Science.gov

Sample records for doe reprocessing waste

  1. Determination of the Structure of Vitrified Hydroceramic/CBC Waste Form Glasses Manufactured from DOE Reprocessing Waste

    SciTech Connect

    Scheetz, B.E.; White, W. B.; Chesleigh, M.; Portanova, A.; Olanrewaju, J.

    2005-05-31

    The selection of a glass-making option for the solidification of nuclear waste has dominated DOE waste form programs since the early 1980's. Both West Valley and Savannah River are routinely manufacturing glass logs from the high level waste inventory in tank sludges. However, for some wastes, direct conversion to glass is clearly not the optimum strategy for immobilization. INEEL, for example, has approximately 4400 m{sup 3} of calcined high level waste with an activity that produces approximately 45 watts/m{sup 3}, a rather low concentration of radioactive constituents. For these wastes, there is value in seeking alternatives to glass. An alternative approach has been developed and the efficacy of the process demonstrated that offers a significant savings in both human health and safety exposures and also a lower cost relative to the vitrification option. The alternative approach utilizes the intrinsic chemical reactivity of the highly alkaline waste with the addition of aluminosilicate admixtures in the appropriate proportions to form zeolites. The process is one in which a chemically bonded ceramic is produced. The driving force for reaction is derived from the chemical system itself at very modest temperatures and yet forms predominantly crystalline phases. Because the chemically bonded ceramic requires an aqueous medium to serve as a vehicle for the chemical reaction, the proposed zeolite-containing waste form can more adequately be described as a hydroceramic. The hydrated crystalline materials are then subject to hot isostatic pressing (HIP) which partially melts the material to form a glass ceramic. The scientific advantages of the hydroceramic/CBC approach are: (1) Low temperature processing; (2) High waste loading and thus only modest volumetric bulking from the addition of admixtures; (3) Ability to immobilize sodium; (4) Ability to handle low levels of nitrate (2-3% NO{sub 3}{sup -}); (5) The flexibility of a vitrifiable waste; and (6) A process that

  2. 77 FR 38789 - Notice of Availability of Draft Waste Incidental to Reprocessing Evaluation for the Concentrator...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-29

    ... incidental to reprocessing and thus are not high-level radioactive waste (HLW) and may be managed and....1-1, Radioactive Waste Management Manual. DOE is consulting with the Nuclear Regulatory Commission... would be disposed of at a suitable off-site LLW waste disposal facility, either the Area 5...

  3. Idaho Nuclear Technology and Engineering Center (INTEC) Sodium Bearing Waste - Waste Incidental to Reprocessing Determination

    SciTech Connect

    Jacobson, Victor Levon

    2002-08-01

    U.S. Department of Energy Manual 435.1-1, Radioactive Waste Management, Section I.1.C, requires that all radioactive waste subject to Department of Energy Order 435.1 be managed as high-level radioactive waste, transuranic waste, or low-level radioactive waste. Determining the radiological classification of the sodium-bearing waste currently in the Idaho Nuclear Technology and Engineering Center Tank Farm Facility inventory is important to its proper treatment and disposition. This report presents the technical basis for making the determination that the sodium-bearing waste is waste incidental to spent fuel reprocessing and should be managed as mixed transuranic waste. This report focuses on the radiological characteristics of the sodiumbearing waste. The report does not address characterization of the nonradiological, hazardous constituents of the waste in accordance with Resource Conservation and Recovery Act requirements.

  4. Waste Minimization Study on Pyrochemical Reprocessing Processes

    SciTech Connect

    Boussier, H.; Conocar, O.; Lacquement, J.

    2006-07-01

    Ideally a new pyro-process should not generate more waste, and should be at least as safe and cost effective as the hydrometallurgical processes currently implemented at industrial scale. This paper describes the thought process, the methodology and some results obtained by process integration studies to devise potential pyro-processes and to assess their capability of achieving this challenging objective. As example the assessment of a process based on salt/metal reductive extraction, designed for the reprocessing of Generation IV carbide spent fuels, is developed. Salt/metal reductive extraction uses the capability of some metals, aluminum in this case, to selectively reduce actinide fluorides previously dissolved in a fluoride salt bath. The reduced actinides enter the metal phase from which they are subsequently recovered; the fission products remain in the salt phase. In fact, the process is not so simple, as it requires upstream and downstream subsidiary steps. All these process steps generate secondary waste flows representing sources of actinide leakage and/or FP discharge. In aqueous processes the main solvent (nitric acid solution) has a low boiling point and evaporate easily or can be removed by distillation, thereby leaving limited flow containing the dissolved substance behind to be incorporated in a confinement matrix. From the point of view of waste generation, one main handicap of molten salt processes, is that the saline phase (fluoride in our case) used as solvent is of same nature than the solutes (radionuclides fluorides) and has a quite high boiling point. So it is not so easy, than it is with aqueous solutions, to separate solvent and solutes in order to confine only radioactive material and limit the final waste flows. Starting from the initial block diagram devised two years ago, the paper shows how process integration studies were able to propose process fittings which lead to a reduction of the waste variety and flows leading at an 'ideal

