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1

PAIRWISE BLENDING OF HIGH LEVEL WASTE (HLW)  

SciTech Connect (OSTI)

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.

CERTA, P.J.

2006-02-22T23:59:59.000Z

2

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

SciTech Connect (OSTI)

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.

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

1994-07-08T23:59:59.000Z

3

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

SciTech Connect (OSTI)

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).

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

2010-01-04T23:59:59.000Z

4

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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5

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

SciTech Connect (OSTI)

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.

Aurah, Mirwaise Y.; Roberts, Mark A.

2013-12-12T23:59:59.000Z

6

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

SciTech Connect (OSTI)

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.

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

2010-01-04T23:59:59.000Z

7

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

SciTech Connect (OSTI)

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

Fox, K. M.

2014-02-27T23:59:59.000Z

8

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...

9

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...

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]

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

Sheffield, University of

11

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

SciTech Connect (OSTI)

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.

Jantzen, C; James Marra, J

2007-09-17T23:59:59.000Z

12

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

SciTech Connect (OSTI)

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.

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

13

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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14

DEVELOPMENT OF GLASS MATRICES FOR HLW RADIOACTIVE WASTES  

SciTech Connect (OSTI)

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.

Jantzen, C.

2010-03-18T23:59:59.000Z

15

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

SciTech Connect (OSTI)

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.

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

1996-02-01T23:59:59.000Z

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)

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.

KRUGER AA; FENG Z; GAN H; PEGG IL

2009-11-05T23:59:59.000Z

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)

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.

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

2009-02-07T23:59:59.000Z

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)

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)

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

19

High Level Waste System Plan Revision 9  

SciTech Connect (OSTI)

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.

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

1998-04-01T23:59:59.000Z

20

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

SciTech Connect (OSTI)

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.

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

2009-12-30T23:59:59.000Z

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21

Optimizing High Level Waste Disposal  

SciTech Connect (OSTI)

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.

Dirk Gombert

2005-09-01T23:59:59.000Z

22

Risk-informing decisions about high-level nuclear waste repositories  

E-Print Network [OSTI]

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. ...

Ghosh, Suchandra Tina, 1973-

2004-01-01T23:59:59.000Z

23

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

SciTech Connect (OSTI)

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.

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

1998-05-01T23:59:59.000Z

24

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

SciTech Connect (OSTI)

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.

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

1996-02-01T23:59:59.000Z

25

High-level waste melter alternatives assessment report  

SciTech Connect (OSTI)

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.

Calmus, R.B.

1995-02-01T23:59:59.000Z

26

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

SciTech Connect (OSTI)

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.

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

27

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

SciTech Connect (OSTI)

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.

Burgard, K.C.

1998-04-09T23:59:59.000Z

28

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

SciTech Connect (OSTI)

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.

Burgard, K.C.

1998-06-02T23:59:59.000Z

29

High-level waste issues and resolutions document  

SciTech Connect (OSTI)

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.

Not Available

1994-05-01T23:59:59.000Z

30

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

SciTech Connect (OSTI)

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).

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

31

FLUIDIZED BED STEAM REFORMING ENABLING ORGANIC HIGH LEVEL WASTE DISPOSAL  

SciTech Connect (OSTI)

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.

Williams, M

2008-05-09T23:59:59.000Z

32

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

33

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

SciTech Connect (OSTI)

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.

SK Sundaram; ML Elliott; D Bickford

1999-11-19T23:59:59.000Z

34

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

E-Print Network [OSTI]

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

Sheffield, University of

35

Defense HLW Glass Degradation Model  

SciTech Connect (OSTI)

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.

D. Strachan

2004-10-20T23:59:59.000Z

36

Review of Alternative Technologies for Pretreatment of Accumulated HLW  

SciTech Connect (OSTI)

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.

Romanovsky, V.; Rimski-Korsakov, A.

2002-02-26T23:59:59.000Z

37

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

SciTech Connect (OSTI)

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.

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

1999-05-01T23:59:59.000Z

38

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

SciTech Connect (OSTI)

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.

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

1999-01-01T23:59:59.000Z

39

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

SciTech Connect (OSTI)

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.

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

2002-02-26T23:59:59.000Z

40

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

SciTech Connect (OSTI)

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.

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

2009-10-05T23:59:59.000Z

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

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

SciTech Connect (OSTI)

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.

James A. King; Vince Maio

2011-09-01T23:59:59.000Z

42

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

SciTech Connect (OSTI)

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.

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

2013-01-09T23:59:59.000Z

43

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

SciTech Connect (OSTI)

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)

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

44

High-Level Waste Requirements  

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

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.

1999-07-09T23:59:59.000Z

45

Low-Level Waste Requirements  

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

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

1999-07-09T23:59:59.000Z

46

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

SciTech Connect (OSTI)

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.

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

1993-12-31T23:59:59.000Z

47

Nondestructive examination of DOE high-level waste storage tanks  

SciTech Connect (OSTI)

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.

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

48

Progress achieved in HLW vitrification techniques at INE  

SciTech Connect (OSTI)

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.

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

1993-12-31T23:59:59.000Z

49

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

50

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

SciTech Connect (OSTI)

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.

N /A

2002-10-11T23:59:59.000Z

51

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...

52

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

SciTech Connect (OSTI)

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.

THIEN MG; WELLS BE; ADAMSON DJ

2010-01-14T23:59:59.000Z

53

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)

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)

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

54

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

SciTech Connect (OSTI)

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.

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

55

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

SciTech Connect (OSTI)

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.

N /A

2002-05-31T23:59:59.000Z

56

DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER  

SciTech Connect (OSTI)

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.

G. Radulesscu; J.S. Tang

2000-06-07T23:59:59.000Z

57

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

58

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

SciTech Connect (OSTI)

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.

Sandra M Birk

2010-10-01T23:59:59.000Z

59

RIVER PROTECTION PROJECT MISSION ANALYSIS WASTE BLENDING STUDY  

SciTech Connect (OSTI)

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.

SHUFORD DH; STEGEN G

2010-04-19T23:59:59.000Z

60

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)

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.

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

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

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

SciTech Connect (OSTI)

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.

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

62

Low Level Radioactive Waste Authority (Michigan)  

Broader source: Energy.gov [DOE]

Federal laws passed in 1980 and 1985 made each state responsible for the low-level radioactive waste produced within its borders. Act 204 of 1987 created the Low-Level Radioactive Waste Authority ...

63

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

SciTech Connect (OSTI)

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.

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

64

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

SciTech Connect (OSTI)

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.

Fellinger, A.

2009-12-08T23:59:59.000Z

65

EMPIRICAL MODEL FOR FORMULATION OF CRYSTAL-TOLERANT HLW GLASSES  

SciTech Connect (OSTI)

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.

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

2012-03-07T23:59:59.000Z

66

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

SciTech Connect (OSTI)

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).

Hansen, E

2005-03-31T23:59:59.000Z

67

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

SciTech Connect (OSTI)

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.

BRIGGS, M.G.

2000-09-22T23:59:59.000Z

68

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

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 Systems EngineeringDepartment of

69

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

70

Development of Crystal-Tolerant High-Level Waste Glasses  

SciTech Connect (OSTI)

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.

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

71

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

SciTech Connect (OSTI)

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.

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

1984-01-01T23:59:59.000Z

72

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

73

Certification Plan, low-level waste Hazardous Waste Handling Facility  

SciTech Connect (OSTI)

The purpose of this plan is to describe the organization and methodology for the certification of low-level radioactive waste (LLW) handled in the Hazardous Waste Handling Facility (HWHF) at Lawrence Berkeley Laboratory (LBL). This plan also incorporates the applicable elements of waste reduction, which include both up-front minimization and end-product treatment to reduce the volume and toxicity of the waste; segregation of the waste as it applies to certification; an executive summary of the Waste Management Quality Assurance Implementing Management Plan (QAIMP) for the HWHF and a list of the current and planned implementing procedures used in waste certification. This plan provides guidance from the HWHF to waste generators, waste handlers, and the Waste Certification Specialist to enable them to conduct their activities and carry out their responsibilities in a manner that complies with the requirements of WHC-WAC. Waste generators have the primary responsibility for the proper characterization of LLW. The Waste Certification Specialist verifies and certifies that LBL LLW is characterized, handled, and shipped in accordance with the requirements of WHC-WAC. Certification is the governing process in which LBL personnel conduct their waste generating and waste handling activities in such a manner that the Waste Certification Specialist can verify that the requirements of WHC-WAC are met.

Albert, R.

1992-06-30T23:59:59.000Z

74

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.

75

Low-level waste forum meeting reports  

SciTech Connect (OSTI)

This report contains highlights from the 1991 fall meeting of the Low Level Radioactive Waste Forum. Topics included legal updates; US NRC updates; US EPA updates; mixed waste issues; financial assistance for waste disposal facilities; and a legislative and policy report.

NONE

1991-12-31T23:59:59.000Z

76

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

SciTech Connect (OSTI)

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.

Reimann, G.A.

1991-10-01T23:59:59.000Z

77

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

SciTech Connect (OSTI)

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.

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

2002-02-27T23:59:59.000Z

78

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

SciTech Connect (OSTI)

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.

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

79

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

SciTech Connect (OSTI)

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.%.

Stefanovsky, Sergey; Marra, James; Lebedev, Vladimir

2014-01-13T23:59:59.000Z

80

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

SciTech Connect (OSTI)

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.

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

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

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

SciTech Connect (OSTI)

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)

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

82

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

SciTech Connect (OSTI)

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.

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

2013-05-31T23:59:59.000Z

83

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)

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.

Alexandre, D.; Baker, R.; Kohout, R.; Marke, J. [eds.

1993-12-31T23:59:59.000Z

84

ESTIMATING HIGH LEVEL WASTE MIXING PERFORMANCE IN HANFORD DOUBLE SHELL TANKS  

SciTech Connect (OSTI)

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.

THIEN MG; GREER DA; TOWNSON P

2011-01-13T23:59:59.000Z

85

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

SciTech Connect (OSTI)

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.

Veronica J Rutledge; Vince Maio

2013-10-01T23:59:59.000Z

86

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

SciTech Connect (OSTI)

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.

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

2012-01-12T23:59:59.000Z

87

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

SciTech Connect (OSTI)

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.

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

2013-11-13T23:59:59.000Z

88

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)

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.

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

2008-12-31T23:59:59.000Z

89

Low-level waste forum meeting reports  

SciTech Connect (OSTI)

This paper provides highlights from the 1992 winter meeting of the Low Level Radioactive Wastes Forum. Topics of discussion included: legal information; state and compact reports; freedom of information requests; and storage.

Sternwheeler, W.D.E.

1992-12-31T23:59:59.000Z

90

Low-level waste forum meeting reports  

SciTech Connect (OSTI)

This report provides highlights from the 1992 fall meeting of the Low LEvel Radioactive Waste Forum. Topics included: disposal options after 1992; interregional agreements; management alternatives; policy; and storage.

NONE

1992-12-31T23:59:59.000Z

91

Mixed low-level waste form evaluation  

SciTech Connect (OSTI)

A scoping level evaluation of polyethylene encapsulation and vitreous waste forms for safe storage of mixed low-level waste was performed. Maximum permissible radionuclide concentrations were estimated for 15 indicator radionuclides disposed of at the Hanford and Savannah River sites with respect to protection of the groundwater and inadvertent intruder pathways. Nominal performance improvements of polyethylene and glass waste forms relative to grout are reported. These improvements in maximum permissible radionuclide concentrations depend strongly on the radionuclide of concern and pathway. Recommendations for future research include improving the current understanding of the performance of polymer waste forms, particularly macroencapsulation. To provide context to these estimates, the concentrations of radionuclides in treated DOE waste should be compared with the results of this study to determine required performance.

Pohl, P.I.; Cheng, Wu-Ching; Wheeler, T.; Waters, R.D.

1997-03-01T23:59:59.000Z

92

Low-level waste forum meeting reports  

SciTech Connect (OSTI)

This paper provides highlights from the spring meeting of the Low Level Radioactive Waste Forum. Topics of discussion included: state and compact reports; New York`s challenge to the constitutionality of the Low-Level Radioactive Waste Amendments Act of 1985; DOE technical assistance for 1993; interregional import/export agreements; Department of Transportation requirements; superfund liability; nonfuel bearing components; NRC residual radioactivity criteria.

NONE

1992-12-31T23:59:59.000Z

93

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

SciTech Connect (OSTI)

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.

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

1998-12-01T23:59:59.000Z

94

Southwestern Low-Level Radioactive Waste Disposal Compact (South Dakota)  

Broader source: Energy.gov [DOE]

This legislation authorizes the state's entrance into the Southwestern Low-Level Radioactive Waste Disposal Compact, which provides for the cooperative management of low-level radioactive waste....

95

WM '04 Conference, February 29 March 4, 2004, Tucson, AZ WM-4010 VITRIFICATION OF LOW AND INTERMEDIATE LEVEL WASTE: TECHNOLOGY  

E-Print Network [OSTI]

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

Sheffield, University of

96

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

SciTech Connect (OSTI)

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.

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

1984-01-01T23:59:59.000Z

97

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

SciTech Connect (OSTI)

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.

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

2003-02-24T23:59:59.000Z

98

Twelfth annual US DOE low-level waste management conference  

SciTech Connect (OSTI)

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

Not Available

1990-01-01T23:59:59.000Z

99

High-Level Waste Melter Study Report  

SciTech Connect (OSTI)

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.

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

100

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)

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.

Aaron, G.; Wilmarth, B.

