. This effort required collaboration between SRR, SRR critical subcontractor EnergySolutions, and Savannah River Nuclear Solutions, including the Savannah River National Laboratory (SRNL). The tasks included design and fabrication of the bubblers and related equipment, testing of the bubblers for various technical issues, the actual installation of the bubblers and related equipment, and the initial successful operation of the bubblers in the DWPF Melter.

  5. Service Oriented Architecture for High Level Applications

    SciTech Connect (OSTI)

    Chu, Chungming; Chevtsov, Sergei; Wu, Juhao; /SLAC; Shen, Guobao; /Brookhaven

    2012-06-28T23:59:59.000Z

    Standalone high level applications often suffer from poor performance and reliability due to lengthy initialization, heavy computation and rapid graphical update. Service-oriented architecture (SOA) is trying to separate the initialization and computation from applications and to distribute such work to various service providers. Heavy computation such as beam tracking will be done periodically on a dedicated server and data will be available to client applications at all time. Industrial standard service architecture can help to improve the performance, reliability and maintainability of the service. Robustness will also be improved by reducing the complexity of individual client applications.

  6. High Level Computational Chemistry Approaches to the Prediction...

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

    High Level Computational Chemistry Approaches to the Prediction of Energetic Properties of Chemical Hydrogen Storage Systems High Level Computational Chemistry Approaches to the...

  7. Application of Synergistic Technologies to Achieve High Levels...

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

    Synergistic Technologies to Achieve High Levels of Gasoline Engine Downsizing Application of Synergistic Technologies to Achieve High Levels of Gasoline Engine Downsizing Discussed...

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

    SciTech Connect (OSTI)

    Cherkas, Dmytro

    2011-10-01T23:59:59.000Z

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

  9. Geological problems in radioactive waste isolation - second worldwide review

    SciTech Connect (OSTI)

    Witherspoon, P.A. [ed.

    1996-09-01T23:59:59.000Z

    The first world wide review of the geological problems in radioactive waste isolation was published by Lawrence Berkeley National Laboratory in 1991. This review was a compilation of reports that had been submitted to a workshop held in conjunction with the 28th International Geological Congress that took place July 9-19, 1989 in Washington, D.C. Reports from 15 countries were presented at the workshop and four countries provided reports after the workshop, so that material from 19 different countries was included in the first review. It was apparent from the widespread interest in this first review that the problem of providing a permanent and reliable method of isolating radioactive waste from the biosphere is a topic of great concern among the more advanced, as well as the developing, nations of the world. This is especially the case in connection with high-level waste (HLW) after its removal from nuclear power plants. The general concensus is that an adequate isolation can be accomplished by selecting an appropriate geologic setting and carefully designing the underground system with its engineered barriers. This document contains the Second Worldwide Review of Geological Problems in Radioactive Waste Isolation, dated September 1996.

  10. Generic Degraded Congiguration Probability Analysis for DOE Codisposal Waste Package

    SciTech Connect (OSTI)

    S.F.A. Deng; M. Saglam; L.J. Gratton

    2001-05-23T23:59:59.000Z

    In accordance with the technical work plan, ''Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages'' (CRWMS M&O 2000c), this Analysis/Model Report (AMR) is developed for the purpose of screening out degraded configurations for U.S. Department of Energy (DOE) spent nuclear fuel (SNF) types. It performs the degraded configuration parameter and probability evaluations of the overall methodology specified in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000, Section 3) to qualifying configurations. Degradation analyses are performed to assess realizable parameter ranges and physical regimes for configurations. Probability calculations are then performed for configurations characterized by k{sub eff} in excess of the Critical Limit (CL). The scope of this document is to develop a generic set of screening criteria or models to screen out degraded configurations having potential for exceeding a criticality limit. The developed screening criteria include arguments based on physical/chemical processes and probability calculations and apply to DOE SNF types when codisposed with the high-level waste (HLW) glass inside a waste package. The degradation takes place inside the waste package and is long after repository licensing has expired. The emphasis of this AMR is on degraded configuration screening and the probability analysis is one of the approaches used for screening. The intended use of the model is to apply the developed screening criteria to each DOE SNF type following the completion of the degraded mode criticality analysis internal to the waste package.

  11. Proceedings of the 1993 international conference on nuclear waste management and environmental remediation. Volume 1: Low and intermediate level radioactive waste management

    SciTech Connect (OSTI)

    Alexandre, D.; Baker, R.; Kohout, R.; Marke, J. [eds.

    1993-12-31T23:59:59.000Z

    This conference was held in 1993 in Prague, Czech Republic to provide a forum for exchange of state-of-the-art information on radioactive waste management. Sections of the proceedings include the following: engineered and geological barriers; radioactive transportation and related issues; national programs for HLW and spent fuel management; recent advances in HLW treatment; science and engineering of the repository system; national examples of spent fuel storage systems; disposal systems for HLW and spent fuel--design and status; reprocessing and its influence on waste quantities; qualification methods for HLW; site characterization and underground testing; science and engineering of HLW and spent fuel storage systems; transmutation and separations of radioactivity; disposal site selection approaches and issues; and safety assessments and methodology for repositories.

  12. Umbra's High Level Architecture (HLA) Interface

    SciTech Connect (OSTI)

    GOTTLIEB, ERIC JOSEPH; MCDONALD, MICHAEL J.; OPPEL III, FRED J.

    2002-04-01T23:59:59.000Z

    This report describes Umbra's High Level Architecture HLA library. This library serves as an interface to the Defense Simulation and Modeling Office's (DMSO) Run Time Infrastructure Next Generation Version 1.3 (RTI NG1.3) software library and enables Umbra-based models to be federated into HLA environments. The Umbra library was built to enable the modeling of robots for military and security system concept evaluation. A first application provides component technologies that ideally fit the US Army JPSD's Joint Virtual Battlespace (JVB) simulation framework for Objective Force concept analysis. In addition to describing the Umbra HLA library, the report describes general issues of integrating Umbra with RTI code and outlines ways of building models to support particular HLA simulation frameworks like the JVB.