  5. Gloves Reprocessing: Does It Really Save Money?

    PubMed

    Arora, Pankaj; Kumari, Santosh; Sodhi, Jitender; Talati, Shweta; Gupta, Anil Kumar

    2015-12-01

    Gloves are reprocessed and reused in health-care facilities in resource-limited settings to reduce the cost of availability of gloves. The study was done with the aim to compute the cost of reprocessing of gloves so that an economically rationale decision can be taken. A retrospective record-based cross-sectional study was undertaken in a central sterile supply department where different steps during reprocessing of gloves were identified and the cost involved in reprocessing per pair of gloves was calculated. The cost of material and manpower was calculated to arrive at the cost of reprocessing per pair of gloves. The cost of a reprocessed pair of surgical gloves was calculated to be Indian Rupee (INR) 14.33 which was greater than the cost of a new pair of disposable surgical gloves (INR 9.90) as the cost of sterilization of one pair of gloves itself came out to  be INR 10.97. The current study showed that the purchase of sterile disposable single-use gloves is cheaper than the process of recycling. Reprocessing of gloves is not economical on tangible terms even in resource-limited settings, and from the perspective of better infection control as well as health-care worker safety, it further justifies the use of disposable gloves.

  6. Mesoscale to plant-scale models of nuclear waste reprocessing.

    SciTech Connect

    Noble, David Frederick; O'Hern, Timothy John; Moffat, Harry K.; Nemer, Martin B.; Domino, Stefan Paul; Rao, Rekha Ranjana; Cipiti, Benjamin B.; Brotherton, Christopher M.; Jove-Colon, Carlos F.; Pawlowski, Roger Patrick

    2010-09-01

    Imported oil exacerabates our trade deficit and funds anti-American regimes. Nuclear Energy (NE) is a demonstrated technology with high efficiency. NE's two biggest political detriments are possible accidents and nuclear waste disposal. For NE policy, proliferation is the biggest obstacle. Nuclear waste can be reduced through reprocessing, where fuel rods are separated into various streams, some of which can be reused in reactors. Current process developed in the 1950s is dirty and expensive, U/Pu separation is the most critical. Fuel rods are sheared and dissolved in acid to extract fissile material in a centrifugal contactor. Plants have many contacts in series with other separations. We have taken a science and simulation-based approach to develop a modern reprocessing plant. Models of reprocessing plants are needed to support nuclear materials accountancy, nonproliferation, plant design, and plant scale-up.

  7. Plasma Mass Filters For Nuclear Waste Reprocessing

    SciTech Connect

    Fetterman, Abraham J.; Fisch, Nathaniel J.

    2011-05-01

    Practical disposal of nuclear waste requires high-throughput separation techniques. The most dangerous part of nuclear waste is the fission product, which contains the most active and mobile radioisotopes and produces most of the heat. We suggest that the fission products could be separated as a group from nuclear waste using plasma mass filters. Plasmabased processes are well suited to separating nuclear waste, because mass rather than chemical properties are used for separation. A single plasma stage can replace several stages of chemical separation, producing separate streams of bulk elements, fission products, and actinoids. The plasma mass filters may have lower cost and produce less auxiliary waste than chemical processing plants. Three rotating plasma configurations are considered that act as mass filters: the plasma centrifuge, the Ohkawa filter, and the asymmetric centrifugal trap.

  8. Plasma Mass Filters For Nuclear Waste Reprocessing

    SciTech Connect

    Abraham J. Fetterman and Nathaniel J. Fisch

    2011-05-26

    Practical disposal of nuclear waste requires high-throughput separation techniques. The most dangerous part of nuclear waste is the fission product, which contains the most active and mobile radioisotopes and produces most of the heat. We suggest that the fission products could be separated as a group from nuclear waste using plasma mass filters. Plasmabased processes are well suited to separating nuclear waste, because mass rather than chemical properties are used for separation. A single plasma stage can replace several stages of chemical separation, producing separate streams of bulk elements, fission products, and actinoids. The plasma mass filters may have lower cost and produce less auxiliary waste than chemical processing plants. Three rotating plasma configurations are considered that act as mass filters: the plasma centrifuge, the Ohkawa filter, and the asymmetric centrifugal trap.