2011-09-19T23:59:59.000Z

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Solid low-level waste forecasting guide  

SciTech Connect (OSTI)

Guidance for forecasting solid low-level waste (LLW) on a site-wide basis is described in this document. Forecasting is defined as an approach for collecting information about future waste receipts. The forecasting approach discussed in this document is based solely on hanford`s experience within the last six years. Hanford`s forecasting technique is not a statistical forecast based upon past receipts. Due to waste generator mission changes, startup of new facilities, and waste generator uncertainties, statistical methods have proven to be inadequate for the site. It is recommended that an approach similar to Hanford`s annual forecasting strategy be implemented at each US Department of Energy (DOE) installation to ensure that forecast data are collected in a consistent manner across the DOE complex. Hanford`s forecasting strategy consists of a forecast cycle that can take 12 to 30 months to complete. The duration of the cycle depends on the number of LLW generators and staff experience; however, the duration has been reduced with each new cycle. Several uncertainties are associated with collecting data about future waste receipts. Volume, shipping schedule, and characterization data are often reported as estimates with some level of uncertainty. At Hanford, several methods have been implemented to capture the level of uncertainty. Collection of a maximum and minimum volume range has been implemented as well as questionnaires to assess the relative certainty in the requested data.

Templeton, K.J.; Dirks, L.L.

1995-03-01T23:59:59.000Z

102

HLW system plan - revision 2  

SciTech Connect (OSTI)

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.

Not Available

1994-01-14T23:59:59.000Z

103

High-level radioactive wastes. Supplement 1  

SciTech Connect (OSTI)

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.

McLaren, L.H. (ed.)

1984-09-01T23:59:59.000Z

104

High-level waste qualification: Managing uncertainty  

SciTech Connect (OSTI)

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.

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

1993-12-31T23:59:59.000Z

105

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

SciTech Connect (OSTI)

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.

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

2005-09-19T23:59:59.000Z

106

Potential Application Of Radionuclide Scaling Factors To High Level Waste Characterization  

SciTech Connect (OSTI)

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.

Reboul, S. H.

2013-09-30T23:59:59.000Z

107

Low level tank waste disposal study  

SciTech Connect (OSTI)

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

Mullally, J.A.

1994-09-29T23:59:59.000Z

108

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

SciTech Connect (OSTI)

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.

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

109

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

SciTech Connect (OSTI)

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.

N /A

2005-06-30T23:59:59.000Z

110

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

SciTech Connect (OSTI)

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.

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

2014-05-15T23:59:59.000Z

111

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

SciTech Connect (OSTI)

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.

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

2002-02-26T23:59:59.000Z

112

Cement encapsulation of intermediate-level waste slurries  

SciTech Connect (OSTI)

Reprocessing of irradiated nuclear fuel at BNFL`s Sellafield site produces a range of radioactive wastes. BNFL has adopted a detailed policy for radioactive waste management in order that: effluent discharges to the environment are minimized, solid low-level waste is safely disposed of as it arises, and all other wastes are stored, conditioned and treated for eventual disposal.

Lewis, H.G.; Cassidy, C.M. [British Nuclear Fuels plc, Risley (United Kingdom)

1993-12-31T23:59:59.000Z

113

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

114

Lid design for low level waste container  

DOE Patents [OSTI]

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

Holbrook, Richard H. (Clinton, TN); Keener, Wendell E. (Lenior City, TN)

1995-01-01T23:59:59.000Z

115

Lid design for low level waste container  

DOE Patents [OSTI]

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

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

1995-02-28T23:59:59.000Z

116

Solid low level waste forms and extended storage  

SciTech Connect (OSTI)

This paper presents regulatory, technical, and economic aspects of selecting solid waste forms for the extended on-site storage of power plant low level wastes (LLW) in the United States. The author explains current uncertainties and disposal site shortages, defines power plant waste types, addresses regulatory requirements for disposal, discusses basic waste form storage considerations, outlines possible strategies for the management of individual waste types, and offers methodological steps for selecting a waste form for extended storage. Broader issues closely associated with waste form selection are also presented.

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

1995-11-01T23:59:59.000Z

117

Low-Level Radioactive Waste Disposal Act (Pennsylvania)  

Broader source: Energy.gov [DOE]

This act provides a comprehensive strategy for the siting of commercial low-level waste compactors and other waste management facilities, and to ensure the proper transportation, disposal and...

118

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

SciTech Connect (OSTI)

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.

Hansen, E; Alex Cozzi, A

2008-06-19T23:59:59.000Z

119

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

SciTech Connect (OSTI)

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.

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

2002-02-25T23:59:59.000Z

120

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

E-Print Network [OSTI]

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

Sheffield, University of

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121

Vitrified waste option study report  

SciTech Connect (OSTI)

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.

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

1998-02-01T23:59:59.000Z

122

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

SciTech Connect (OSTI)

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.

Vince Maio

2014-08-01T23:59:59.000Z

123

Handbook of high-level radioactive waste transportation  

SciTech Connect (OSTI)

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.

Sattler, L.R.

1992-10-01T23:59:59.000Z

124

Appalachian States Low-Level Radioactive Waste Compact (Maryland)  

Broader source: Energy.gov [DOE]

This legislation authorizes Maryland's entrance into the Appalachian States Low-Level Radioactive Waste Compact, which seeks to promote interstate cooperation for the proper management and disposal...

125

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

SciTech Connect (OSTI)

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.

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

126

Low-Level Burial Grounds Waste Analysis Plan  

SciTech Connect (OSTI)

The purpose of this waste analysis plan (WAP) is to document the waste acceptance process, sampling methodologies, analytical techniques, and overall processes that are undertaken for waste accepted for storage and/or disposal at the Low-Level Burial Grounds which are located in the 200 East and West Areas of the Hanford Facility, Richland, Washington. This WAP documents the methods used to characterize, obtain and analyze representative samples of waste managed at this unit.

ELLEFSON, M.D.

2000-03-02T23:59:59.000Z

127

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

SciTech Connect (OSTI)

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.

Quigley, K.D.; Wessman, D

2003-02-27T23:59:59.000Z

128

A model for a national low level waste program  

SciTech Connect (OSTI)

A national program for the management of low level waste is essential to the success of environmental clean-up, decontamination and decommissioning, current operations and future missions. The value of a national program is recognized through procedural consistency and a shared set of resources. A national program requires a clear waste definition and an understanding of waste characteristics matched against available and proposed disposal options. A national program requires the development and implementation of standards and procedures for implementing the waste hierarchy, with a specitic emphasis on waste avoidance, minimization and recycling. It requires a common set of objectives for waste characterization based on the disposal facility's waste acceptance criteria, regulatory and license requirements and performance assessments. Finally, a national waste certification program is required to ensure compliance. To facilitate and enhance the national program, a centralized generator services organization, tasked with providing technical services to the generators on behalf of the national program, is necessary. These subject matter experts are the interface between the generating sites and the disposal facility(s). They provide an invaluable service to the generating organizations through their involvement in waste planning prior to waste generation and through championing implementation of the waste hierarchy. Through their interface, national treatment and transportation services are optimized and new business opportunities are identified. This national model is based on extensive experience in the development and on-going management of a national transuranic waste program and management of the national repository, the Waste Isolation Pilot Plant. The Low Level Program at the Savannah River Site also successfully developed and implemented the waste hierarchy, waste certification and waste generator services concepts presented below. The Savannah River Site services over forty generators and has historically managed over 12,000 cubic meters of low level waste annually. The results of the waste minimization program at the site resulted in over 900 initiatives, avoiding over 220,000 cubic meters of waste for a life cycle cost savings of $275 million. At the Los Alamos National Laboratory, the low level waste program services over 20 major generators and several hundred smaller generators that produce over 4,000 cubic meters of low level waste annually. The Los Alamos National Laboratory low level waste program utilizes both on-site and off-site disposal capabilities. Off-site disposal requires the implementation of certification requirements to utilize both federal and commercial options. The Waste Isolation Pilot Plant is the US Department of Energy's first deep geological repository for the permanent disposal of Transuanic waste. Transuranic waste was generated and retrievably stored at 39 sites across the US. Transuranic waste is defined as waste with a radionuclide concentration equal to or greater than 100 nCi/g consisting of radionuclides with half-lives greater than 20 years and with an atomic mass greater than uranium. Combining the lessons learned from the national transuranic waste program, the successful low level waste program at Savannah River Site and the experience of off-site disposal options at Los Alamos National Laboratory provides the framework and basis for developing a viable national strategy for managing low level waste.

Blankenhorn, James A [Los Alamos National Laboratory

2009-01-01T23:59:59.000Z

129

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

SciTech Connect (OSTI)

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

Not Available

1990-10-01T23:59:59.000Z

130

Melter Throughput Enhancements for High-Iron HLW  

SciTech Connect (OSTI)

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.

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

131

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)

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.

KETUSKY, EDWARD

2005-10-31T23:59:59.000Z

132

Identification of potential transuranic waste tanks at the Hanford Site  

SciTech Connect (OSTI)

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.

Colburn, R.P.

1995-05-05T23:59:59.000Z

133

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

SciTech Connect (OSTI)

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.

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

134

HLW Glass Studies: Development of Crystal-Tolerant HLW Glasses  

SciTech Connect (OSTI)

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.

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

2012-04-02T23:59:59.000Z

135

System for chemically digesting low level radioactive, solid waste material  

DOE Patents [OSTI]

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

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

1982-01-01T23:59:59.000Z

136

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

SciTech Connect (OSTI)

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.

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

2006-04-01T23:59:59.000Z

137

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)

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.

Vince Maio

2011-08-01T23:59:59.000Z

138

Exploratory study of complexant concentrate waste processing  

SciTech Connect (OSTI)

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.

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

1993-02-01T23:59:59.000Z

139

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

SciTech Connect (OSTI)

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.

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

140

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

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

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

SciTech Connect (OSTI)

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

Kirner, N.; Kelly, J.; Faison, G.; Johnson, D. [Foster Wheeler Environmental Corp. (United States)

1995-05-01T23:59:59.000Z

142

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

143

Low-level radioactive waste regulation: Science, politics and fear  

SciTech Connect (OSTI)

An inevitable consequence of the use of radioactive materials is the generation of radioactive wastes and the public policy debate over how they will be managed. In 1980, Congress shifted responsibility for the disposal of low-level radioactive wastes from the federal government to the states. This act represented a sharp departure from more than 30 years of virtually absolute federal control over radioactive materials. Though this plan had the enthusiastic support of the states in 1980, it now appears to have been at best a chimera. Radioactive waste management has become an increasingly complicated and controversial issue for society in recent years. This book discusses only low-level wastes, however, because Congress decided for political reasons to treat them differently than high-level wastes. The book is based in part on three symposia sponsored by the division of Chemistry and the Law of the American Chemical Society. Each chapter is derived in full or in part from presentations made at these meetings, and includes: (1) Low-level radioactive wastes in the nuclear power industry; (2) Low-level radiation cancer risk assessment and government regulation to protect public health; and (3) Low-level radioactive waste: can new disposal sites be found.

Burns, M.E. (ed.)

1988-01-01T23:59:59.000Z

144

Combined Waste Form Cost Trade Study  

SciTech Connect (OSTI)

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.

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

2008-11-01T23:59:59.000Z

145

Criteria for releases and disposal of low level and intermediate level waste in Sweden  

SciTech Connect (OSTI)

In Sweden there exists a complete system for management, including final disposal, of all radioactive wastes which are not classified as long-lived or high-level waste. This paper will present the disposal options and the requirements set on the waste categories as well as Sweden`s four different engineered shallow land disposals. The advantages of having a shallow land disposal together with exemption of waste and a final storage facility for low-level and intermediate-level waste are discussed. Finally, the paper will give a summary of why Sweden has succeeded in establishing a full system for low-level and intermediate-level waste. The discussion is from regulatory point of view.

Lindbom, G. [Swedish Radiation Protection Inst., Stockholm (Sweden). Div. of Waste Management and Environmental Protection

1993-12-31T23:59:59.000Z

146

Alpha low-level stored waste systems design study  

SciTech Connect (OSTI)

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

Feizollahi, F.; Teheranian, B. [Morrison Knudson Corp., San Francisco, CA (United States). Environmental Services Div.; Quapp, W.J. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1992-08-01T23:59:59.000Z

147

Alpha low-level stored waste systems design study  

SciTech Connect (OSTI)

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

Feizollahi, F.; Teheranian, B. (Morrison Knudson Corp., San Francisco, CA (United States). Environmental Services Div.); Quapp, W.J. (EG and G Idaho, Inc., Idaho Falls, ID (United States))

1992-08-01T23:59:59.000Z

148

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

SciTech Connect (OSTI)

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.

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

1996-08-01T23:59:59.000Z

149

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...

150

Low-Level Radioactive Waste Disposal Regional Facility Act (Pennsylvania)  

Broader source: Energy.gov [DOE]

This act establishes a low-level radioactive waste disposal regional facility siting fund that requires nuclear power reactor constructors and operators to pay to the Department of Environmental...

151

Costs of mixed low-level waste stabilization options  

SciTech Connect (OSTI)

Selection of final waste forms to be used for disposal of DOE`s mixed low-level waste (MLLW) depends on the waste form characteristics and total life cycle cost. In this paper the various cost factors associated with production and disposal of the final waste form are discussed and combined to develop life-cycle costs associated with several waste stabilization options. Cost factors used in this paper are based on a series of treatment system studies in which cost and mass balance analyses were performed for several mixed low-level waste treatment systems and various waste stabilization methods including vitrification, grout, phosphate bonded ceramic and polymer. Major cost elements include waste form production, final waste form volume, unit disposal cost, and system availability. Production of grout costs less than the production of a vitrified waste form if each treatment process has equal operating time (availability) each year; however, because of the lower volume of a high temperature slag, certification and handling costs and disposal costs of the final waste form are less. Both the total treatment cost and life cycle costs are higher for a system producing grout than for a system producing high temperature slag, assuming equal system availability. The treatment costs decrease with increasing availability regardless of the waste form produced. If the availability of a system producing grout is sufficiently greater than a system producing slag, then the cost of treatment for the grout system will be less than the cost for the slag system, and the life cycle cost (including disposal) may be less depending on the unit disposal cost. Treatment and disposal costs will determine the return on investment in improved system availability.