  13. Characteristics of potential repository wastes. Volume 2

    SciTech Connect (OSTI)

    Not Available

    1992-07-01T23:59:59.000Z

    The LWR spent fuels discussed in Volume 1 of this report comprise about 99% of all domestic non-reprocessed spent fuel. In this report we discuss other types of spent fuels which, although small in relative quantity, consist of a number of diverse types, sizes, and compositions. Many of these fuels are candidates for repository disposal. Some non-LWR spent fuels are currently reprocessed or are scheduled for reprocessing in DOE facilities at the Savannah River Site, Hanford Site, and the Idaho National Engineering Laboratory. It appears likely that the reprocessing of fuels that have been reprocessed in the past will continue and that the resulting high-level wastes will become part of defense HLW. However, it is not entirely clear in some cases whether a given fuel will be reprocessed, especially in cases where pretreatment may be needed before reprocessing, or where the enrichment is not high enough to make reprocessing attractive. Some fuels may be canistered, while others may require special means of disposal. The major categories covered in this chapter include HTGR spent fuel from the Fort St. Vrain and Peach Bottom-1 reactors, research and test reactor fuels, and miscellaneous fuels, and wastes generated from the decommissioning of facilities.

  14. Melt Processed Single Phase Hollandite Waste Forms For Nuclear Waste Immobilization: Ba{sub 1.0}Cs{sub 0.3}A{sub 2.3}Ti{sub 5.7}O{sub 16}; A = Cr, Fe, Al

    SciTech Connect (OSTI)

    Brinkman, Kyle [Savannah River National Laboratory, Aiken, SC 29808 (United States); Marra, James [Savannah River National Laboratory, Aiken, SC 29808 (United States); Amoroso, Jake [Savannah River National Laboratory, Aiken, SC 29808 (United States); Conradson, Steven D. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Tang, Ming [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2013-09-23T23:59:59.000Z

    Cs is one of the more problematic fission product radionuclides to immobilize due to its high volatility at elevated temperatures, ability to form water soluble compounds, and its mobility in many host materials. The hollandite structure is a promising crystalline host for Cs immobilization and has been traditionally fabricated by solid state sintering methods. This study presents the structure and performance of Ba{sub 1.0}Cs{sub 0.3}A{sub 2.3}Ti{sub 5.7}O{sub 16}; A = Cr, Fe, Al hollandite fabricated by melt processing. Melt processing is considered advantageous given that melters are currently in use for High Level Waste (HLW) vitrification in several countries. This work details the impact of Cr additions that were demonstrated to i) promote the formation of a Cs containing hollandite phase and ii) maintain the stability of the hollandite phase in reducing conditions anticipated for multiphase waste form processing.

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

    SciTech Connect (OSTI)

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

    2012-01-12T23:59:59.000Z

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

  16. Glass Ceramic Waste Forms for Combined CS+LN+TM Fission Products Waste Streams

    SciTech Connect (OSTI)

    Crum, Jarrod V.; Turo, Laura A.; Riley, Brian J.; Tang, Ming; Kossoy, Anna; Sickafus, Kurt E.

    2010-09-23T23:59:59.000Z

    In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.[1] The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development of a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both combined waste stream options in terms of waste loading and/or decay time required before treatment. For Option 1, glass ceramics show an increase in waste loading of 15 mass % and reduction in decay time of 24 years. Decay times of {approx}50 years or longer are close to the expected age of the fuel that will be reprocessed when the modified open or closed fuel cycle is expected to be put into action. Option 2 shows a 2x to 2.5x increase in waste loading with decay times of only 45 years. Note that for Option 2 glass, the required decay time before treatment is only 35 years because of the waste loading limits related to the solubility of MoO{sub 3} in glass. If glass was evaluated for similar waste loadings as those achieved in Option 2 glass ceramics, the decay time would be significantly longer than 45 years. These glass ceramics are not optimized, but already they show the potential to dramatically reduce the amount of waste generated while still utilizing the proven processing technology used for glass production.

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

    Office of Environmental Management (EM)

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

  18. Integrated data base for 1993: US spent fuel and radioactive waste inventories, projections, and characteristics. Revision 9

    SciTech Connect (OSTI)

    Klein, J.A.; Storch, S.N.; Ashline, R.C. [and others

    1994-03-01T23:59:59.000Z

    The Integrated Data Base (IDB) Program has compiled historic data on inventories and characteristics of both commercial and DOE spent fuel; also, commercial and U.S. government-owned radioactive wastes through December 31, 1992. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest U.S. Department of Energy/Energy Information Administration (DOE/EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste (HLW), transuranic (TRU), waste, low-level waste (LLW), commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) LLW. For most of these categories, current and projected inventories are given through the calendar-year (CY) 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal.

  19. PROGRESS & CHALLENGES IN CLEANUP OF HANFORDS TANK WASTES

    SciTech Connect (OSTI)

    HEWITT, W.M.; SCHEPENS, R.

    2006-01-23T23:59:59.000Z

    The River Protection Project (RPP), which is managed by the Department of Energy (DOE) Office of River Protection (ORP), is highly complex from technical, regulatory, legal, political, and logistical perspectives and is the largest ongoing environmental cleanup project in the world. Over the past three years, ORP has made significant advances in its planning and execution of the cleanup of the Hartford tank wastes. The 149 single-shell tanks (SSTs), 28 double-shell tanks (DSTs), and 60 miscellaneous underground storage tanks (MUSTs) at Hanford contain approximately 200,000 m{sup 3} (53 million gallons) of mixed radioactive wastes, some of which dates back to the first days of the Manhattan Project. The plan for treating and disposing of the waste stored in large underground tanks is to: (1) retrieve the waste, (2) treat the waste to separate it into high-level (sludge) and low-activity (supernatant) fractions, (3) remove key radionuclides (e.g., Cs-137, Sr-90, actinides) from the low-activity fraction to the maximum extent technically and economically practical, (4) immobilize both the high-level and low-activity waste fractions by vitrification, (5) interim store the high-level waste fraction for ultimate disposal off-site at the federal HLW repository, (6) dispose the low-activity fraction on-site in the Integrated Disposal Facility (IDF), and (7) close the waste management areas consisting of tanks, ancillary equipment, soils, and facilities. Design and construction of the Waste Treatment and Immobilization Plant (WTP), the cornerstone of the RPP, has progressed substantially despite challenges arising from new seismic information for the WTP site. We have looked closely at the waste and aligned our treatment and disposal approaches with the waste characteristics. For example, approximately 11,000 m{sup 3} (2-3 million gallons) of metal sludges in twenty tanks were not created during spent nuclear fuel reprocessing and have low fission product concentrations. We plan to treat these wastes as transuranic waste (TRU) for disposal at the Waste Isolation Pilot Plant (WIPP), which will reduce the WTP system processing time by three years. We are also developing and testing bulk vitrification as a technology to supplement the WTP LAW vitrification facility for immobilizing the massive volume of LAW. We will conduct a full-scale demonstration of the Demonstration Bulk Vitrification System by immobilizing up to 1,100 m{sup 3} (300,000 gallons) of tank S-109 low-curie soluble waste from which Cs-137 had previously been removed. This past year has been marked by both progress and new challenges. The focus of our tank farm work has been retrieving waste from the old single-shell tanks (SSTs). We have completed waste retrieval from three SSTs and are conducting retrieval operations on an additional three SSTs. While most waste retrievals have gone about as expected, we have faced challenges with some recalcitrant tank heel wastes that required enhanced approaches. Those enhanced approaches ranged from oxalic acid additions to deploying a remote high-pressure water lance. As with all large, long-term projects that employ first of a kind technologies, we continue to be challenged to control costs and maintain schedule. However, it is most important to work safely and to provide facilities that will do the job they are intended to do.