  9. Repository disposal requirements for commercial transuranic wastes (generated without reprocessing)

    SciTech Connect

    Daling, P.M.; Ludwick, J.D.; Mellinger, G.B.; McKee, R.W.

    1986-06-01

    This report forms a preliminary planning basis for disposal of commercial transuranic (TRU) wastes in a geologic repository. Because of the unlikely prospects for commercial spent nuclear fuel reprocessing in the near-term, this report focuses on TRU wastes generated in a once-through nuclear fuel cycle. The four main objectives of this study were to: develop estimates of the current inventories, projected generation rates, and characteristics of commercial TRU wastes; develop proposed acceptance requirements for TRU wastes forms and waste canisters that ensure a safe and effective disposal system; develop certification procedures and processing requirements that ensure that TRU wastes delivered to a repository for disposal meet all applicable waste acceptance requirements; and identify alternative conceptual strategies for treatment and certification of commercial TRU first objective was accomplished through a survey of commercial producers of TRU wastes. The TRU waste acceptance and certification requirements that were developed were based on regulatory requirements, information in the literature, and from similar requirements already established for disposal of defense TRU wastes in the Waste Isolation Pilot Plant (WIPP) which were adapted, where necessary, to disposal of commercial TRU wastes. The results of the TRU waste-producer survey indicated that there were a relatively large number of producers of small quantities of TRU wastes.

  10. Process for recovery of palladium from nuclear fuel reprocessing wastes

    DOEpatents

    Campbell, D.O.; Buxton, S.R.

    1980-06-16

    Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M; (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound; (c) heating the solution at reflux temperature until precipitation is complete; and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

  11. Process for recovery of palladium from nuclear fuel reprocessing wastes

    DOEpatents

    Campbell, David O.; Buxton, Samuel R.

    1981-01-01

    Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M, (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound, (c) heating the solution at reflux temperature until precipitation is complete, and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

  12. Vitrification of IFR and MSBR halide salt reprocessing wastes

    SciTech Connect

    Siemer, D.D.

    2013-07-01

    Both of the genuinely sustainable (breeder) nuclear fuel cycles (IFR - Integral Fast Reactor - and MSBR - Molten Salt Breeder Reactor -) studied by the USA's national laboratories would generate high level reprocessing waste (HLRW) streams consisting of a relatively small amount ( about 4 mole %) of fission product halide (chloride or fluoride) salts in a matrix comprised primarily (about 95 mole %) of non radioactive alkali metal halide salts. Because leach resistant glasses cannot accommodate much of any of the halides, most of the treatment scenarios previously envisioned for such HLRW have assumed a monolithic waste form comprised of a synthetic analog of an insoluble crystalline halide mineral. In practice, this translates to making a 'substituted' sodalite ('Ceramic Waste Form') of the IFR's chloride salt-based wastes and fluoroapatite of the MSBR's fluoride salt-based wastes. This paper discusses my experimental studies of an alternative waste management scenario for both fuel cycles that would separate/recycle the waste's halide and immobilize everything else in iron phosphate (Fe-P) glass. It will describe both how the work was done and what its results indicate about how a treatment process for both of those wastes should be implemented (fluoride and chloride behave differently). In either case, this scenario's primary advantages include much higher waste loadings, much lower overall cost, and the generation of a product (glass) that is more consistent with current waste management practices. (author)

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

    SciTech Connect

    Sullivan, Dan; Suttora, Linda; Goldston, Sonny; Petras, Robert; Rowell, Laurene; McNeil, Jim

    2012-07-01

    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)

  14. Reduction of Sodium Nitrate Liquid Waste in Nuclear Reprocessing Plants

    SciTech Connect

    Numata, M.; Mihara, S.; Kojima, S.; Ito, H.; Kato, T.