Schwinkendorf, W.E.; Cooley, C.R.

1998-03-01T23:59:59.000Z

152

Characteristics of potential repository wastes. Volume 1  

SciTech Connect (OSTI)

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.

Not Available

1992-07-01T23:59:59.000Z

153

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

SciTech Connect (OSTI)

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)

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

154

Modeling and low-level waste management: an interagency workshop  

SciTech Connect (OSTI)

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

Little, C.A.; Stratton, L.E. (comps.)

1980-01-01T23:59:59.000Z

155

Immobilized low-level waste disposal options configuration study  

SciTech Connect (OSTI)

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

Mitchell, D.E.

1995-02-01T23:59:59.000Z

156

Low-level radioactive waste disposal facility closure  

SciTech Connect (OSTI)

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

White, G.J.; Ferns, T.W.; Otis, M.D.; Marts, S.T.; DeHaan, M.S.; Schwaller, R.G.; White, G.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

1990-11-01T23:59:59.000Z

157

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

SciTech Connect (OSTI)

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

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

1997-06-23T23:59:59.000Z

158

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

SciTech Connect (OSTI)

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.

Amoroso, J.

2011-10-09T23:59:59.000Z

159

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

E-Print Network [OSTI]

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 ...

Cabeche, Dion Tunick

2011-01-01T23:59:59.000Z

160

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...

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Low-Level Waste Disposal Alternatives Analysis Report  

SciTech Connect (OSTI)

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

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

2006-09-01T23:59:59.000Z

162

National Low-Level Waste Management Program Radionuclide Report Series  

SciTech Connect (OSTI)

This volume serves as an introduction to the National Low-Level Radioactive Waste Management Program Radionuclide Report Series. This report includes discussions of radionuclides listed in Title 10 of the Code of Federal Regulations Part 61.55, Tables 1 and 2 (including alpha-emitting transuranics with half-lives greater than five years). Each report includes information regarding radiological and chemical characteristics of specific radionuclides. Information is also included discussing waste streams and waste forms that may contain each radionuclide, and radionuclide behavior in the environment and in the human body. Not all radionuclides commonly found at low-level radioactive waste sites are included in this report. The discussion in this volume explains the rationale of the radionuclide selection process.

Rudin, M.J.; Garcia, R.S.

1992-02-01T23:59:59.000Z

163

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

SciTech Connect (OSTI)

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

Mayberry, J.L.; Huebner, T.L. [Science Applications International Corp., Idaho Falls, ID (United States); Ross, W. [Pacific Northwest Lab., Richland, WA (United States); Nakaoka, R. [Los Alamos National Lab., NM (United States); Schumacher, R. [Westinghouse Savannah River Co., Aiken, SC (United States); Cunnane, J.; Singh, D. [Argonne National Lab., IL (United States); Darnell, R. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Greenhalgh, W. [Westinghouse Hanford Co., Richland, WA (United States)

1993-08-01T23:59:59.000Z

164

Spanish high level radioactive waste management system issues  

SciTech Connect (OSTI)

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.

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

1993-12-31T23:59:59.000Z

165

Corrosion and failure processes in high-level waste tanks  

SciTech Connect (OSTI)

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.

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

1992-11-01T23:59:59.000Z

166

Hanford low-level tank waste interim performance assessment  

SciTech Connect (OSTI)

The Hanford Low-Level Tank Waste Interim Performance Assessment examines the long-term environmental and human health effects associated with the disposal of the low-level fraction of the Hanford single and double-shell tank waste in the Hanford Site 200 East Area. This report was prepared as a good management practice to provide needed information about the relationship between the disposal system design and performance early in the disposal system project cycle. The calculations in this performance assessment show that the disposal of the low-level fraction can meet environmental and health performance objectives.

Mann, F.M.

1997-09-12T23:59:59.000Z

167

Hanford low-level tank waste interim performance assessment  

SciTech Connect (OSTI)

The Hanford Low-Level Tank Waste Interim Performance Assessment examines the long-term environmental and human health effects associated with the disposal of the low-level fraction of the Hanford single- and double-shell tank waste in the Hanford Site 200 East Area. This report was prepared as a good management practice to provide needed information about the relationship between the disposal system design and its performance as early as possible in the project cycle. The calculations in this performance assessment show that the disposal of the low-level fraction can meet environmental and health performance objectives.

Mann, F.M.

1996-09-16T23:59:59.000Z

168

Flammable gas tank waste level reconciliation for 241-S-111  

SciTech Connect (OSTI)

Fluor Daniel Northwest (FDNW) was authorized to address flammable gas issues by reconciling the unexplained surface level increases in Tank 241-S-111. The trapped gas evaluation document states that Tank S-111 exceeds the 25% of the lower flammable-limit criterion, based on a surface level rise evaluation. The Waste Storage Tank Status and Leak Detection Criteria document, commonly referred to as the Welty Report is the basis for this letter report. The unexplained waste level rises were attributed to the production and retention of gas in the column of waste corresponding to the unaccounted for surface level rise. From 1973 through 1980, the Welty Report tracked Tank S-111 transfers. This surface level increase is from an unknown source or is unaccounted for. Duke Engineering and Services Hanford and Lockheed Martin Hanford Corporation are interested in determining the validity of the unexplained surface level changes reported in the Welty Report based upon other corroborative sources of data. The purpose of this letter report is to assemble detailed surface level and waste addition data from daily tank records, logbooks, and other corroborative data that indicate surface levels, and to reconcile the cumulative unaccounted for surface level changes as shown in the Welty Report from 1973 through 1980. Tank S-111 initially received waste from REDOX in 1952, and after April 1974, primarily received processed waste slurry from the 242-S Evaporator/Crystallizer and transferred supernatant waste to Tank S-102. From the FDNW review and comparisons of the Welty Report versus other daily records for Tank S-111, FDNW determined that the majority of the time, the Welty Report is consistent with daily records. Surface level decreases that occurred following saltwell pumping were identified as unaccounted for decreases in the Welty Report, however they were probably a continued settlement caused by saltwell pumping of the interstitial liquids. Because the flammable/trapped gas issue is linked to the unexplained increase in the surface level, FDNW recommends that all occurrence reports, concerning tank waste level increases or decreases from 1970 through 1980, be reevaluated for acceptability of the evaluation as to the root cause of the occurrence.

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

1997-06-23T23:59:59.000Z

169

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

SciTech Connect (OSTI)

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.

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

1987-10-01T23:59:59.000Z

170

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

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

Greater-than-Class C Low-Level Radioactive Waste (GTCC LLW) Greater-than-Class C Low-Level Radioactive Waste (GTCC LLW) A transuranic (TRU) waste shipment makes its way to the...

171

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

SciTech Connect (OSTI)

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

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

1980-10-01T23:59:59.000Z

172

ENVIROCARE OF UTAH: EXPANDING WASTE ACCEPTANCE CRITERIA TO PROVIDE LOW-LEVEL AND MIXED WASTE DISPOSAL OPTIONS  

SciTech Connect (OSTI)

Envirocare of Utah operates a low-level radioactive waste disposal facility 80 miles west of Salt Lake City in Clive, Utah. Accepted waste types includes NORM, 11e2 byproduct material, Class A low-level waste, and mixed waste. Since 1988, Envirocare has offered disposal options for environmental restoration waste for both government and commercial remediation projects. Annual waste receipts exceed 12 million cubic feet. The waste acceptance criteria (WAC) for the Envirocare facility have significantly expanded to accommodate the changing needs of restoration projects and waste generators since its inception, including acceptable physical waste forms, radiological acceptance criteria, RCRA requirements and treatment capabilities, PCB acceptance, and liquids acceptance. Additionally, there are many packaging, transportation, and waste management options for waste streams acceptable at Envirocare. Many subcontracting vehicles are also available to waste generators for both government and commercial activities.

Rogers, B.; Loveland, K.

2003-02-27T23:59:59.000Z

173

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

SciTech Connect (OSTI)

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

Not Available

1992-04-01T23:59:59.000Z

174

Cementitious waste option scoping study report  

SciTech Connect (OSTI)

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.

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

1998-02-01T23:59:59.000Z

175

Remote-Handled Low Level Waste Disposal Project Alternatives Analysis  

SciTech Connect (OSTI)

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

David Duncan

2010-10-01T23:59:59.000Z

176

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...

177

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

SciTech Connect (OSTI)

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.

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

2012-09-30T23:59:59.000Z

178

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

SciTech Connect (OSTI)

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.

WILLIS, W.L.

2000-06-15T23:59:59.000Z

179

Overview of resuspension model: application to low level waste management  

SciTech Connect (OSTI)

Resuspension is one of the potential pathways to man for radioactive or chemical contaminants that are in the biosphere. In waste management, spills or other surface contamination can serve as a source for resuspension during the operational phase. After the low-level waste disposal area is closed, radioactive materials can be brought to the surface by animals or insects or, in the long term, the surface can be removed by erosion. Any of these methods expose the material to resuspension in the atmosphere. Intrusion into the waste mass can produce resuspension of potential hazard to the intruder. Removal of items from the waste mass by scavengers or archeologists can result in potential resuspension exposure to others handling or working with the object. The ways in which resuspension can occur are wind resuspension, mechanical resuspension and local resuspension. While methods of predicting exposure are not accurate, they include the use of the resuspension factor, the resuspension rate and mass loading of the air.

Healy, J.W.

1980-01-01T23:59:59.000Z

180

Developments in Very Low Level Waste/Exempt Waste Assay at AWE - 12000  

SciTech Connect (OSTI)

Portable High Resolution Gamma Spectrometry (HRGS) has been developed, for Very Low Level Waste (VLLW) and Exempt Waste (EW) assay at AWE, in order to meet the latest reduced clearance levels of < 1 Bq/g (or Bq/cm{sup 2}) for uranium (U) contaminated wastes and < 0.15 Bq/g (or Bq/cm{sup 2}) for plutonium (Pu) wastes. Studies have focused on a 10 kg bag of low density soft waste monitored either as a rotating cylinder, contained within a shortened plastic drum liner, or as a contained disk monitored on each broad side. Liquid and surface contaminated metal wastes have also been studied. It was established that monitoring the disk gave the best detection levels, but uncertainties rose more sharply, compared to the cylinder, as detector offset was reduced. Exempt detection levels were readily achieved for all U compositions encountered at AWE and for most Pu compositions (via Am-241 measurement). However, performance will need to be enhanced for those Pu compositions with relatively high Pu/Am-241 activity ratios. Recommendations are made for further developments to enhance the performance of this technique so that exempt clearance can be achieved for all Pu compositions encountered. (author)

Miller, T.J. [AWE, Aldermaston, Reading, Berkshire, RG7 4PR (United Kingdom)

2012-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

Permitting plan for the immobilized low-activity waste project  

SciTech Connect (OSTI)

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.

Deffenbaugh, M.L.

1997-09-04T23:59:59.000Z

182

Steam Reforming of Low-Level Mixed Waste  

SciTech Connect (OSTI)

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

None

1998-01-01T23:59:59.000Z

183

International low level waste disposal practices and facilities  

SciTech Connect (OSTI)

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

Nutt, W.M. (Nuclear Engineering Division)

2011-12-19T23:59:59.000Z

184

DOE high-level waste tank safety program. Final report  

SciTech Connect (OSTI)

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.

NONE

1998-11-01T23:59:59.000Z

185

Mixed and low-level waste treatment facility project  

SciTech Connect (OSTI)

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

Not Available

1992-04-01T23:59:59.000Z

186

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

SciTech Connect (OSTI)

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.

Kelly, Steven E.

2013-11-11T23:59:59.000Z

187

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...

188

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...

189

Proceedings: 2001 EPRI International Low-Level Waste Conference  

SciTech Connect (OSTI)

Nuclear utilities are continually evaluating methods to improve operations and minimize cost. EPRI's tenth annual International Low Level Waste (LLW) Conference--coupled with the 22nd annual ASME/EPRI Radwaste Workshop--offered valuable insights into this effort by presenting papers covering new or improved technology developed worldwide for LLW management, processing, shipment, disposal, and regulation.

None

2001-12-01T23:59:59.000Z

190

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

SciTech Connect (OSTI)

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.

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

191

Waste minimization for commercial radioactive materials users generating low-level radioactive waste  

SciTech Connect (OSTI)

The objective of this document is to provide a resource for all states and compact regions interested in promoting the minimization of low-level radioactive waste (LLW). This project was initiated by the Commonwealth of Massachusetts, and Massachusetts waste streams have been used as examples; however, the methods of analysis presented here are applicable to similar waste streams generated elsewhere. This document is a guide for states/compact regions to use in developing a system to evaluate and prioritize various waste minimization techniques in order to encourage individual radioactive materials users (LLW generators) to consider these techniques in their own independent evaluations. This review discusses the application of specific waste minimization techniques to waste streams characteristic of three categories of radioactive materials users: (1) industrial operations using radioactive materials in the manufacture of commercial products, (2) health care institutions, including hospitals and clinics, and (3) educational and research institutions. Massachusetts waste stream characterization data from key radioactive materials users in each category are used to illustrate the applicability of various minimization techniques. The utility group is not included because extensive information specific to this category of LLW generators is available in the literature.

Fischer, D.K.; Gitt, M.; Williams, G.A.; Branch, S. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Otis, M.D.; McKenzie-Carter, M.A.; Schurman, D.L. (Science Applications International Corp., Idaho Falls, ID (United States))

1991-07-01T23:59:59.000Z

192

Waste minimization for commercial radioactive materials users generating low-level radioactive waste. Revision 1  

SciTech Connect (OSTI)

The objective of this document is to provide a resource for all states and compact regions interested in promoting the minimization of low-level radioactive waste (LLW). This project was initiated by the Commonwealth of Massachusetts, and Massachusetts waste streams have been used as examples; however, the methods of analysis presented here are applicable to similar waste streams generated elsewhere. This document is a guide for states/compact regions to use in developing a system to evaluate and prioritize various waste minimization techniques in order to encourage individual radioactive materials users (LLW generators) to consider these techniques in their own independent evaluations. This review discusses the application of specific waste minimization techniques to waste streams characteristic of three categories of radioactive materials users: (1) industrial operations using radioactive materials in the manufacture of commercial products, (2) health care institutions, including hospitals and clinics, and (3) educational and research institutions. Massachusetts waste stream characterization data from key radioactive materials users in each category are used to illustrate the applicability of various minimization techniques. The utility group is not included because extensive information specific to this category of LLW generators is available in the literature.