  20. Nuclear Waste Technical Review Board Performance Plan

    E-Print Network [OSTI]

    to disposing of spent nuclear fuel and high-level radioactive waste were set forth by Congress in the NWPA. The goals are to develop a repository or repositories for disposing of high-level radioactive waste spent nuclear fuel and high- level radioactive waste. The Board's general goals and strategic objectives

  1. TECHNICAL ASSESSMENT OF FRACTIONAL CRYSTALLIZATION FOR TANK WASTE PRETREATMENT AT THE DOE HANFORD SITE

    SciTech Connect (OSTI)

    HAMILTON, D.W.

    2006-01-03T23:59:59.000Z

    Radioactive wastes from one hundred seventy-seven underground storage tanks in the 200 Area of the Department of Energy (DOE) Hanford Site in Washington State will be retrieved, treated and stored either on site or at an approved off-site repository. DOE is currently planning to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions, which would be treated and permanently disposed in separate facilities. A significant volume of the wastes in the Hanford tanks is currently classified as medium Curie waste, which will require separation and treatment at the Waste Treatment Plant (WTP). Because of the specific challenges associated with treating this waste stream, DOE EM-21 funded a project to investigate the feasibility of using fractional crystallization as a supplemental pretreatment technology. The two process requirements for fractional crystallization to be successfully applied to Hanford waste include: (1) evaporation of water from the aqueous solution to enrich the activity of soluble {sup 137}Cs, resulting in a higher activity stream to be sent to the WTP, and (2) separation of the crystalline salts that are enriched in sodium, carbonate, sulfate, and phosphate and sufficiently depleted in {sup 137}Cs, to produce a second stream to be sent to Bulk Vitrification. Phase I of this project has just been completed by COGEMA/Georgia Institute of Technology. The purpose of this report is to document an independent expert review of the Phase I results with recommendations for future testing. A team of experts with significant experience at both the Hanford and Savannah River Sites was convened to conduct the review at Richland, Washington the week of November 14, 2005.

  2. RADIOACTIVE DEMONSTRATION OF FINAL MINERALIZED WASTE FORMS FOR HANFORD WASTE TREATMENT PLANT SECONDARY WASTE BY FLUIDIZED BED STEAM REFORMING USING THE BENCH SCALE REFORMER PLATFORM

    SciTech Connect (OSTI)

    Crawford, C.; Burket, P.; Cozzi, A.; Daniel, W.; Jantzen, C.; Missimer, D.

    2012-02-02T23:59:59.000Z

    The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as {sup 137}Cs, {sup 129}I, {sup 99}Tc, Cl, F, and SO{sub 4} that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap (that could minimize volatilization). The current waste disposal path for the WTP-SW is to process it through the Effluent Treatment Facility (ETF). Fluidized Bed Steam Reforming (FBSR) is being considered for immobilization of the ETF concentrate that would be generated by processing the WTP-SW. The focus of this current report is the WTP-SW. FBSR offers a moderate temperature (700-750 C) continuous method by which WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product is being investigated to prevent dispersion during transport or burial/storage, but is not necessary for performance. A Benchscale Steam Reformer (BSR) was designed and constructed at the SRNL to treat actual radioactive wastes to confirm the findings of the non-radioactive FBSR pilot scale tests and to qualify the waste form for applications at Hanford. BSR testing with WTP SW waste surrogates and associated analytical analyses and tests of granular products (GP) and monoliths began in the Fall of 2009, and then was continued from the Fall of 2010 through the Spring of 2011. Radioactive testing commenced in 2010 with a demonstration of Hanford's WTP-SW where Savannah River Site (SRS) High Level Waste (HLW) secondary waste from the Defense Waste Processing Facility (DWPF) was shimmed with a mixture of {sup 125/129}I and {sup 99}Tc to chemically resemble WTP-SW. Prior to these radioactive feed tests, non-radioactive simulants were also processed. Ninety six grams of radioactive granular product were made for testing and comparison to the non-radioactive pilot scale tests. The same mineral phases were found in the radioactive and non-radioactive testing.

  3. WM '04 Conference, February 29 March 4, 2004, Tucson, AZ WM-4010 VITRIFICATION OF LOW AND INTERMEDIATE LEVEL WASTE: TECHNOLOGY

    E-Print Network [OSTI]

    Sheffield, University of

    waste (HLW) from spent nuclear fuel reprocessing. Development of vitrification technologies for low waste from nuclear power plants (NPP) and institutional radioactive waste in borosilicate glass matrices radioactive waste (LILW), may result in significant reduction of transport and disposal costs. Development

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

    SciTech Connect (OSTI)

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

    2011-02-24T23:59:59.000Z

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

  5. Deployment of Performance Management Methodology as part of Liquid Waste Program at Savannah River Site - 12178

    SciTech Connect (OSTI)

    Prod'homme, A.; Drouvot, O.; Gregory, J. [AREVA, Paris (France); Barnes, B.; Hodges, B.; Hart, M. [SRR, Aiken, SC (United States)