    2006-07-01

    Sodium nitrate solution has been generated from nuclear reprocessing plant as a result of neutralization of nitric acid. The sodium nitrate has been immobilized by bitumen, cement or other material in the site and waste packages have been produced. In order to reduce an environmental impact of the waste packages from the reprocessing plant, it is preferable to decompose nitrate ion to harmless gases such as nitrogen. A combination of formic acid and catalyst has been proposed for this purpose. But, the method is inadequate for a full decomposition of the nitrate ion. In addition, a mixture of NO and NO{sub 2} is produced during the reaction. Formaldehyde and hydrazine were selected as reductants and a combined use of Pd-Cu catalyst was tried to decompose the nitrate ion. As a result, the nitrate ion can almost entirely be decomposed without any generation of NO and NO{sub 2}. The test was conducted by 1 L flask. In case of formaldehyde, nitrate ion concentration can be reduced from 0.017 mol/l to 3.9x10{sup -4} mol/l. In case of hydrazine, nitrate concentration can be decreased from 2.8 mol/l to 9.5 x 10{sup -3} mol/l and ammonium ion is detected. The ammonium ion concentration in the final solution is 0.12 mol/l when 2.8 mol/l nitrate is reduced by hydrazine. Chemical reactions for formaldehyde on the Pd-Cu catalyst are estimated as combination of: NO{sub 3-} + HCHO = NO{sub 2-} + HCOOH; 2NO{sub 2-} + 3HCOOH = N{sub 2} + 3CO{sub 2} + 2H{sub 2}O + 2OH-; 4NO{sub 2-} + 3HCHO = 2N{sub 2} + 3CO{sub 2} + H{sub 2}O + 4OH-. the other hand, for hydrazine with the Pd-Cu catalyst: 3N{sub 2}H{sub 4} = 2NH{sub 3} + 2N{sub 2} + 3H{sub 2}; NO{sub 3-} + H{sub 2} = NO{sub 2-} + H{sub 2}O; NO{sub 2-} + NH{sub 3} = N{sub 2} + H{sub 2}O + OH-. The fundamental research shows that the combination usage of the Pd-Cu catalyst and formaldehyde or hydrazine is applicable for the reduction of nitrate liquid waste in the nuclear reprocessing plant. (authors)

  15. Waste management system alternatives for treatment of wastes from spent fuel reprocessing

    SciTech Connect

    McKee, R.W.; Swanson, J.L.; Daling, P.M.; Clark, L.L.; Craig, R.A.; Nesbitt, J.F.; McCarthy, D.; Franklin, A.L.; Hazelton, R.F.; Lundgren, R.A.

    1986-09-01

    This study was performed to help identify a preferred TRU waste treatment alternative for reprocessing wastes with respect to waste form performance in a geologic repository, near-term waste management system risks, and minimum waste management system costs. The results were intended for use in developing TRU waste acceptance requirements that may be needed to meet regulatory requirements for disposal of TRU wastes in a geologic repository. The waste management system components included in this analysis are waste treatment and packaging, transportation, and disposal. The major features of the TRU waste treatment alternatives examined here include: (1) packaging (as-produced) without treatment (PWOT); (2) compaction of hulls and other compactable wastes; (3) incineration of combustibles with cementation of the ash plus compaction of hulls and filters; (4) melting of hulls and failed equipment plus incineration of combustibles with vitrification of the ash along with the HLW; (5a) decontamination of hulls and failed equipment to produce LLW plus incineration and incorporation of ash and other inert wastes into HLW glass; and (5b) variation of this fifth treatment alternative in which the incineration ash is incorporated into a separate TRU waste glass. The six alternative processing system concepts provide progressively increasing levels of TRU waste consolidation and TRU waste form integrity. Vitrification of HLW and intermediate-level liquid wastes (ILLW) was assumed in all cases.