Fischer, D.K.; Gitt, M.; Williams, G.A.; Branch, S. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Otis, M.D.; McKenzie-Carter, M.A.; Schurman, D.L. [Science Applications International Corp., Idaho Falls, ID (United States)

1991-07-01T23:59:59.000Z

193

Mobile plant for low-level radioactive waste reprocessing  

SciTech Connect (OSTI)

Along with nuclear power plants, many scientific and industrial enterprises generate radioactive wastes, especially low-level liquid wastes. Some of these facilities generate only small amounts on the order of several dozen cubic meters per year. The Moscow scientific industrial association, Radon, developed a mobile pilot system, EKO, for the processing of LLW with a low salt content. The plant consists of three modules: ultrafiltration module; electrodialysis module; and filtration module. The paper describes the technical parameters and test results from the plant on real LLW.

Sobolev, I.A.; Panteleyev, V.I.; Demkin, V.I. [Government of Moscow (Russian Federation). Dept. of Engineering Supply

1993-12-31T23:59:59.000Z

194

Management of low-level radioactive wastes around the world  

SciTech Connect (OSTI)

This paper reviews the status of various practices used throughout the world for managing low-level radioactive wastes. Most of the information in this review was obtained through the DOE-sponsored International Program Support Office (IPSO) activities at Pacific Northwest Laboratory (PNL) at Richland, Washington. The objective of IPSO is to collect, evaluate, and disseminate information on international waste management and nuclear fuel cycle activities. The center's sources of information vary widely and include the proceedings of international symposia, papers presented at technical society meetings, published topical reports, foreign trip reports, and the news media. Periodically, the information is published in topical reports. Much of the information contained in this report was presented at the Fifth Annual Participants' Information Meeting sponsored by DOE's Low-Level Waste Management Program Office at Denver, Colorado, in September of 1983. Subsequent to that presentation, the information has been updated, particularly with information provided by Dr. P. Colombo of Brookhaven National Laboratory who corresponded with low-level waste management specialists in many countries. The practices reviewed in this paper generally represent actual operations. However, major R and D activities, along with future plans, are also discussed. 98 refs., 6 tabls.

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

1985-04-01T23:59:59.000Z

195

Steam reforming of low-level mixed waste  

SciTech Connect (OSTI)

The U.S. department of Energy (DOE) is responsible for the treatment and disposal of an inventory of approximately 160,000 tons of Low-Level Mixed Waste (LLMW). Most of this LLMW is stored in drums, barrels and steel boxes at 20 different sites throughout the DOE complex. The basic objective of low-level mixed waste treatment systems is to completely destroy the hazardous constituents and to simultaneously isolate and capture the radionuclides in a superior final waste form such as glass. The DOE is sponsoring the development of advanced technologies that meet this objective while achieving maximum volume reduction, low-life cycle costs and maximum operational safety. ThermoChem, Inc. is in the final stages of development of a steam-reforming system capable of treating a wide variety of DOE low-level mixed waste that meets these objectives. The design, construction, and testing of a nominal 1 ton/day Process Development Unit is described.

Voelker, G.E.; Steedman, W.G. [Thermochem, Inc., Columbia, MD (United States); Chandran, R.R. [Manufacturing and Technology Conversion International, Inc., Columbia, MD (United States)

1996-12-31T23:59:59.000Z

196

Screening Level Risk Assessment for the New Waste Calcining Facility  

SciTech Connect (OSTI)

This screening level risk assessment evaluates potential adverse human health and ecological impacts resulting from continued operations of the calciner at the New Waste Calcining Facility (NWCF) at the Idaho Nuclear Technology and Engineering Center (INTEC), Idaho National Engineering and Environmental Laboratory (INEEL). The assessment was conducted in accordance with the Environmental Protection Agency (EPA) report, Guidance for Performing Screening Level Risk Analyses at Combustion Facilities Burning Hazardous Waste. This screening guidance is intended to give a conservative estimate of the potential risks to determine whether a more refined assessment is warranted. The NWCF uses a fluidized-bed combustor to solidify (calcine) liquid radioactive mixed waste from the INTEC Tank Farm facility. Calciner off volatilized metal species, trace organic compounds, and low-levels of radionuclides. Conservative stack emission rates were calculated based on maximum waste solution feed samples, conservative assumptions for off gas partitioning of metals and organics, stack gas sampling for mercury, and conservative measurements of contaminant removal (decontamination factors) in the off gas treatment system. Stack emissions were modeled using the ISC3 air dispersion model to predict maximum particulate and vapor air concentrations and ground deposition rates. Results demonstrate that NWCF emissions calculated from best-available process knowledge would result in maximum onsite and offsite health and ecological impacts that are less then EPA-established criteria for operation of a combustion facility.

M. L. Abbott; K. N. Keck; R. E. Schindler; R. L. VanHorn; N. L. Hampton; M. B. Heiser

1999-05-01T23:59:59.000Z

197

Secondary Waste Cast Stone Waste Form Qualification Testing Plan  

SciTech Connect (OSTI)

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

Westsik, Joseph H.; Serne, R. Jeffrey

2012-09-26T23:59:59.000Z

198

WRAP low level waste (LLW) glovebox operational test report  

SciTech Connect (OSTI)

The Low Level Waste (LLW) Process Gloveboxes are designed to: receive a 55 gallon drum in an 85 gallon overpack in the Entry glovebox (GBIOI); and open and sort the waste from the 55 gallon drum, place the waste back into drum and relid in the Sorting glovebox (GB 102). In addition, waste which requires further examination is transferred to the LLW RWM Glovebox via the Drath and Schraeder Bagiess Transfer Port (DO-07-201) or sent to the Sample Transfer Port (STC); crush the drum in the Supercompactor glovebox (GB 104); place the resulting puck (along with other pucks) into another 85 gallon overpack in the Exit glovebox (GB 105). The status of the waste items is tracked by the Data Management System (DMS) via the Plant Control System (PCS) barcode interface. As an item is moved from the entry glovebox to the exit glovebox, the Operator will track an items location using a barcode reader and enter any required data on the DMS console. The Operational Test Procedure (OTP) will perform evolution`s (described below) using the Plant Operating Procedures (POP) in order to verify that they are sufficient and accurate for controlled glovebox operation.

Kersten, J.K.

1998-02-19T23:59:59.000Z

199

Mixed low-level waste minimization at Los Alamos  

SciTech Connect (OSTI)

During the first six months of University of California 98 Fiscal Year (July--December) Los Alamos National Laboratory has achieved a 57% reduction in mixed low-level waste generation. This has been accomplished through a systems approach that identified and minimized the largest MLLW streams. These included surface-contaminated lead, lead-lined gloveboxes, printed circuit boards, and activated fluorescent lamps. Specific waste minimization projects have been initiated to address these streams. In addition, several chemical processing equipment upgrades are being implemented. Use of contaminated lead is planned for several high energy proton beam stop applications and stainless steel encapsulated lead is being evaluated for other radiological control area applications. INEEL is assisting Los Alamos with a complete systems analysis of analytical chemistry derived mixed wastes at the CMR building and with a minimum life-cycle cost standard glovebox design. Funding for waste minimization upgrades has come from several sources: generator programs, waste management, the generator set-aside program, and Defense Programs funding to INEEL.

Starke, T.P.

1998-12-01T23:59:59.000Z

200

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

SciTech Connect (OSTI)

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.

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

1994-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

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

SciTech Connect (OSTI)

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

Not Available

1992-04-01T23:59:59.000Z

202

Steam reforming of low-level mixed waste. Final report  

SciTech Connect (OSTI)

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

NONE

1998-06-01T23:59:59.000Z

203

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

E-Print Network [OSTI]

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...

Dontha, Lakshman

2012-07-16T23:59:59.000Z

204

Microbial degradation of low-level radioactive waste. Final report  

SciTech Connect (OSTI)

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

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

1996-06-01T23:59:59.000Z

205

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

SciTech Connect (OSTI)

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.

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

206

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

SciTech Connect (OSTI)

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)

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

207

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

SciTech Connect (OSTI)

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.

Swanson, J.L.

1993-09-01T23:59:59.000Z

208

Maximum surface level and temperature histories for Hanford waste tanks  

SciTech Connect (OSTI)

Radioactive defense waste resulting from the chemical processing of spent nuclear fuel has been accumulating at the Hanford Site since 1944. This waste is stored in underground waste-storage tanks. The Hanford Site Tank Farm Facilities Interim Safety Basis (ISB) provides a ready reference to the safety envelope for applicable tank farm facilities and installations. During preparation of the ISB, tank structural integrity concerns were identified as a key element in defining the safety envelope. These concerns, along with several deficiencies in the technical bases associated with the structural integrity issues and the corresponding operational limits/controls specified for conduct of normal tank farm operations are documented in the ISB. Consequently, a plan was initiated to upgrade the safety envelope technical bases by conducting Accelerated Safety Analyses-Phase 1 (ASA-Phase 1) sensitivity studies and additional structural evaluations. The purpose of this report is to facilitate the ASA-Phase 1 studies and future analyses of the single-shell tanks (SSTs) and double-shell tanks (DSTs) by compiling a quantitative summary of some of the past operating conditions the tanks have experienced during their existence. This report documents the available summaries of recorded maximum surface levels and maximum waste temperatures and references other sources for more specific data.

Flanagan, B.D.; Ha, N.D.; Huisingh, J.S.

1994-09-02T23:59:59.000Z

209

Atlantic Interstate Low-Level Radioactive Waste Management Compact (South Carolina)  

Broader source: Energy.gov [DOE]

The Atlantic (Northeast) Interstate Low-Level Radioactive Waste Management Compact is a cooperative effort to plan, regulate, and administer the disposal of low-level radioactive waste in the...

210

Radioactive Waste Management Complex low-level waste radiological performance assessment  

SciTech Connect (OSTI)

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

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

1994-04-01T23:59:59.000Z

211

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)

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.

Lomenick, T.F.

1996-03-01T23:59:59.000Z

212

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

SciTech Connect (OSTI)

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.

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

213

Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility |  

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

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214

Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility  

Office of Environmental Management (EM)

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215

1989 Annual report on low-level radioactive waste management progress  

SciTech Connect (OSTI)

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

Not Available

1990-10-01T23:59:59.000Z

216

Commercial low-level radioactive waste transportation liability and radiological risk  

SciTech Connect (OSTI)

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

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

1992-08-01T23:59:59.000Z

217

Deep borehole disposal of high-level radioactive waste.  

SciTech Connect (OSTI)

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).

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

218

High-level waste tank farm set point document  

SciTech Connect (OSTI)

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.

Anthony, J.A. III

1995-01-15T23:59:59.000Z

219

Life-Cycle Cost Study for a Low-Level Radioactive Waste Disposal Facility in Texas  

SciTech Connect (OSTI)

This report documents the life-cycle cost estimates for a proposed low-level radioactive waste disposal facility near Sierra Blanca, Texas. The work was requested by the Texas Low-Level Radioactive Waste Disposal Authority and performed by the National Low-Level Waste Management Program with the assistance of Rogers and Associates Engineering Corporation.

B. C. Rogers; P. L. Walter (Rogers and Associates Engineering Corporation); R. D. Baird

1999-08-01T23:59:59.000Z

220

Use of the Waste-Incidental-to-Reprocessing Citation Process at the West Valley Demonstration Project - 12250  

SciTech Connect (OSTI)

The West Valley Demonstration Project recently achieved a breakthrough in management of radioactive waste from reprocessing of spent nuclear fuel by taking advantage of lessons learned at other Department of Energy (DOE) sites in implementation of the waste-incidental-to-reprocessing citation process of DOE Manual 435.1-1, Radioactive Waste Management. This breakthrough involved a revision to the site procedure on waste-incidental to reprocessing. This procedure revision served as the basis for a determination by the DOE West Valley field office using the citation process that three secondary waste streams consisting of equipment that had once been contaminated by association with HLW are not HLW following decontamination and may be disposed of as low-level waste (LLW) or transuranic waste. These waste streams, which comprised much of the approximately 380 cubic meters of West Valley waste contaminated by association with HLW, included several vessels and certain tank farm equipment. By making use of lessons learned in use of the citation process by other DOE sites and information developed to support use of the citation process at the Hanford site and the Savannah River Site, the team developed a technical basis for showing that use of the citation process of DOE Manual 435.1-1 for the three new waste stream was appropriate and technically justified. The Waste Management Working Group of the EFCOG assisted in transferring lessons learned by drawing on experience from around the DOE complex. This process shared knowledge about effective implementation of the citation process in a manner that proved to be beneficial to the West Valley Demonstration Project and resulted in a technical basis document that could be used to determine that the three new waste streams were not HLW. (authors)

Sullivan, Dan; Suttora, Linda [U.S. Department of Energy (United States); Goldston, Sonny [EnergySolutions - USA (United States); Petras, Robert [Savannah River Remediation (United States); Rowell, Laurene [URS Corporation (United States); McNeil, Jim [Consultant - USA (United States)

2012-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

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

SciTech Connect (OSTI)

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.

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

2012-01-10T23:59:59.000Z

222

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

SciTech Connect (OSTI)

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

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

2011-12-29T23:59:59.000Z

223

Southeast Interstate Low-Level Radioactive Waste Management Compact (multi-state)  

Broader source: Energy.gov [DOE]

The Southeast Interstate Low-Level Radioactive Waste Management Compact is administered by the Compact Commission. The Compact provides for rotating responsibility for the region's low-level...