    2012-07-01T23:59:59.000Z

    In 2009, Savannah River Remediation LLC (SRR) assumed the management lead of the Liquid Waste (LW) Program at the Savannah River Site (SRS). The four SRR partners and AREVA, as an integrated subcontractor are performing the ongoing effort to safely and reliably: - Close High Level Waste (HLW) storage tanks; - Maximize waste throughput at the Defense Waste Processing Facility (DWPF); - Process salt waste into stable final waste form; - Manage the HLW liquid waste material stored at SRS. As part of these initiatives, SRR and AREVA deployed a performance management methodology based on Overall Equipment Effectiveness (OEE) at the DWPF in order to support the required production increase. This project took advantage of lessons learned by AREVA through the deployment of Total Productive Maintenance and Visual Management methodologies at the La Hague reprocessing facility in France. The project also took advantage of measurement data collected from different steps of the DWPF process by the SRR team (Melter Engineering, Chemical Process Engineering, Laboratory Operations, Plant Operations). Today the SRR team has a standard method for measuring processing time throughout the facility, a reliable source of objective data for use in decision-making at all levels, and a better balance between engineering department goals and operational goals. Preliminary results show that the deployment of this performance management methodology to the LW program at SRS has already significantly contributed to the DWPF throughput increases and is being deployed in the Saltstone facility. As part of the liquid waste program on Savannah River Site, SRR committed to enhance production throughput of DWPF. Beyond technical modifications implemented at different location of the facility, SRR deployed performance management methodology based on OEE metrics. The implementation benefited from the experience gained by AREVA in its own facilities in France. OEE proved to be a valuable tool in order to support the enhancement program in DWPF by providing unified metrics to measure plant performances, identify bottleneck location, and rank the most time consuming causes from objective data shared between the different groups belonging to the organization. Beyond OEE, the Visual Management tool adapted from the one used at La Hague were also provided in order to further enhance communication within the operating teams. As a result of all the initiatives implemented on DWPF, achieved production has been increased to record rates from FY10 to FY11. It is expected that thanks to the performance management tools now available within DWPF, these results will be sustained and even improved in the future to meet system plan targets. (authors)

  6. Final Report - "Foaming and Antifoaming and Gas Entrainment in Radioactive Waste Pretreatment and Immobilization Processes"

    SciTech Connect (OSTI)

    Wasan, Darsh T.

    2007-10-09T23:59:59.000Z

    The Savannah River Site (SRS) and Hanford site are in the process of stabilizing millions of gallons of radioactive waste slurries remaining from production of nuclear materials for the Department of Energy (DOE). The Defense Waste Processing Facility (DWPF) at SRS is currently vitrifying the waste in borosilicate glass, while the facilities at the Hanford site are in the construction phase. Both processes utilize slurry-fed joule-heated melters to vitrify the waste slurries. The DWPF has experienced difficulty during operations. The cause of the operational problems has been attributed to foaming, gas entrainment and the rheological properties of the process slurries. The rheological properties of the waste slurries limit the total solids content that can be processed by the remote equipment during the pretreatment and meter feed processes. Highly viscous material can lead to air entrainment during agitation and difficulties with pump operations. Excessive foaming in waste evaporators can cause carryover of radionuclides and non-radioactive waste to the condensate system. Experimental and theoretical investigations of the surface phenomena, suspension rheology and bubble generation of interactions that lead to foaming and air entrainment problems in the DOE High Level and Low Activity Radioactive Waste separation and immobilization processes were pursued under this project. The first major task accomplished in the grant proposal involved development of a theoretical model of the phenomenon of foaming in a three-phase gas-liquid-solid slurry system. This work was presented in a recently completed Ph.D. thesis (9). The second major task involved the investigation of the inter-particle interaction and microstructure formation in a model slurry by the batch sedimentation method. Both experiments and modeling studies were carried out. The results were presented in a recently completed Ph.D. thesis. The third task involved the use of laser confocal microscopy to study the effectiveness of three slurry rheology modifiers. An effective modifier was identified which resulted in lowering the yield stress of the waste simulant. Therefore, the results of this research have led to the basic understanding of the foaming/antifoaming mechanism in waste slurries as well as identification of a rheology modifier, which enhances the processing throughput, and accelerates the DOE mission. The objectives of this research effort were to develop a fundamental understanding of the physico-chemical mechanisms that produced foaming and air entrainment in the DOE High Level (HLW) and Low Activity (LAW) radioactive waste separation and immobilization processes, and to develop and test advanced antifoam/defoaming/rheology modifier agents. Antifoams/rheology modifiers developed from this research ere tested using non-radioactive simulants of the radioactive wastes obtained from Hanford and the Savannah River Site (SRS).

  7. Geological problems in radioactive waste isolation - A world wide review

    SciTech Connect (OSTI)

    Witherspoon, P.A. [Lawrence Berkeley Lab., CA (United States)

    1991-06-01T23:59:59.000Z

    The problem of isolating radioactive wastes from the biosphere presents specialists in the earth sciences with some of the most complicated problems they have ever encountered. This is especially true for high-level waste (HLW), which must be isolated in the underground and away from the biosphere for thousands of years. The most widely accepted method of doing this is to seal the radioactive materials in metal canisters that are enclosed by a protective sheath and placed underground in a repository that has been carefully constructed in an appropriate rock formation. Much new technology is being developed to solve the problems that have been raised, and there is a continuing need to publish the results of new developments for the benefit of all concerned. Table 1 presents a summary of the various formations under investigation according to the reports submitted for this world wide review. It can be seen that in those countries that are searching for repository sites, granitic and metamorphic rocks are the prevalent rock type under investigation. Six countries have developed underground research facilities that are currently in use. All of these investigations are in saturated systems below the water table, except the United States project, which is in the unsaturated zone of a fractured tuff.

  8. The solubilities of significant organic compounds in HLW tank supernate solutions -- FY 1995 progress report

    SciTech Connect (OSTI)

    Barney, G.S.