224

Defense High-Level Waste Leaching Mechanisms Program. Final report  

SciTech Connect (OSTI)

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.

Mendel, J.E. (compiler)

1984-08-01T23:59:59.000Z

225

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

SciTech Connect (OSTI)

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)

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

226

Proceedings: 2002 EPRI International Low Level Waste Conference  

SciTech Connect (OSTI)

Nuclear utilities are continually evaluating methods to improve operations and minimize cost. EPRI's 11th annual International Low Level Waste (LLW) Conference--coupled with the 25th annual Radwaste Workshop cosponsored by the American Society of Mechanical Engineers (ASME) and EPRI--offered valuable insights into this effort. Industry representatives presented papers covering new or improved technology developed worldwide for LLW management, processing, shipment, disposal, and regulation. This year, in collaboration with the International Atomic Energy Agency (IAEA), foreign participation increased, with papers from Canada, Korea, Germany, Finland, Ukraine, Belgium, the Slovak Republic, and the United Kingdom expanding the conference scope.

None

2002-09-01T23:59:59.000Z

227

Proceedings: 2003 EPRI International Low Level Waste Conference  

SciTech Connect (OSTI)

Nuclear utilities are continually evaluating methods to improve operations and minimize cost. EPRI's Twelfth Annual International Low Level Waste (LLW) Conference--coupled with the 24th Annual ASME/EPRI Radwaste Workshop--offered valuable insights into this effort by presenting papers covering new or improved technology developed worldwide for LLW management, processing, shipment, disposal, and regulation. EPRI accomplished the conference planning in collaboration with the International Atomic Energy Agency (IAEA). In addition to the United States, international representatives from the IAEA, Korea, Hungary, Canada, the United Kingdom, Japan, and Germany presented papers.

None

2004-04-01T23:59:59.000Z

228

Low-Level Waste Overview of the Nevada Test Site  

SciTech Connect (OSTI)

This paper provides an overview and the impacts of new policies, processes, and opportunities at the Nevada Test Site. Operational changes have been implemented, such as larger trench sizes and more efficient soil management as have administrative processes to address U.S. Department of Energy and U.S. Code of Federal Regulation analyses. Some adverse conditions have prompted changes in transportation and mixed low-level waste polices, and a new funding mechanism was developed. This year has seen many changes to the Nevada Test Site disposal family.

J. T. Carilli; M. G. Skougard; S. K. Krenzien; J.K Wrapp; C. Ramirez; V. Yucel; G.J. Shott; S.J. Gordon; K.C. Enockson; L.T. Desotell

2008-02-01T23:59:59.000Z

229

High-Level Waste Corporate Board Performance Assessment Subcommittee  

Office of Environmental Management (EM)

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230

Low-Level Waste Disposal Facility Federal Review Group Manual  

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 | DepartmentEnergy InvitationLegaltoLizLow- LEVEL WASTE DISPOSAL FACILITY

231

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

SciTech Connect (OSTI)

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.

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

232

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

SciTech Connect (OSTI)

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.

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

2013-11-13T23:59:59.000Z

233

THERMAL ANALYSIS OF WASTE GLASS MELTER FEEDS  

SciTech Connect (OSTI)

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.

KRUGER AA; HRMA PR; POKORNY R; PIERCE DA

2011-10-21T23:59:59.000Z

234

Direct cementitious waste option study report  

SciTech Connect (OSTI)

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.

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

1998-02-01T23:59:59.000Z

235

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

SciTech Connect (OSTI)

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

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

1994-09-01T23:59:59.000Z

236

JET MIXING ANALYSIS FOR SRS HIGH-LEVEL WASTE RECOVERY  

SciTech Connect (OSTI)

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.

Lee, S.

2011-07-05T23:59:59.000Z

237

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

SciTech Connect (OSTI)

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.

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

2012-10-19T23:59:59.000Z

238

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

SciTech Connect (OSTI)

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

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

1990-09-18T23:59:59.000Z

239

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

SciTech Connect (OSTI)

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

Li, H.

1996-03-01T23:59:59.000Z

240

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

SciTech Connect (OSTI)

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

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

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

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

242

PLUTONIUM/HIGH-LEVEL VITRIFIED WASTE BDBE DOSE CALCULATION  

SciTech Connect (OSTI)

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.

D.C. Richardson

2003-03-19T23:59:59.000Z

243

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

SciTech Connect (OSTI)

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

Mohamed, Yasser T. [Hot Laboratories and Waste Management Center, Atomic Energy Authority, 3 Ahmed El-Zomor St., El-Zohour District, Naser City, 11787, Cairo (Egypt)] [Hot Laboratories and Waste Management Center, Atomic Energy Authority, 3 Ahmed El-Zomor St., El-Zohour District, Naser City, 11787, Cairo (Egypt)

2013-07-01T23:59:59.000Z

244

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

SciTech Connect (OSTI)

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.

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

1994-12-31T23:59:59.000Z

245

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

SciTech Connect (OSTI)

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.

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

1994-04-01T23:59:59.000Z

246

Engineered sorbent barriers for low-level waste disposal.  

SciTech Connect (OSTI)

The Engineered Sorbent Barriers Program at Pacific Northwest Laboratory is investigating sorbent materials to prevent the migration of soluble radio nuclides from low-level waste sites. These materials would allow water to pass, preventing the bathtub effect at humid sites. Laboratory studies identifield promising sorbent materials for three key radionuclides: for cesium, greensand; for cobalt, activated charcoal; and for strontium, synthetic zeolite or clinoptilolite. Mixtures of these sorbent materials were tested in 0.6-m-diameter columns using radioactive leachates. To simulate expected worst-case conditions, the leachate solution contained the radionuclides, competing cations, and a chelating agent and was adjusted to a pH of 5. A sorbent barrier comprised of greensand (1 wt%), activated charcoal (6 wt%), synthetic zeolite (20 wt%), and local soil (73 wt%) achieved the decontamination factors necessary to meet the regulatory performance requirements established for this study. Sorbent barriers can be applied to shallow-land burial, as backfill around the waste or engineered structures, or as backup to other liner systems. 7 refs., 14 figs., 12 tabs.

Freeman, H.D.; Mitchell, S.J.; Buelt, J.L.

1986-12-01T23:59:59.000Z

247

High level waste tank farm setpoint document. Revision 1  

SciTech Connect (OSTI)

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.

Anthony, J.A. III

1995-01-31T23:59:59.000Z

248

Selected radionuclides important to low-level radioactive waste management  

SciTech Connect (OSTI)

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

NONE

1996-11-01T23:59:59.000Z

249

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

SciTech Connect (OSTI)

This paper provides an overview and the impacts of new policies, processes, and opportunities at the Nevada Test Site (NTS). Operational changes have been implemented, such as larger trench sizes and more efficient soil management as have administrative processes to address U.S. Department of Energy and U.S. Code of Federal Regulation analyses. Some adverse conditions have prompted changes in transportation and mixed low-level waste polices, and a new funding mechanism was developed. This year has seen many changes to the NTS disposal family. (authors)

Carilli, J.T.; Skougard, M.G. [U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, Las Vegas, NV (United States); Krenzien, S.K. [Navarro Research and Engineering, Inc., Las Vegas, NV (United States); Wrapp, J.K.; Ramirez, C.; Yucel, V.; Shott, G.J.; Gordon, S.J.; Enockson, K.C.; Desotell, L.T. [National Security Technologies, LLC, Las Vegas, Nevada (United States)

2008-07-01T23:59:59.000Z

250

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

SciTech Connect (OSTI)

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

Adams, J.P.

1995-09-01T23:59:59.000Z

251

Transmutation of high-level radioactive waste - Perspectives  

E-Print Network [OSTI]

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.

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

2014-01-01T23:59:59.000Z

252

WRAP low level waste (LLW) glovebox acceptance test report  

SciTech Connect (OSTI)

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

Leist, K.J.

1998-02-17T23:59:59.000Z

253

High Level Waste System Impacts from Acid Dissolution of Sludge  

SciTech Connect (OSTI)

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.

KETUSKY, EDWARD

2006-04-20T23:59:59.000Z

254

Preliminary low-level waste feed staging plan  

SciTech Connect (OSTI)

A Preliminary Low-Level Waste Feed Staging Plan was prepared. The plan supports the Phase I privatization effort by providing recommendations that may influence the technical content of the final request for proposal, and the interface control documents for the turnover of two double-shell tanks (DST) to the private contractors for use as feed tanks and the transfer of supernate to these tanks. Additionally, the preliminary schedule of feed staging activities will be useful to both RL and the private bidders during the contract negotiation period. A revised feed staging plan will be issued in August 1996 reflecting anticipated changes in the request for proposal, resolution of issues identified in this report, and completion of additional work scope.

Certa, P.J.

1996-02-05T23:59:59.000Z

255

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

SciTech Connect (OSTI)

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.

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

1984-04-01T23:59:59.000Z

256

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."

257

Low-level waste certification plan for the Lawrence Berkeley Laboratory Hazardous Waste Handling Facility. Revision 1  

SciTech Connect (OSTI)

The purpose of this plan is to describe the organization and methodology for the certification of low-level radioactive waste (LLW) handled in the Hazardous Waste Handling Facility (HWHF) at Lawrence Berkeley Laboratory (LBL). This plan is composed to meet the requirements found in the Westinghouse Hanford Company (WHC) Solid Waste Acceptance Criteria (WAC) and follows the suggested outline provided by WHC in the letter of April 26, 1990, to Dr. R.H. Thomas, Occupational Health Division, LBL. LLW is to be transferred to the WHC Hanford Site Central Waste Complex and Burial Grounds in Hanford, Washington.

NONE

1995-01-10T23:59:59.000Z

258

Low-level Waste Forum meeting report. Winter meeting, January 26--28, 1994  

SciTech Connect (OSTI)

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

NONE

1994-12-31T23:59:59.000Z

259

Low-level Waste Forum meeting report. Summer meeting, July 21--23, 1993  

SciTech Connect (OSTI)

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

NONE

1993-12-31T23:59:59.000Z

260

Low-level Waste Forum meeting report. Quarterly meeting, July 25--26, 1991  

SciTech Connect (OSTI)

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

NONE

1991-12-31T23:59:59.000Z

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Low-level Waste Forum meeting report. Spring meeting, April 28--30, 1993  

SciTech Connect (OSTI)

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

NONE

1993-12-31T23:59:59.000Z

262

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

SciTech Connect (OSTI)

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

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

1996-03-01T23:59:59.000Z

263

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)

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.

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

264

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

SciTech Connect (OSTI)

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.

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

1988-08-01T23:59:59.000Z

265

A team effort: Reducing the volume of low-level radioactive waste  

SciTech Connect (OSTI)

This article describes the team effort at Entergy Operation`s River Bend Station in Louisiana to reduce the volume of low-level radioactive waste. Topic areas covered include the following: Assessment - waste composition analysis using EPRI guidelines; grassroots effort; release facility - managing the waste; emerging technologies; spreading the success. 4 fig.

Zimmermann, K.

1996-09-01T23:59:59.000Z

266

National Low-Level Waste Management Program Radionuclide Report Series, Volume 17: Plutonium-239  

SciTech Connect (OSTI)

This report, Volume 17 of the National Low-Level Waste Management Program Radionuclide Report Series, discusses the radiological and chemical characteristics of plutonium-239 (Pu-239). This report also discusses waste types and forms in which Pu-239 can be found, waste and disposal information on Pu-239, and Pu-239 behavior in the environment and in the human body.

J. P. Adams; M. L. Carboneau

1999-03-01T23:59:59.000Z

267

Nondestructive characterization of low-level transuranic waste  

SciTech Connect (OSTI)

The use of nondestructive evaluation (NDE) methods is proposed for characterization of transuranic (TRU) waste stored at the Radioactive Waste Management Complex. These NDE methods include real-time x-ray radiography, real-time neutron radiography, x-ray and neutron computed tomography, thermal imaging, container weighing, visual examination, and acoustic measurements. An integrated NDE system is proposed for characterization and certification of TRU waste destined for eventual shipment to the Waste Isolation Pilot Plant in New Mexico. Methods for automating both the classification waste and control of a complete nondestructive evaluation/nondestructive assay system are presented. Feasibility testing of the different NDE methods, including real-time x-ray radiography, and development of automated waste classification techniques are covered as part of a five year effort designed to yield a production waste characterization system.

Barna, B.A.; Reinhardt, W.W.

1981-10-01T23:59:59.000Z

268

Rules and Regulations for the Disposal of Low-Level Radioactive Waste (Nebraska)  

Broader source: Energy.gov [DOE]

These regulations, promulgated by the Department of Environmental Quality, contain provisions pertaining to the disposal of low-level radioactive waste, disposal facilities, and applicable fees.

269

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

SciTech Connect (OSTI)

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

NONE

1997-11-01T23:59:59.000Z

270

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)

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.

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

2011-12-29T23:59:59.000Z

271

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

SciTech Connect (OSTI)

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

O`Holleran, T.O.; Feng, X.; Kalb, P. [and others

1996-12-01T23:59:59.000Z

272

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

SciTech Connect (OSTI)

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.

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

2003-02-27T23:59:59.000Z

273

Cementitious Grout for Closing SRS High Level Waste Tanks - 12315  

SciTech Connect (OSTI)

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

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

274

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

SciTech Connect (OSTI)

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.

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

1993-12-31T23:59:59.000Z

275

Identification of permit and waste acceptance criteria provisions requiring modification for acceptance of commercial mixed waste. National Low-Level Waste Management Program  

SciTech Connect (OSTI)

In October 1990, representatives of States and compact regions requested that the US Department of Energy (DOE) explore an agreement with host States and compact regions under which DOE would accept commercial mixed low-level radioactive waste (LLW) at DOE`s own treatment and disposal facilities. A program for DOE management of commercial mixed waste is made potentially more attractive in light of the low commercial mixed waste volumes, high regulatory burdens, public opposition to new disposal sites, and relatively high cost of constructing commercial disposal facilities. Several studies were identified as essential in determining the feasibility of DOE accepting commercial mixed waste for disposal. The purpose of this report is to identify any current or proposed waste acceptance criteria (WAC) or Resource Conservation and Recovery Act (RCRA) provisions that would have to be modified for commercial mixed waste acceptance at specified DOE facilities. Following the introduction, Section 2 of this report (a) provides a background summary of existing and proposed mixed waste disposal facilities at each DOE site, and (b) summarizes the status of any RCRA Part B permit and WAC provisions relating to the disposal of mixed waste, including provisions relating to acceptance of offsite waste. Section 3 provides overall conclusions regarding the current status and permit modifications that must be implemented in order to grant DOE sites authority under their permits to accept commercial mixed waste for disposal. Section 4 contains a list of references.

Not Available

1994-03-01T23:59:59.000Z

276

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

SciTech Connect (OSTI)

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

Timothy Solack; Carol Mason

2012-03-01T23:59:59.000Z

277

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

SciTech Connect (OSTI)

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.

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

2013-11-13T23:59:59.000Z

278

PLUTONIUM SOLUBILITY IN HIGH-LEVEL WASTE ALKALI BOROSILICATE GLASS  

SciTech Connect (OSTI)

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.

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

2011-01-04T23:59:59.000Z

279

DATA SUMMARY REPORT SMALL SCALE MELTER TESTING OF HLW ALGORITHM GLASSES MATRIX1 TESTS VSL-07S1220-1 REV 0 7/25/07  

SciTech Connect (OSTI)

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.

KRUGER AA; MATLACK KS; PEGG IL

2011-12-29T23:59:59.000Z

280

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)

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.

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

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Determination of total cyanide in Hanford Site high-level wastes  

SciTech Connect (OSTI)

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.

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

282

Los Alamos low-level waste performance assessment status  

SciTech Connect (OSTI)

This report reviews the documented Los Alamos studies done to assess the containment of buried hazardous wastes. Five sections logically present the environmental studies, operational source terms, transport pathways, environmental dosimetry, and computer model development and use. This review gives a general picture of the Los Alamos solid waste disposal and liquid effluent sites and is intended for technical readers with waste management and environmental science backgrounds but without a detailed familiarization with Los Alamos. The review begins with a wide perspective on environmental studies at Los Alamos. Hydrology, geology, and meteorology are described for the site and region. The ongoing Laboratory-wide environmental surveillance and waste management environmental studies are presented. The next section describes the waste disposal sites and summarizes the current source terms for these sites. Hazardous chemical wastes and liquid effluents are also addressed by describing the sites and canyons that are impacted. The review then focuses on the transport pathways addressed mainly in reports by Healy and Formerly Utilized Sites Remedial Action Program. Once the source terms and potential transport pathways are described, the dose assessment methods are addressed. Three major studies, the waste alternatives, Hansen and Rogers, and the Pantex Environmental Impact Statement, contributed to the current Los Alamos dose assessment methodology. Finally, the current Los Alamos groundwater, surface water, and environmental assessment models for these mesa top and canyon sites are described.

Wenzel, W.J.; Purtymun, W.D.; Dewart, J.M.; Rodgers, J.E. (comps.)

1986-06-01T23:59:59.000Z

283

Hanford high-level waste evaporator/crystallizer corrosion evaluation  

SciTech Connect (OSTI)

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.

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

1993-10-01T23:59:59.000Z

284

Surrogate formulations for thermal treatment of low-level mixed waste, Part II: Selected mixed waste treatment project waste streams  

SciTech Connect (OSTI)

This report summarizes the formulation of surrogate waste packages, representing the major bulk constituent compositions for 12 waste stream classifications selected by the US DOE Mixed Waste Treatment Program. These waste groupings include: neutral aqueous wastes; aqueous halogenated organic liquids; ash; high organic content sludges; adsorbed aqueous and organic liquids; cement sludges, ashes, and solids; chloride; sulfate, and nitrate salts; organic matrix solids; heterogeneous debris; bulk combustibles; lab packs; and lead shapes. Insofar as possible, formulation of surrogate waste packages are referenced to authentic wastes in inventory within the DOE; however, the surrogate waste packages are intended to represent generic treatability group compositions. The intent is to specify a nonradiological synthetic mixture, with a minimal number of readily available components, that can be used to represent the significant challenges anticipated for treatment of the specified waste class. Performance testing and evaluation with use of a consistent series of surrogate wastes will provide a means for the initial assessment (and intercomparability) of candidate treatment technology applicability and performance. Originally the surrogate wastes were intended for use with emerging thermal treatment systems, but use may be extended to select nonthermal systems as well.

Bostick, W.D.; Hoffmann, D.P.; Chiang, J.M.; Hermes, W.H.; Gibson, L.V. Jr.; Richmond, A.A. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States)] [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States); Mayberry, J. [Science Applications International Corp., Idaho Falls, ID (United States)] [Science Applications International Corp., Idaho Falls, ID (United States); Frazier, G. [Univ. of Tennessee, Knoxville, TN (United States)] [Univ. of Tennessee, Knoxville, TN (United States)

1994-01-01T23:59:59.000Z

285

Conversion of transuranic waste to low level waste by decontamination: a technical and economic evaluation  

SciTech Connect (OSTI)

A study was conducted to evaluate the technical and economic feasibility of using in-situ decontamination techniques to convert glove boxes and other large TRU-contaminated components directly into LLW. The results of the technical evaluation indicate that in-situ decontamination of these types of components to non-TRU levels is technically feasible. Applicable decontamination techniques include electropolishing, hand scrubbing, chemical washes/sprays, strippable coatings and Freon spray-cleaning. The removal of contamination from crevices and other holdup areas remains a problem, but may be solved through further advances in decontamination technology. Also, the increase in the allowable maximum TRU level from 10 nCi/g to 100 nCi/g as defined in DOE Order 5820.2 reduces the removal requirement and facilitates measurement of the remaining quantities. The major emphasis of the study was on a cost/benefit evaluation that included a review and update of previous analyses and evaluations of TRU-waste volume reduction and conversion options. The results of the economic evaluation show, for the assumptions used, that there is a definite cost incentive to size reduce large components, and that decontamination of sectioned material has become cost competitive with the size reduction options. In-situ decontamination appears to be the lowest cost option when based on routine-type operations conducted by well-trained and properly equipped personnel. 16 references, 1 figure, 7 tables.

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

1984-12-01T23:59:59.000Z

286

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

SciTech Connect (OSTI)

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.

Kruger, Albert A.

2013-01-16T23:59:59.000Z

287

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

SciTech Connect (OSTI)

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)

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

288

Report on waste burial charges. Escalation of decommissioning waste disposal costs at low-level waste burial facilities, Revision 4  

SciTech Connect (OSTI)

One of the requirements placed upon nuclear power reactor licensees by the U.S. Nuclear Regulatory Commission (NRC) is for the licensees to periodically adjust the estimate of the cost of decommissioning their plants, in dollars of the current year, as part of the process to provide reasonable assurance that adequate funds for decommissioning will be available when needed. This report, which is scheduled to be revised periodically, contains the development of a formula for escalating decommissioning cost estimates that is acceptable to the NRC. The sources of information to be used in the escalation formula are identified, and the values developed for the escalation of radioactive waste burial costs, by site and by year, are given. The licensees may use the formula, the coefficients, and the burial escalation factors from this report in their escalation analyses, or they may use an escalation rate at least equal to the escalation approach presented herein. This fourth revision of NUREG-1307 contains revised spreadsheet results for the disposal costs for the reference PWR and the reference BWR and the ratios of disposal costs at the Washington, Nevada, and South Carolina sites for the years 1986, 1988, 1991 and 1993, superseding the values given in the May 1993 issue of this report. Burial cost surcharges mandated by the Low-Level Radioactive Waste Policy Amendments Act of 1985 (LLRWPAA) have been incorporated into the revised ratio tables for those years. In addition, spreadsheet results for the disposal costs for the reference reactors and ratios of disposal costs at the two remaining burial sites in Washington and South Carolina for the year 1994 are provided. These latter results do not include any LLRWPAA surcharges, since those provisions of the Act expired at the end of 1992. An example calculation for escalated disposal cost is presented, demonstrating the use of the data contained in this report.

Not Available

1994-06-01T23:59:59.000Z

289

Report on waste burial charges: Escalation of decommissioning waste disposal costs at Low-Level Waste Burial facilities. Revision 5  

SciTech Connect (OSTI)

One of the requirements placed upon nuclear power reactor licensees by the US Nuclear Regulatory Commission (NRC) is for the licensees to periodically adjust the estimate of the cost of decommissioning their plants, in dollars of the current year, as part of the process to provide reasonable assurance that adequate funds for decommissioning will be available when needed. This report, which is scheduled to be revised periodically, contains the development of a formula for escalating decommissioning cost estimates that is acceptable to the NRC. The sources of information to be used in the escalation formula are identified, and the values developed for the escalation of radioactive waste burial costs, by site and by year, are given. The licensees may use the formula, the coefficients, and the burial escalation factors from this report in their escalation analyses, or they may use an escalation rate at least equal to the escalation approach presented herein. This fifth revision of NUREG-1307 contains revised spreadsheet results for the disposal costs for the reference PWR and the reference BWR and the ratios of disposal costs at the Washington, Nevada, and South Carolina sites for the years 1986, 1988, 1991, 1993, and 1994, superseding the values given in the June 1994 issue of this report. Burial cost surcharges mandated by the Low-Level Radioactive Waste Policy Amendments Act of 1985 (LLRWPAA) have been incorporated into the revised ratio tables for those years. In addition, spreadsheet results for the disposal costs for the reference reactors and ratios of disposal costs at the two remaining burial sites in Washington and South Carolina for the year 1995 are provided. These latter results do not include any LLRWPAA surcharges, since those provisions of the Act expired at the end of 1992. An example calculation for escalated disposal cost is presented, demonstrating the use of the data contained in this report.

NONE

1995-08-01T23:59:59.000Z

290

High level waste facilities -- Continuing operation or orderly shutdown  

SciTech Connect (OSTI)

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.

Decker, L.A.

1998-04-01T23:59:59.000Z

291

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

SciTech Connect (OSTI)

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

Tyacke, M.

1993-08-01T23:59:59.000Z

292

Solid low-level radioactive waste radiation stability studies  

E-Print Network [OSTI]

properties and condition; on the other, on the specific features of thermal and radiation influences on it (Spitsyn et al. 1983). For the average composition of the fission products going to wastes repositories, the mean energy of irradiation may vary from... to the container determines, in part, the life of the container. Cormsion studies of containers by solidified wastes has indicated no problem areas in limited measurements to date; however very long-term effects have not been evaluated. The useful life...

Williams, Arnold Andre?

1989-01-01T23:59:59.000Z

293

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

SciTech Connect (OSTI)

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.

Meacham, J.E.

1996-10-03T23:59:59.000Z

294

Florida State Briefing Book for low-level radioactive-waste management  

SciTech Connect (OSTI)

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

none,

1981-06-01T23:59:59.000Z

295

Wyoming State Briefing Book for low-level radioactive waste management  

SciTech Connect (OSTI)

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

Not Available

1981-10-01T23:59:59.000Z

296

North Carolina State Briefing Book for low-level radioactive waste management  

SciTech Connect (OSTI)

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

Not Available

1981-08-01T23:59:59.000Z

297

Puerto Rico State Briefing Book for low-level radioactive waste management  

SciTech Connect (OSTI)

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

Not Available

1981-10-01T23:59:59.000Z

298

South Carolina State Briefing Book for low-level radioactive waste management  

SciTech Connect (OSTI)

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

Not Available

1981-08-01T23:59:59.000Z

299

Texas State Briefing Book for low-level radioactive waste management  

SciTech Connect (OSTI)

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

Not Available

1981-08-01T23:59:59.000Z

300

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

SciTech Connect (OSTI)

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

Not Available

1981-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Massachusetts State Briefing Book for low-level radioactive waste management  

SciTech Connect (OSTI)

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

Not Available

1981-03-12T23:59:59.000Z

302

North Dakota State Briefing Book for low-level radioactive waste management  

SciTech Connect (OSTI)

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

none,

1981-10-01T23:59:59.000Z

303

Utah State Briefing Book for low-level radioactive waste management  

SciTech Connect (OSTI)

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

Not Available

1981-10-01T23:59:59.000Z

304

Hanford Tank Waste - Near Source Treatment of Low Activity Waste  

SciTech Connect (OSTI)

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.

Ramsey, William Gene

2013-08-15T23:59:59.000Z

305

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

SciTech Connect (OSTI)

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.

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

1981-10-01T23:59:59.000Z

306

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

SciTech Connect (OSTI)

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.

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

2011-09-12T23:59:59.000Z

307

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

SciTech Connect (OSTI)

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.

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

308

Criticality Safety Evaluation of Hanford Site High Level Waste Storage Tanks  

SciTech Connect (OSTI)

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.

ROGERS, C.A.

2000-02-17T23:59:59.000Z

309

Vitrification as a low-level radioactive mixed waste treatment technology at Argonne National Laboratory  

SciTech Connect (OSTI)

Argonne National Laboratory-East (ANL-E) is developing plans to use vitrification to treat low-level radioactive mixed wastes (LLMW) generated onsite. The ultimate objective of this project is to install a full-scale vitrification system at ANL-E capable of processing the annual generation and historic stockpiles of selected LLMW streams. This project is currently in the process of identifying a range of processible glass compositions that can be produced from actual mixed wastes and additives, such as boric acid or borax. During the formulation of these glasses, there has been an emphasis on maximizing the waste content in the glass (70 to 90 wt %), reducing the overall final waste volume, and producing a stabilized low-level radioactive waste glass. Crucible glass studies with actual mixed waste streams have produced alkali borosilicate glasses that pass the Toxic Characteristic Leaching Procedure (TCLP) test. These same glass compositions, spiked with toxic metals well above the expected levels in actual wastes, also pass the TCLP test. These results provide compelling evidence that the vitrification system and the glass waste form will be robust enough to accommodate expected variations in the LLMW streams from ANL-E. Approximately 40 crucible melts will be studied to establish a compositional envelope for vitrifying ANL-E mixed wastes. Also being determined is the identity of volatilized metals or off-gases that will be generated.