    1996-04-26T23:59:59.000Z

    At the Hanford Site organic compounds were measured in tank supernate simulant solutions during FY 1995. This solubility information will be used to determine if these organic salts could exist in solid phases (saltcake or sludges) in the waste where they might react violently with the nitrate or nitrite salts present in the tanks. Solubilities of sodium glycolate, succinate, and caproate salts; iron and aluminum and butylphosphate salts; and aluminum oxalate were measured in simulated waste supernate solutions at 25 {degree}C, 30 {degree}C, 40 {degree}C, and 50 {degree}C. The organic compounds were selected because they are expected to exist in relatively high concentrations in the tanks. The solubilities of sodium glycolate, succinate, caproate, and butylphosphate in HLW tank supernate solutions were high over the temperature and sodium hydroxide concentration ranges expected in the tanks. High solubilities will prevent solid sodium salts of these organic acids from precipitating from tank supernate solutions. The total organic carbon concentrations (YOC) of actual tank supernates are generally much lower than the TOC ranges for simulated supernate solutions saturated (at the solubility limit) with the organic salts. This is so even if all the dissolved carbon in a given tank and supernate is due to only one of these eight soluble compounds (an unlikely situation). Metal ion complexes of and butylphosphate and oxalate in supernate solutions were not stable in the presence of the hydroxide concentrations expected in most tanks. Iron and aluminum dibutylphosphate compounds reacted with hydroxide to form soluble sodium dibutylphosphate and precipitated iron and aluminum hydroxides. Aluminum oxalate complexes were also not stable in the basic simulated supernate solutions. Solubilities of all the organic salts decrease with increasing sodium hydroxide concentration because of the common ion effect of Na+. Increasing temperatures raised the solubilities of the organic salts, especially the succinate and caproate salts.

  9. Evaluation of Options for Permanent Geologic Disposal of Spent...

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

    policy decisions regarding strategies for the management and permanent disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) in the United States requiring...

  10. DEVELOPMENT OF A KINETIC MODEL OF BOEHMITE DISSOLUTION IN CAUSTIC SOLUTIONS APPLIED TO OPTIMIZE HANFORD WASTE PROCESSING

    SciTech Connect (OSTI)

    DISSELKAMP RS

    2011-01-06T23:59:59.000Z

    Boehmite (e.g., aluminum oxyhydroxide) is a major non-radioactive component in Hanford and Savannah River nuclear tank waste sludge. Boehmite dissolution from sludge using caustic at elevated temperatures is being planned at Hanford to minimize the mass of material disposed of as high-level waste (HLW) during operation of the Waste Treatment Plant (WTP). To more thoroughly understand the chemistry of this dissolution process, we have developed an empirical kinetic model for aluminate production due to boehmite dissolution. Application of this model to Hanford tank wastes would allow predictability and optimization of the caustic leaching of aluminum solids, potentially yielding significant improvements to overall processing time, disposal cost, and schedule. This report presents an empirical kinetic model that can be used to estimate the aluminate production from the leaching of boehmite in Hanford waste as a function of the following parameters: (1) hydroxide concentration; (2) temperature; (3) specific surface area of boehmite; (4) initial soluble aluminate plus gibbsite present in waste; (5) concentration of boehmite in the waste; and (6) (pre-fit) Arrhenius kinetic parameters. The model was fit to laboratory, non-radioactive (e.g. 'simulant boehmite') leaching results, providing best-fit values of the Arrhenius A-factor, A, and apparent activation energy, E{sub A}, of A = 5.0 x 10{sup 12} hour{sup -1} and E{sub A} = 90 kJ/mole. These parameters were then used to predict boehmite leaching behavior observed in previously reported actual waste leaching studies. Acceptable aluminate versus leaching time profiles were predicted for waste leaching data from both Hanford and Savannah River site studies.

  11. Dataflow Transformations in High-level DSP System Design

    E-Print Network [OSTI]

    Bhattacharyya, Shuvra S.

    in the application that can be analyzed and transformed for effective optimization. Because of the high level of abstraction at which they operate, the transformations employed in this process have a large impact on keyDataflow Transformations in High-level DSP System Design (Invited Paper) Sankalita Saha, Sebastian

  12. WIPP TRANSURANIC WASTE How has the WIPP TRU Waste Inventory Changed

    E-Print Network [OSTI]

    of tank waste from the Hanford site that is currently managed as high-level waste. None of this waste has that these Hanford tank wastes will be treated and will eventually be able to meet the WIPP waste acceptance criteria on the Hanford Tank Waste and K-Basin Sludges that were included in the waste inventory for recertifica- tion

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

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

    due to releases of radionuclides and chemicals from the high-level radioactive waste tanks, Fast Flux Test Facility decommissioning, and waste management activities over long...

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

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

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

  15. aqueous radioactive waste: Topics by E-print Network

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

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

  16. acidic radioactive waste: Topics by E-print Network

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

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

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

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

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

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

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

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

  19. Hanford Waste Treatment Plant completes critical system design for High-Level Waste Vitrification Facility

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsingFun with Bigfront.jpgcommunity200cell 9 Hanford Traffic Safety144&

  20. Hanford Waste Treatment Plant completes critical system design for High-Level Waste Vitrification Facility

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsingFun with Bigfront.jpgcommunity200cell 9 Hanford Traffic Safety144&8, 2010

  1. EM Waste Acceptance Product Specification (WAPS) for Vitrified High-Level Waste Forms

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power SystemsResources DOE ZeroThreeEnergyDepartment0: DOE512:Shines with Five DOE One2/2012EM

  2. Radioactive Waste Management Manual

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

    1999-07-09T23:59:59.000Z

    This Manual further describes the requirements and establishes specific responsibilities for implementing DOE O 435.1, Radioactive Waste Management, for the management of DOE high-level waste, transuranic waste, low-level waste, and the radioactive component of mixed waste. Change 1 dated 6/19/01 removes the requirement that Headquarters is to be notified and the Office of Environment, Safety and Health consulted for exemptions for use of non-DOE treatment facilities. Certified 1-9-07.

  3. High Level Overview of DOE Biomass Logistics II Project Activities

    Broader source: Energy.gov [DOE]

    Breakout Session 1B—Integration of Supply Chains I: Breaking Down Barriers High Level Overview of DOE Biomass Logistics II Project Activities Kevin Comer, Associate Principal, Antares Group Inc.

  4. HANFORD MEDIUM & LOW CURIE WASTE PRETREATMENT PROJECT PHASE 1 LAB REPORT

    SciTech Connect (OSTI)

    HAMILTON, D.W.

    2006-01-30T23:59:59.000Z

    A fractional crystallization (FC) process is being developed to supplement tank waste pretreatment capabilities provided by the Waste Treatment and Immobilization Plant (WTP). FC can process many tank wastes, separating wastes into a low-activity fraction (LAW) and high-activity fraction (HLW). The low-activity fraction can be immobilized in a glass waste form by processing in the bulk vitrification (BV) system.

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

    SciTech Connect (OSTI)

    Herman, Connie C.