Mazer, J.J.; No, Hyo J.

1995-08-01T23:59:59.000Z

310

Evaluation of interim and final waste forms for the newly generated liquid low-level waste flowsheet  

SciTech Connect (OSTI)

The purpose of this review is to evaluate the final forms that have been proposed for radioactive-containing solid wastes and to determine their application to the solid wastes that will result from the treatment of newly generated liquid low-level waste (NGLLLW) and Melton Valley Storage Tank (MVST) supernate at the Oak Ridge National Laboratory (ORNL). Since cesium and strontium are the predominant radionuclides in NGLLLW and MVST supernate, this review is focused on the stabilization and solidification of solid wastes containing these radionuclides in cement, glass, and polymeric materials-the principal waste forms that have been tested with these types of wastes. Several studies have shown that both cesium and strontium are leached by distilled water from solidified cement, although the leachabilities of cesium are generally higher than those of strontium under similar conditions. The situation is exacerbated by the presence of sulfates in the solution, as manifested by cracking of the grout. Additives such as bentonite, blast-furnace slag, fly ash, montmorillonite, pottery clay, silica, and zeolites generally decrease the cesium and strontium release rates. Longer cement curing times (>28 d) and high ionic strengths of the leachates, such as those that occur in seawater, also decrease the leach rates of these radionuclides. Lower cesium leach rates are observed from vitrified wastes than from grout waste forms. However, significant quantities of cesium are volatilized due to the elevated temperatures required to vitrify the waste. Hence, vitrification will generally require the use of cleanup systems for the off-gases to prevent their release into the atmosphere.

Abotsi, G.M.K. [Clark Atlanta Univ., GA (United States); Bostick, D.T.; Beck, D.E. [Oak Ridge National Lab., TN (United States)] [and others

1996-05-01T23:59:59.000Z

311

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)

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.

Not Available

1983-06-01T23:59:59.000Z

312

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

SciTech Connect (OSTI)

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.

Swita, W.R.

1998-01-09T23:59:59.000Z

313

National Low-Level Waste Management Program radionuclide report series. Volume 13, Curium-242  

SciTech Connect (OSTI)

This report, Volume 13 of the National Low-Level Waste Management Program Radionuclide Report Series, discusses the radiological and chemical characteristics of curium-242 ({sup 242}Cm). This report also includes discussions about waste types and forms in which {sup 242}Cm can be found and {sup 242}Cm behavior in the environment and in the human body.

Adams, J.P.

1995-08-01T23:59:59.000Z

314

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

E-Print Network [OSTI]

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

Boyer, Edmond

315

State-of-the-art report on low-level radioactive waste treatment  

SciTech Connect (OSTI)

An attempt is made to identify the main sources of low-level radioactive wastes that are generated in the United States. To place the waste problem in perspective, rough estimates are given of the annual amounts of each generic type of waste that is generated. Most of the wet solid wastes arise from the cleanup of gaseous and liquid radioactive streams prior to discharge or recycle. The treatment of the process streams and the secondary wet solid wastes thus generated is described for each type of government or fuel cycle installation. Similarly, the institutional wet wastes are also described. The dry wastes from all sources have smilar physical and chemical characteristics in that they can be classified as compactible, noncompactible, combustible, noncombustible, or combinations thereof. The various treatment options for concentrated or solid wet wastes and for dry wastes are discussed. Among the dry-waste treatment methods are compaction, baling, and incineration, as well as chopping, cutting, and shredding. Organic materials can usually be incinerated or, in some cases, biodegraded. The filter sludges, spent resins, incinerator ashes, and concentrated liquids are usually solidified in cement, urea-formaldehyde, or unsaturated polyester resins prior to burial. Asphalt has not yet been used as a solidificaton agent in the United States, but it probably will be used in the near future. The treatment of radioactive medical and bioresearch wastes is described, but the waste from radiochenmical, pharmaceutical, and other industries is not well defined at the present time. Recovery of waste metals and treatment of hazardous contaminated wastes are discussed briefly. Some areas appearing to need more research, development, and demonstration are specifically pointed out.

Kibbey, A.H.; Godbee, H.W.

1980-09-01T23:59:59.000Z

316

Determination of Iodine-129 in Low Level Radioactive Wastes - 13334  

SciTech Connect (OSTI)

For the radioactivity determination of {sup 129}I in the radioactive wastes, alkali fusion and anion-exchange resin separation methods, which are sample pretreatment methods, have been investigated in this study. To separate and quantify the {sup 129}I radionuclide in an evaporator bottom and spent resin, the radionuclide was chemically leached from the wastes and adsorbed on an anion exchange resin at pH 4, 7, 9. In the case of dry active waste and another solid type, the alkali fusion method was applied. KNO{sub 3} was added as a KOH and oxidizer to the wastes. It was then fused at 450 deg. C for 1 hour. The radioactivity of the separated iodine was measured with a low energy gamma spectrometer after the sample pretreatment. Finally, it was confirmed that the recovery rate of the iodine for the alkali fusion method was 83.6±3.8%, and 86.4±1.6% for the anionic exchange separation method. (authors)

Choi, K.C.; Ahn, J.H.; Park, Y.J.; Song, K.S. [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daejeon, 305-600 (Korea, Republic of)] [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daejeon, 305-600 (Korea, Republic of)

2013-07-01T23:59:59.000Z

317

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

SciTech Connect (OSTI)

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.

R.A. Levich; J.S. Stuckless

2006-09-25T23:59:59.000Z

318

Progress in Low and Intermediate Level Operational Waste Characterization and Preparation for Disposal at Ignalina NPP  

SciTech Connect (OSTI)

In Lithuania about 70-80% of all electricity is generated at a single power station, Ignalina NPP, which has two RBMK-1500 type reactors. Units 1 and 2 will be closed by 2005 and 2010, respectively, taking into account the conditions of the long-term substantial financial assistance rendered by the European Union, G-7 countries and other states as well as international institutions. The Government approved the Strategy on Radioactive Waste Management. Objectives of this strategy are to develop the radioactive waste management infrastructure based on modern technologies and provide for the set of practical actions that shall bring management of radioactive waste in Lithuania in compliance with radioactive waste management principles of IAEA and with good practices in force in European Union Member States. SKB-SWECO International-Westinghouse Atom Joint Venture with participation of Lithuanian Energy Institute has prepared a reference design of a near surface repository for short-lived low and intermediate level waste. This reference design is applicable to the needs in Lithuania, considering its hydro-geological, climatic and other environmental conditions and is able to cover the expected needs in Lithuania for at least thirty years ahead. Development of waste acceptance criteria is in practice an iterative process concerning characterization of existing waste, repository development, safety and environmental impact assessment etc. This paper describes the position in Lithuania with regard to the long-term management of low and intermediate level waste in the absence of finalized waste acceptance criteria and a near surface repository.

Poskas, P.; Adomaitis, J. E.; Ragaisis, V.

2003-02-25T23:59:59.000Z

319

An experimental survey of the factors that affect leaching from low-level radioactive waste forms  

SciTech Connect (OSTI)

This report represents the results of an experimental survey of the factors that affect leaching from several types of solidified low-level radioactive waste forms. The goal of these investigations was to determine those factors that accelerate leaching without changing its mechanism(s). Typically, although not in every case,the accelerating factors include: increased temperature, increased waste loading (i.e., increased waste to binder ratio), and decreased size (i.e., decreased waste form volume to surface area ratio). Additional factors that were studied were: increased leachant volume to waste form surface area ratio, pH, leachant composition (groundwaters, natural and synthetic chelating agents), leachant flow rate or replacement frequency and waste form porosity and surface condition. Other potential factors, including the radiation environment and pressure, were omitted based on a survey of the literature. 82 refs., 236 figs., 13 tabs.

Dougherty, D.R.; Pietrzak, R.F.; Fuhrmann, M.; Colombo, P.

1988-09-01T23:59:59.000Z

320

Phosphate ceramic process for macroencapsulation and stabilization of low-level debris wastes  

SciTech Connect (OSTI)

Across the DOE complex, large quantities of contaminated debris and irradiated lead bricks have been accumulated for disposal. Under the US Environmental Protection Agency`s Alternative Treatment Standards, the preferred method of disposal of these wastes is macroencapsulation. Chemically bonded phosphate ceramic (CBPC) is a novel binder that was developed at Argonne National Laboratory to stabilize and solidify various low-level mixed wastes. This binder is extremely strong, dense, and impervious to water. In this investigation, CBPC has been used to demonstrate macroencapsulation of various contaminated debris wastes, including cryofractured debris, lead bricks, lead-lined plastic gloves, and mercury-contaminated crushed glass. This paper describes the fabrication of the waste forms, as well as the results of various characterizations performed on the waste forms. The results show that the simple and low-cost CBPC is an excellent material system for the macroencapsulation of debris wastes.

Singh, D.; Wagh, A.S.; Tlustochowicz, M.; Jeong, S.Y. [Argonne National Lab., IL (United States). Energy Technology Div.] [Argonne National Lab., IL (United States). Energy Technology Div.

1998-12-31T23:59:59.000Z

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Macroencapsulation of low-level debris waste with the phosphate ceramic process  

SciTech Connect (OSTI)

Across the DOE complex, large quantities of contaminated debris and irradiated lead bricks require disposal. The preferred method for disposing of these wastes is macroencapsulation under U.S. Environmental Protection Agency Alternative Treatment Standards. Chemically bonded phosphate ceramics serve as a novel binder, developed at Argonne National Laboratory, for stabilizing and solidifying various low-level mixed wastes. Extremely strong, dense, and impervious to water intrusion, this material was developed with support from the U.S. Department of Energy`s Office of Science and Technology (DOE OST). In this investigation, CBPCs have been used to demonstrate macroencapsulation of various contaminated debris wastes, including cryofractured debris, lead bricks, and lead-lined plastic gloves. This paper describes the processing steps for fabricating the waste forms and the results of various characterizations performed on the waste forms. The conclusion is that simple and low-cost CBPCs are excellent material systems for macroencapsulating debris wastes.

Singh, D.; Wagh, A.S.; Tlustochowicz, M.; Jeong, S.Y.

1997-03-01T23:59:59.000Z

322

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

SciTech Connect (OSTI)

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

NONE

1993-09-01T23:59:59.000Z

323

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

Broader source: Energy.gov [DOE]

This EIS assesses the environmental impacts of alternatives for the disposal of low-level waste and by-product materials generated by the three major plants on the Oak Ridge Reservation (ORR). In addition to the no-action alternative, two classes of alternatives are evaluated: facility design alternatives and siting alternatives.

324

Radiological, physical, and chemical characterization of low-level alpha contaminated wastes stored at the Idaho National Engineering Laboratory  

SciTech Connect (OSTI)

This document provides radiological, physical, and chemical characterization data for low-level alpha-contaminated radioactive and low-level alpha-contaminated radioactive and hazardous (i.e., mixed) wastes stored at the Idaho National Engineering Laboratory and considered for treatment under the Private Sector Participation Initiative Program. Waste characterization data are provided in the form of INEL Waste Profile Sheets. These documents provide, for each content code, information on waste identification, waste description, waste storage configuration, physical/chemical waste composition, radionuclide and associated alpha activity waste characterization data, and hazardous constituents present in the waste. Information is provided for 97 waste streams which represent an estimated total volume of 25,450 m 3 corresponding to a total mass of approximately 12,000,000 kg. In addition, considerable information concerning alpha, beta, gamma, and neutron source term data specific to Rocky Flats-generated waste forms stored at the INEL are provided to assist in facility design specification.

Apel, M.L.; Becker, G.K.; Ragan, Z.K.; Frasure, J.; Raivo, B.D.; Gale, L.G.; Pace, D.P.

1994-03-01T23:59:59.000Z

325

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

Broader source: Energy.gov [DOE]

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

326

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

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

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

327

Northwest Interstate Compact on Low-Level Radioactive Waste Management (Multiple States)  

Broader source: Energy.gov [DOE]

The Northwest Interstate Compact on Low-Level Radioactive Waste Management, enacted in 1981, was ratified by Congress in 1985. The Compact is a cooperative effort of the party states to protect...

328

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

E-Print Network [OSTI]

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. ...

Driscoll, Michael J.

1988-01-01T23:59:59.000Z

329

Midwest Interstate Compact on Low-Level Radioactive Waste (Multiple States)  

Broader source: Energy.gov [DOE]

The Midwest Interstate Low-Level Radioactive Waste Compact is an agreement between the states of Indiana, Iowa, Minnesota, Missouri, Ohio, and Wisconsin that provides for the cooperative and safe...

330

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

Broader source: Energy.gov [DOE]

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

331

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

SciTech Connect (OSTI)

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.

Peters, T.B.

2001-09-10T23:59:59.000Z

332

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

Broader source: Energy.gov [DOE]

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

333

Maine State Briefing Book on low-level radioactive waste management  

SciTech Connect (OSTI)

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

Not Available

1981-08-01T23:59:59.000Z

334

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

Broader source: Energy.gov [DOE]

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

335

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

E-Print Network [OSTI]

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 ...

Gibbs, Jonathan Sutton

2010-01-01T23:59:59.000Z

336

1994 annual report on low-level radioactive waste management progress  

SciTech Connect (OSTI)

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

NONE

1995-04-01T23:59:59.000Z

337

Managing low-level radioactive waste in a democratic society: Requirements and accommodations  

SciTech Connect (OSTI)

This paper will focus on public policy needs to ensure the involvement of the general public in effective decision-making related to the handling of low-level radioactive waste. It highlights difficulties experienced in involving the public in siting low-level radioactive waste disposal facilities. It reviews the process recently developed by Illinois to locate a disposal facility and discusses that process`s potential as a general model for siting such facilities and involving citizens in a democratic fashion.

Ortciger, T. [Illinois Dept. of Nuclear Safety, Springfield, IL (United States); Ayers, M. [Sangamon State Univ., Springfield, IL (United States)

1993-12-31T23:59:59.000Z

338

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

SciTech Connect (OSTI)

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.

Howden, G.F.

1994-10-24T23:59:59.000Z

339

Analysis of the low-level waste radionuclide inventory for the Radioactive Waste Management Complex performance assessment  

SciTech Connect (OSTI)

This report summarizes the results of a study to improve the estimates of the radionuclides in the low-level radioactive waste (LLW) inventory which is buried in the Idaho National Engineering Laboratory (INEL) Radioactive Waste Management Complex (RWMC) Subsurface Disposal Area (SDA). The work is done to support the RWMC draft performance assessment (PA). Improved radionuclide inventory estimates are provided for the INEL LLW generators. Engineering, environmental assessment or other research areas may find use for the information in this report. It may also serve as a LLW inventory baseline for data quality assurance. The individual INEL LLW generators, their history and their activities are also described in detail.

Plansky, L.E.; Hoiland, S.A.

1992-02-01T23:59:59.000Z

340

3-D MAPPING TECHNOLOGIES FOR HIGH LEVEL WASTE TANKS  

SciTech Connect (OSTI)

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.

Marzolf, A.; Folsom, M.

2010-08-31T23:59:59.000Z

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

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)

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.

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

2013-11-13T23:59:59.000Z

342

National profile on commercially generated low-level radioactive mixed waste  

SciTech Connect (OSTI)

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

Klein, J.A.; Mrochek, J.E.; Jolley, R.L.; Osborne-Lee, I.W.; Francis, A.A.; Wright, T. [Oak Ridge National Lab., TN (United States)

1992-12-01T23:59:59.000Z

343

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

SciTech Connect (OSTI)

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.

HILL RC PHILIP; REYNOLDS JG; RUTLAND PL

2011-02-23T23:59:59.000Z

344

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

SciTech Connect (OSTI)

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

G. Becker; M. Connolly; M. McIlwain

1999-02-01T23:59:59.000Z

345

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

SciTech Connect (OSTI)

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

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

1981-12-01T23:59:59.000Z

346

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

SciTech Connect (OSTI)

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

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

1995-01-01T23:59:59.000Z

347

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

SciTech Connect (OSTI)

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

Mayberry, J.L.; DeWitt, L.M. [Science Applications International Corp., Idaho Falls, ID (United States); Darnell, R. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

1993-08-01T23:59:59.000Z

348

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

SciTech Connect (OSTI)

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.

Jolly, R; Bruce Martin, B

2008-01-15T23:59:59.000Z

349

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)

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.

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

2011-12-29T23:59:59.000Z

350

High-Level Liquid Waste Tank Integrity Workshop - 2008  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: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 toLiquid Waste Tank

351

High-Level Waste Corporate Board Meeting Agenda  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: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 toLiquid Waste

352

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

SciTech Connect (OSTI)

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.

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

1982-11-01T23:59:59.000Z

353

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

SciTech Connect (OSTI)

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

Boggs, C.J. [Argonne National Lab., Germantown, MD (United States); Shaddoan, W.T. [Lockheed Martin Energy Systems, Paducah, KY (United States)

1996-03-01T23:59:59.000Z

354

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

DOE Patents [OSTI]

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.

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

1989-01-01T23:59:59.000Z

355

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

SciTech Connect (OSTI)

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

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

2011-09-20T23:59:59.000Z

356

Waste Treatment Plant - 12508  

SciTech Connect (OSTI)

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)

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

2012-07-01T23:59:59.000Z

357

High-level waste vitrification off-gas cleanup technology  

SciTech Connect (OSTI)

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.

Hanson, M.S.

1980-01-01T23:59:59.000Z

358

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

SciTech Connect (OSTI)

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.

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

2004-08-31T23:59:59.000Z

359

Low level mixed waste thermal treatment technical basis report  

SciTech Connect (OSTI)

Detailed characterization of the existing and projected Hanford Site Radioactive Mixed Waste (RMW) inventory was initiated in 1993 (Place 1993). This report presents an analysis of the existing and projected RMW inventory. The subject characterization effort continues to be in support of the following engineering activities related to thermal treatment of Hanford Site RMW: (1) Contracting for commercial thermal treatment; (2) Installation and operation of an onsite thermal treatment facility (Project W-242); (3) Treatment at another Department of Energy (DOE) site. The collation of this characterization information (data) has emphasized the establishment of a common data base for the entire existing RMW inventory so that the specification of feed streams destined for different treatment facilities can be coordinated.

Place, B.G.

1994-12-01T23:59:59.000Z

360

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

SciTech Connect (OSTI)

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

Hill, Carolyn; Olson, Eric A.J.; Elmer, John [S.M. Stoller Corporation, Broomfield, Colorado 80021 (United States)

2012-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "level waste hlw" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM - 2011  

SciTech Connect (OSTI)

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.

West, B.; Waltz, R.

2012-06-21T23:59:59.000Z

362

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

SciTech Connect (OSTI)

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.

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

363

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

SciTech Connect (OSTI)

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

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

1983-12-01T23:59:59.000Z

364

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

SciTech Connect (OSTI)

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.

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

1989-09-01T23:59:59.000Z

365

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)

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.

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

2011-12-29T23:59:59.000Z

366

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

SciTech Connect (OSTI)

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

None

1980-06-01T23:59:59.000Z

367

Current Regulations and Guidance - New Approaches for Risk-Informed Low-Level Radioactive Waste Management  

SciTech Connect (OSTI)

This paper presents the historical foundations and future challenges for commercial low-level radioactive waste (LLRW) management in the United States. LLRW has been managed at government facilities since the beginning of the nuclear age and in the commercial sector since the early 1960's. Over the intervening years many technical, management and regulatory changes have occurred. Significant progress has been made in waste form, waste packaging and in recognizing radionuclides important to performance of disposal technologies and disposal facilities. This presentation will examine approaches using existing regulations and risk-informed approaches to improve guidance, licensing and management of LLRW. (authors)

Ryan, M.T. Ph.D.; CHP [Advisory Committee on Nuclear Waste and Materials, U.S. Nuclear Regulatory Commission, Washington, D.C. (United States)

2008-07-01T23:59:59.000Z

368

The dilemma of siting a high-level nuclear waste repository  

SciTech Connect (OSTI)

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.

Easterline, D.; Kunreuther, H.

1995-12-31T23:59:59.000Z

369

Waste Package Component Design Methodology Report  

SciTech Connect (OSTI)

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.

D.C. Mecham

2004-07-12T23:59:59.000Z

370

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

SciTech Connect (OSTI)

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

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

2012-05-01T23:59:59.000Z

371

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

SciTech Connect (OSTI)

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.

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

372

Low-level waste shallow burial assessment code  

SciTech Connect (OSTI)

PRESTO (Prediction of Radiation Exposures from Shallow Trench Operationns) is a computer code developed under United States Environmental Protection Agency funding to evaluate possible health effects from radionuclide releases from shallow, radioctive-waste disposal trenches and from areas contaminated with operational spillage. The model is intended to predict radionuclide transport and the ensuing exposure and health impact to a stable, local population for a 1000-year period following closure of the burial grounds. Several classes of submodels are used in PRESTO to represent scheduled events, unit system responses, and risk evaluation processes. The code is modular to permit future expansion and refinement. Near-surface transport mechanisms considered in the PRESTO code are cap failure, cap erosion, farming or reclamation practices, human intrusion, chemical exchange within an active surface soil layer, contamination from trench overflow, and dilution by surface streams. Subsurface processes include infiltration and drainage into the trench, the ensuing solubilization of radionuclides, and chemical exchange between trench water and buried solids. Mechanisms leading to contaminated outflow include trench overflow and downwad vertical percolation. If the latter outflow reaches an aquifer, radiological exposure from irrigation or domestic consumption is considered. Airborne exposure terms are evaluated using the Gaussian plume atmospheric transport formulation as implemented by Fields and Miller (1980).

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

1981-01-01T23:59:59.000Z

373

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)

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374

Remote-Handled Low-Level Waste Disposal Project Alternatives Analysis  

SciTech Connect (OSTI)

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

David Duncan

2009-10-01T23:59:59.000Z

375

Phase I high-level waste pretreatment and feed staging plan  

SciTech Connect (OSTI)

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).

Manuel, A.F.

1996-02-05T23:59:59.000Z

376

Remote-Handled Low-Level Waste Disposal Project Alternatives Analysis  

SciTech Connect (OSTI)

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

David Duncan

2010-06-01T23:59:59.000Z

377

The Savannah River Site Replacement High Level Radioactive Waste Evaporator Project  

SciTech Connect (OSTI)

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.

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

1992-01-01T23:59:59.000Z

378

The Savannah River Site Replacement High Level Radioactive Waste Evaporator Project  

SciTech Connect (OSTI)

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.

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

1992-08-01T23:59:59.000Z

379

Remote-Handled Low-Level Waste Disposal Project Alternatives Analysis  

SciTech Connect (OSTI)

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

David Duncan

2011-03-01T23:59:59.000Z

380

Remote-Handled Low-Level Waste Disposal Project Alternatives Analysis  

SciTech Connect (OSTI)

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

David Duncan

2011-04-01T23:59:59.000Z

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381

Thermal and Radiolytic Gas Generation in Hanford High-Level Waste  

SciTech Connect (OSTI)

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.

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

2000-01-31T23:59:59.000Z

382

On-Site Decontamination System for Liquid Low Level Radioactive Waste - 13010  

SciTech Connect (OSTI)

This study is based on an evaluation of purification methods for liquid low-level radioactive waste (LLLW) by using natural zeolite. Generally the volume of liquid low-level waste is relatively large and the specific activity is rather low when compared to other radioactive waste types. In this study, a pilot scale column was used with natural zeolite as an ion exchanger media. Decontamination and minimization of LLLW especially at the generation site decrease operational cost in waste management operations. Portable pilot scale column was constructed for decontamination of LLW on site. Effect of temperature on the radionuclide adsorption of the zeolite was determined to optimize the waste solution temperature for the plant scale operations. In addition, effect of pH on the radionuclide uptake of the zeolite column was determined to optimize the waste solution pH for the plant scale operations. The advantages of this method used for the processing of LLLW are discussed in this paper. (authors)

OSMANLIOGLU, Ahmet Erdal [Cekmece Nuclear Research and Training Center, Kucukcekmece Istanbul (Turkey)] [Cekmece Nuclear Research and Training Center, Kucukcekmece Istanbul (Turkey)

2013-07-01T23:59:59.000Z

383

Session 35 - Panel: Remaining US Disposition Issues for Orphan or Small Volume Low Level and Low Level Mixed Waste Streams  

SciTech Connect (OSTI)

Faced with closure schedules as a driving force, significant progress has been made during the last 2 years on the disposition of DOE mixed waste streams thought previously to be problematic. Generators, the Department of Energy and commercial vendors have combined to develop unique disposition paths for former orphan streams. Recent successes and remaining issues will be discussed. The session will also provide an opportunity for Federal agencies to share lessons learned on low- level and mixed low-level waste challenges and identify opportunities for future collaboration. This panel discussion was organized by PAC member Dick Blauvelt, Navarro Research and Engineering Inc who served as co-chair along with Dave Eaton from INL. In addition, George Antonucci, Duratek Barnwell and Rich Conley, AFSC were invited members of the audience, prepared to contribute the Barnwell and DOD perspective to the issues as needed. Mr. Small provide information regarding the five year 20K M3 window of opportunity at the Nevada Test Site for DOE contractors to dispose of mixed waste that cannot be received at the Energy Solutions (Envirocare) site in Utah because of activity levels. He provided a summary of the waste acceptance criteria and the process sites must follow to be certified to ship. When the volume limit or time limit is met, the site will undergo a RCRA closure. Ms. Gelles summarized the status of the orphan issues, commercial options and the impact of the EM reorganization on her program. She also announced that there would be a follow-on meeting in 2006 to the very successful St. Louis meeting of last year. It will probably take place in Chicago in July. Details to be announced. Mr. McKenney discussed progress made at the Hanford Reservation regarding disposal of their mixed waste inventory. The news is good for the Hanford site but not good for the rest of the DOE complex since shipment for out of state of both low level and low level mixed waste will continue to be prohibited until the completion of a new NEPA study. This is anticipated to take several years. Bill Franz from Portsmouth and Dave Eaton representing the INL provided the audience with information regarding some of the problematic mixed waste streams at their respective sites. Portsmouth has some unique radiological issues with isotopes such as Tc-99 while the INL is trying to deal with mixed waste in the 10-100 nCi/g range. Kaylin Loveland spoke of the new,Energy Solutions organization and provided information on mixed waste treatment capabilities at the Clive site. Mike Lauer described the licensing activities at the WCS site in Texas where they are trying to eventually have disposal capabilities for Class A, B and C mixed waste from both DOE and the commercial sector. The audience included about 75 WM'06 attendees who asked some excellent questions and provided an active and informative exchange of information on the topic. (authors)

Blauvelt, Richard [Navarro Engineering Research Inc. (United States); Small, Ken [Doe Nevada (United States); Gelles, Christine [DOE EM HQ (United States); McKenney, Dale [Fluor Hanford (United States); Franz, Bill [LATA Portsmouth (United States); Loveland, Kaylin [Energy Solutions Inc. (United States); Lauer, Mike [Waste Control Specialists (United States)

2006-07-01T23:59:59.000Z

384

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

SciTech Connect (OSTI)

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.

Crawford, T.W.

1995-09-22T23:59:59.000Z

385

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

SciTech Connect (OSTI)

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 disadv