    2013-09-30T23:59:59.000Z

    The Savannah River National Laboratory (SRNL) has evaluated the existing waste feed qualification strategy for the Hanford Tank Waste Treatment and Immobilization Plant (WTP) based on experience from the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) waste qualification program. The current waste qualification programs for each of the sites are discussed in the report to provide a baseline for comparison. Recommendations on strategies are then provided that could be implemented at Hanford based on the successful Macrobatch qualification strategy utilized at SRS to reduce the risk of processing upsets or the production of a staged waste campaign that does not meet the processing requirements of the WTP. Considerations included the baseline WTP process, as well as options involving Direct High Level Waste (HLW) and Low Activity Waste (LAW) processing, and the potential use of a Tank Waste Characterization and Staging Facility (TWCSF). The main objectives of the Hanford waste feed qualification program are to demonstrate compliance with the Waste Acceptance Criteria (WAC), determine waste processability, and demonstrate unit operations at a laboratory scale. Risks to acceptability and successful implementation of this program, as compared to the DWPF Macro-Batch qualification strategy, include: Limitations of mixing/blending capability of the Hanford Tank Farm; The complexity of unit operations (i.e., multiple chemical and mechanical separations processes) involved in the WTP pretreatment qualification process; The need to account for effects of blending of LAW and HLW streams, as well as a recycle stream, within the PT unit operations; and The reliance on only a single set of unit operations demonstrations with the radioactive qualification sample. This later limitation is further complicated because of the 180-day completion requirement for all of the necessary waste feed qualification steps. The primary recommendations/changes include the following: Collection and characterization of samples for relevant process analytes from the tanks to be blended during the staging process; Initiation of qualification activities earlier in the staging process to optimize the campaign composition through evaluation from both a processing and glass composition perspective; Definition of the parameters that are important for processing in the WTP facilities (unit operations) across the anticipated range of wastes and as they relate to qualification-scale equipment; Performance of limited testing with simulants ahead of the waste feed qualification sample demonstration as needed to determine the available processing window for that campaign; and Demonstration of sufficient mixing in the staging tank to show that the waste qualification sample chemical and physical properties are representative of the transfers to be made to WTP. Potential flowcharts for derivatives of the Hanford waste feed qualification process are also provided in this report. While these recommendations are an extension of the existing WTP waste qualification program, they are more in line with the processes currently performed for SRS. The implementation of these processes at SRS has been shown to offer flexibility for processing, having identified potential processing issues ahead of the qualification or facility processing, and having provided opportunity to optimize waste loading and throughput in the DWPF.

  6. Alkaline-Side Extraction of Cesium from Savannah River Tank Waste Using a Calixarene-Crown Ether Extractant

    SciTech Connect (OSTI)

    Bonnesen, P.V.; Delmau, L.H.; Haverlock, T.J.; Moyer, B.A.

    1998-12-01T23:59:59.000Z

    Results are presented supporting the viability of the alkaline-side CSEX process as a potential replacement for the In-Tank Precipitation process for removal of cesium from aqueous high-level waste (HLW) at the Savannah River Site (SRS). Under funding from the USDOE Efficient Separations and Crosscutting program, a flowsheet was suggested in early June of 1998, and in the following four months, this flowsheet underwent extensive testing, both in batch tests at ORNL and ANL and in two centrifugal-contactor tests at ANL. To carry out these tests, the initial ESP funding was augmented by direct funds from Westinghouse Savannah River Corporation. The flowsheet employed a solvent containing a calixarene-crown hybrid compound called BoBCalixC6 that was invented at ORNL and can now be obtained commercially for government use from IBC Advanced Technologies. This special extractant is so powerful and selective that it can be used at only 0.01 M, compensating for its expense, but a modifier is required for use in an aliphatic diluent, primarily to increase the cesium distribution ratio D{sub Cs} in extraction. The modifier selected is a relatively economical fluorinated alcohol called Cs3, invented at ORNL and so far available. only from ORNL. For the flowsheet, the modifier is used at 0.2 M in the branched aliphatic kerosene Isopar{reg_sign} L. Testing at ORNL and ANL involved simulants of the SRS HLW. After extraction of the Cs from the waste simulant, the solvent is scrubbed with 0.05 M HNO{sub 3} and stripped with a solution comprised of 0.0005 M HNO{sub 3} and 0.0001 M CsNO{sub 3}. The selection of these conditions is justified in this report, both on the basis of experimental data and underlying theory.

  7. Quasicrystalline Approach to Prediting the Spinel-Nepheline Liquidus: Application to Nuclear Waste Glass Processing

    SciTech Connect (OSTI)

    Jantzen, Carol

    2005-10-10T23:59:59.000Z

    The crystal-melt equilibria in complex fifteen component melts are modeled based on quasicrystalline concepts. A pseudobinary phase diagram between acmite (which melts incongruently to a transition metal ferrite spinel) and nepheline is defined. The pseudobinary lies within the Al{sub 2}O{sub 3}-Fe{sub 2}O{sub 3}-Na{sub 2}O-SiO{sub 2} quaternary system that defines the crystallization of basalt glass melts. The pseudobinary provides the partitioning of species between the melt and the primary liquidus phases. The medium range order of the melt and the melt-crystal exchange equilibria are defined based on a constrained mathematical treatment that considers the crystallochemical coordination of the elemental species in acmite and nepheline. The liquidus phases that form are shown to be governed by the melt polymerization and the octahedral site preference energies. This quasicrystalline liquidus model has been used to prevent unwanted crystallization in the world's largest high level waste (HLW) melter for the past three years while allowing >10 wt% higher waste loadings to be processed.

  8. Reliability-Centric High-Level Synthesis , N. Mansouri

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    1 Reliability-Centric High-Level Synthesis S. Tosun , N. Mansouri , E. Arvas , M. Kandemir-down. As a result, it is becoming necessary to treat reliability as a first-class citizen in system design. In particular, reliability decisions taken early in system design can have significant benefits in terms

  9. The R Environment A high-level overview

    E-Print Network [OSTI]

    Bandyopadhyay, Antar

    The R Environment A high-level overview Deepayan Sarkar Indian Statistical Institute, Delhi 6 October 2010 Deepayan Sarkar The R Environment #12;An article in the New York Times #12;From the article R use it. #12;What exactly is R? R is a language and environment for statistical computing and graphics

  10. High-Level Synthesis of Digital Microfluidic Biochips

    E-Print Network [OSTI]

    Chakrabarty, Krishnendu

    16 High-Level Synthesis of Digital Microfluidic Biochips FEI SU Intel Corporation and KRISHNENDU CHAKRABARTY Duke University Microfluidic biochips offer a promising platform for massively parallel DNA-generation system-on-chip (SOC) designs that are expected to include microfluidic components. We propose a system

  11. Analyses of high-level radioactive glasses and sludges at the Savannah River Site

    SciTech Connect (OSTI)

    Coleman, C.J.; Bibler, N.E.; Dewberry, R.A.

    1990-01-01T23:59:59.000Z

    Reliable analyses of high level radioactive glass and sludge are necessary for successful operation of the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). This facility will convert the radioactive waste sludges at SRS into durable borosilicate glasses for final disposal in a geologic repository. Analyses that are crucial to DWPF operation and repository acceptance of the glass are measurement of the radioactive and nonradioactive composition of the waste sludges and final glasses and measurement of the Fe(II)/Fe(III) ratio in a vitrified sample of melter feed. These measurements are based on the remote dissolutions of the glass and sludge followed by appropriate chemical analyses. Glasses are dissolved by a peroxide fusion method and a method using HF, HNO{sub 3}, H{sub 3}BO{sub 3}, and HCl acids where the solutions are heated in a microwave oven. The resulting solutions are analyzed by inductively coupled plasma-atomic emission spectroscopy (ICP-AES) and atomic absorption spectroscopy (AAS) for nonradioactive elements and appropriate counting techniques for radioactive elements. Results for two radioactive glasses containing actual radioactive waste are also presented. Sludges are dissolved by the Na{sub 2}O{sub 2} fusion method and an aqua regia method. 8 refs., 4 tabs.

  12. Technical Design and Optimization of a HLW-Repository in the Gorleben Salt Dome including Detailed Design of the Sealing System - 13305

    SciTech Connect (OSTI)

    Bollingerfehr, W.; Filbert, W.; Herold, P.; Lerch, C.; Mueller-Hoeppe, N. [DBE TECHNOLOGY GmbH, Eschenstrasse 55, D-31224 Peine (Germany)] [DBE TECHNOLOGY GmbH, Eschenstrasse 55, D-31224 Peine (Germany); Charlier, F. [International nuclear safety engineering GmbH, Jesuitenstrasse 4, D-52062 Aachen (Germany)] [International nuclear safety engineering GmbH, Jesuitenstrasse 4, D-52062 Aachen (Germany); Kilger, R. [Gesellschaft fuer Anlagen und Reaktorsicherheit mbH (GRS), Boltzmannstr. 14, D-85748 Garching (Germany)] [Gesellschaft fuer Anlagen und Reaktorsicherheit mbH (GRS), Boltzmannstr. 14, D-85748 Garching (Germany)

    2013-07-01T23:59:59.000Z

    The preliminary safety analysis for a HLW repository at Gorleben, the potential repository site in Germany, takes 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 'Safety Requirements Governing the Final Disposal of Heat-Generating Radioactive Waste' as at 30 September 2010. A repository design was developed for two emplacement concepts (drift disposal and borehole disposal) mainly influenced by the thermal impact of the heat-generating waste on the host rock and taking into account mining constrains. According to the objective to create the conditions for a safe containment of the waste containers in the host rock, a closure concept consisting of backfilling and sealing measures was developed. The repository was designed in such a way that retrievability requirements can be met for all waste containers within the operating phase of the repository. In addition, it could be shown that sufficient measures for ensuring subcriticality are provided both during the operational and the post-closure phases of the repository. (authors)

  13. DATA SUMMARY REPORT SMALL SCALE MELTER TESTING OF HLW ALGORITHM GLASSES MATRIX1 TESTS VSL-07S1220-1 REV 0 7/25/07

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; PEGG IL

    2011-12-29T23:59:59.000Z

    Eight tests using different HLW feeds were conducted on the DM100-BL to determine the effect of variations in glass properties and feed composition on processing rates and melter conditions (off-gas characteristics, glass processing, foaming, cold cap, etc.) at constant bubbling rate. In over seven hundred hours of testing, the property extremes of glass viscosity, electrical conductivity, and T{sub 1%}, as well as minimum and maximum concentrations of several major and minor glass components were evaluated using glass compositions that have been tested previously at the crucible scale. Other parameters evaluated with respect to glass processing properties were +/-15% batching errors in the addition of glass forming chemicals (GFCs) to the feed, and variation in the sources of boron and sodium used in the GFCs. Tests evaluating batching errors and GFC source employed variations on the HLW98-86 formulation (a glass composition formulated for HLW C-106/AY-102 waste and processed in several previous melter tests) in order to best isolate the effect of each test variable. These tests are outlined in a Test Plan that was prepared in response to the Test Specification for this work. The present report provides summary level data for all of the tests in the first test matrix (Matrix 1) in the Test Plan. Summary results from the remaining tests, investigating minimum and maximum concentrations of major and minor glass components employing variations on the HLW98-86 formulation and glasses generated by the HLW glass formulation algorithm, will be reported separately after those tests are completed. The test data summarized herein include glass production rates, the type and amount of feed used, a variety of measured melter parameters including temperatures and electrode power, feed sample analysis, measured glass properties, and gaseous emissions rates. More detailed information and analysis from the melter tests with complete emission chemistry, glass durability, and melter operating details will be provided in the final report. A summary of the tests that were conducted is provided in Table 1. Each of the seven tests was of nominally one hundred hours in duration. Test B was conducted in two equal segments: the first with nominal additives, and the second with the replacement of borax with a mixture of boric acid and soda ash to determine the effect of alternative OPC sources on production rates and processing characteristics. Interestingly, sugar additions were required near mid points of Tests W and Z to reduce excessive foaming that severely limited feed processing rates. The sugar additions were very effective in recovering manageable processing conditions, albeit over the relatively short remainder of the test duration. Tests W and Z employed the highest melt viscosities but not by a particularly wide margin. Other tests, which did not exhibit such foaming Issues, employed higher concentrations of manganese or iron or both. These results highlight the need for the development of protocols for the a priori determination of which HLW feeds will require sugar additions and the appropriate amounts of sugar to be added in order to control foaming (and maintain throughput) without over-reduction of the melt (which could lead to molten metal formation). In total, over 8,800 kg of feed was processed to produce over 3200 kg of glass. Steady-state processing rates were achieved, and no secondary sulfate phases were observed during any of the tests. Analysis was performed on samples of the glass product taken throughout the tests to verify composition and properties. Sampling and analysis was also performed on melter exhaust to determine the effect of the feed and glass changes on melter emissions.

  14. Radioactive Waste Management Manual

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

    1999-07-09T23:59:59.000Z

    This Manual further describes the requirements and establishes specific responsibilities for implementing DOE O 435.1, Radioactive Waste Management, for the management of DOE high-level waste, transuranic waste, low-level waste, and the radioactive component of mixed waste. The purpose of the Manual is to catalog those procedural requirements and existing practices that ensure that all DOE elements and contractors continue to manage DOE's radioactive waste in a manner that is protective of worker and public health and safety, and the environment. Does not cancel other directives.

  15. Operational considerations for high level blast furnace fuel injection

    SciTech Connect (OSTI)

    Poveromo, J.J. [Quebec Cartier Mining Co., Bethlehem, PA (United States)

    1996-12-31T23:59:59.000Z

    Injection levels of over 400 lbs/NTHM for coal, over 250 lbs/NTHM for natural gas and over 200 lbs/NTHM for oil have been achieved. Such high levels of fuel injection has a major impact on many aspects of blast furnace operation. In this paper the author begins by reviewing the fundamentals of fuel injection with emphasis on raceway thermochemical phenomena. The operational impacts which are generic to high level injection of any injectant are then outlined. The author will then focus on the particular characteristics of each injectant, with major emphasis on coal and natural gas. Operational considerations for coping with these changes and methods of maximizing the benefits of fuel injection will be reviewed.

  16. Granite Recrystallization The Key to an Alternative Strategy for HLW Disposal? Fergus G.F. Gibb

    E-Print Network [OSTI]

    Sheffield, University of

    Granite Recrystallization ­ The Key to an Alternative Strategy for HLW Disposal? Fergus G.F. Gibb, and hence the key to the entire strategy, depends on whether sufficient melting of granite host rock can-temperature, high-pressure experiments reported here demonstrate that granite can be partially melted and completely

  17. HLW Salt Disposition Alternatives Identification Preconceptual Phase I Summary Report (Including Attachments)

    SciTech Connect (OSTI)

    Piccolo, S.F.

    1999-07-09T23:59:59.000Z

    The purpose of this report is to summarize the process used by the Team to systematically develop alternative methods or technologies for final disposition of HLW salt. Additionally, this report summarizes the process utilized to reduce the total list of identified alternatives to an ''initial list'' for further evaluation. This report constitutes completion of the team charter major milestone Phase I Deliverable.

  18. E-Print Network 3.0 - account radioactive waste Sample Search...

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

    42 U... of repositories for high-level radioactive waste and spent nuclear fuel. Low Level Radioactive Waste Policy Act... adequate disposal ... Source: Yucca Mountain...

  19. alamos low-level waste: Topics by E-print Network

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

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

  20. Understanding radioactive waste

    SciTech Connect (OSTI)

    Murray, R.L.

    1981-12-01T23:59:59.000Z

    This document contains information on all aspects of radioactive wastes. Facts are presented about radioactive wastes simply, clearly and in an unbiased manner which makes the information readily accessible to the interested public. The contents are as follows: questions and concerns about wastes; atoms and chemistry; radioactivity; kinds of radiation; biological effects of radiation; radiation standards and protection; fission and fission products; the Manhattan Project; defense and development; uses of isotopes and radiation; classification of wastes; spent fuels from nuclear reactors; storage of spent fuel; reprocessing, recycling, and resources; uranium mill tailings; low-level wastes; transportation; methods of handling high-level nuclear wastes; project salt vault; multiple barrier approach; research on waste isolation; legal requiremnts; the national waste management program; societal aspects of radioactive wastes; perspectives; glossary; appendix A (scientific American articles); appendix B (reference material on wastes). (ATT)

  1. Relative yields of U-235 fission products measured in a high level radioactive sludge at Savannah River Site

    SciTech Connect (OSTI)

    Bibler, N.E.; Coleman, C.J. [Westinghouse Savannah River Co., Aiken, SC (United States); Kinard, W.F. [Charleston Coll., SC (United States). Dept. of Chemistry

    1992-10-01T23:59:59.000Z

    This paper presents measurements of the concentrations of 42 of the long-lived U-235 fission products in a high-level radioactive waste sludge stored at Savannah River Site. The 42 fision products make up 98% of the waste sludge. We used inductively coupled plasma-mass spectroscopy for the analysis. The relative yields for most of the fission products are in complete agreement with the known relative yields for the beta decay chains of the two asymmetric branches of the slow neutron fission of U-235. Disagreements can be reconciled based on the chemistry of the fission products in the caustic waste sludges, the neutron fluences in SRS reactors, or interferences in the ICP-MS analyses. This paper presents measurements of the concentrations of 42 (98%) of the long-lived U-235 fission products in a high-level radioactive waste sludge stored at the Savannah River Site. We analyzed the sludge with inductively coupled plasma-mass spectroscopy. The relative yields for most of the fission products agree completely with the known relative vields for the beta decay chains of the two asymmetric: branches of the slow neutron fission of U-235. The chemistry of the fission products in the caustic waste sludges, the neutron fluences in SRS reactors, or interferences in the ICP-MS analyses explain the differences in the measured and calculated results.

  2. Relative yields of U-235 fission products measured in a high level radioactive sludge at Savannah River Site

    SciTech Connect (OSTI)

    Bibler, N.E.; Coleman, C.J. (Westinghouse Savannah River Co., Aiken, SC (United States)); Kinard, W.F. (Charleston Coll., SC (United States). Dept. of Chemistry)

    1992-01-01T23:59:59.000Z

    This paper presents measurements of the concentrations of 42 of the long-lived U-235 fission products in a high-level radioactive waste sludge stored at Savannah River Site. The 42 fision products make up 98% of the waste sludge. We used inductively coupled plasma-mass spectroscopy for the analysis. The relative yields for most of the fission products are in complete agreement with the known relative yields for the beta decay chains of the two asymmetric branches of the slow neutron fission of U-235. Disagreements can be reconciled based on the chemistry of the fission products in the caustic waste sludges, the neutron fluences in SRS reactors, or interferences