  16. Waste-Incidental-to-Reprocessing Evaluation for the West Valley Demonstration Project Vitrification Melter - 12167

    SciTech Connect

    McNeil, Jim; Kurasch, David; Sullivan, Dan; Crandall, Thomas

    2012-07-01

    The Department of Energy (DOE) has determined that the vitrification melter used in the West Valley Demonstration Project can be disposed of as low-level waste (LLW) after completion of a waste-incidental-to-reprocessing evaluation performed in accordance with the evaluation process of DOE Manual 435.1-1, Radioactive Waste Management Manual. The vitrification melter - which consists of a ceramic lined, electrically heated box structure - was operated for more than 5 years melting and fusing high-level waste (HLW) slurry and glass formers and pouring the molten glass into 275 stainless steel canisters. Prior to shutdown, the melter was decontaminated by processing low-activity decontamination flush solutions and by extracting molten glass from the melter cavity. Because it could not be completely emptied, residual radioactivity conservatively estimated at approximately 170 TBq (4,600 Ci) remained in the vitrification melter. To establish whether the melter was incidental to reprocessing, DOE prepared an evaluation to demonstrate that the vitrification melter: (1) had been processed to remove key radionuclides to the maximum extent technically and economically practical; (2) would be managed to meet safety requirements comparable to the performance objectives for LLW established by the Nuclear Regulatory Commission (NRC); and (3) would be managed by DOE in accordance with DOE's requirements for LLW after it had been incorporated in a solid physical form with radionuclide concentrations that do not exceed the NRC concentration limits for Class C LLW. DOE consulted with the NRC on the draft evaluation and gave other stakeholders an opportunity to submit comments before the determination was made. The NRC submitted a request for additional information in connection with staff review of the draft evaluation; DOE provided the additional information and made improvements to the evaluation, which was issued in January 2012. DOE considered the NRC Technical Evaluation Report

  17. Is Spent Fuel or Waste from Reprocessed Spent Fuel Simpler to Dispose of?

    DTIC Science & Technology

    1981-06-12

    the space in a repository for processed high-level waste has not included the disposal of transuranic waste from the co.mercial fuel cycle which will...AD-AI06 573 GENERAL ACCOUNTING OFFICE WASHINGTON OC ENERGY AND M-ETC F/G 18/7 i S PENT FUEL OR WASTE FROM REPROCESSED SPENT FUEL SIMPLER TO D--TC(U...the Congress entitled "Is Spent Fuel or Waste from Reprocessed Spent Fuel Simpler to Dispose of?" (EID-81-78, June 12, 1981)z Insert the following

  18. Materials used in low-level liquid waste reprocessing/treatment studies at Oak Ridge National Laboratory

    SciTech Connect

    McDaniel, E.W.; Weeren, H.O.; Delzer, D.B.; Sams, T.L.; Tallent, O.K.

    1987-01-01

    The importance of effective waste management in the nuclear fuel cycle cannot be overestimated. At Oak Ridge National Laboratory (ORNL), development work in waste reprocessing and treatment includes the testing and use of various additives for the purpose of facilitating adherence to both process and regulatory performance criteria. Three waste reprocessing/treatment technologies and the associated materials are discussed in this paper: (1) suspension and transfer of sludge from waste storage tanks; (2) treatment to render a waste in compliance with regulatory requirements; and (3) fluoride-rich waste reprocessing. 7 refs., 3 figs.

  19. Method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions

    DOEpatents

    Horwitz, E. Philip; Delphin, Walter H.

    1979-07-24

    A method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions containing these and other values by contacting the waste solution with an extractant of tricaprylmethylammonium nitrate in an inert hydrocarbon diluent which extracts the palladium and technetium values from the waste solution. The palladium and technetium values are recovered from the extractant and from any other coextracted values with a strong nitric acid strip solution.

  20. Removal of actinides from nuclear reprocessing wastes: a pilot plant study using non-radioactive simulants

    SciTech Connect

    Maxey, H.R.; McIsaac, L.D.; Chamberlain, D.B.; McManus, G.J.

    1980-01-01

    Nuclear fuel reprocessing wastes generated at the ICPP contain small amounts of actinides, primarily Pu and Am. Removal of these actinides reduces the long term storage hazards of the waste. The development of a flowsheet to remove trivalent actinides is discussed in this paper. Pilot plant studies used actinide simulants. As a result of these studies, the Height of a Transfer Unit (HTU) was selected as the better measure of pulse column separation efficiency.

  1. A novel waste form for disposal of spent-nuclear-fuel reprocessing waste: A vitrifiable cement

    SciTech Connect

    Gougar, M.L.D.; Scheetz, B.E.; Siemer, D.D.

    1999-01-01

    A cement capable of being hot isostatically pressed into a glass ceramic has been proposed as the waste form for spent-nuclear-fuel reprocessing wastes at the Idaho National Engineering and Environmental Laboratory (INEEL). This intermediate cement, with a composition based on that of common glasses, has been designed and tested. The cement formulations included mixed INEEL wastes, blast furnace slag, reactive silica, and INEEL soil or vermiculite, which were activated with potassium or sodium hydroxide. Following autoclave processing, the cements were characterized. X-ray diffraction analysis revealed three notable crystalline phases: quartz, calcite, and fluorite. Results of compressive strength testing ranged from 1452 and 4163 psi, exceeding the US Nuclear Regulatory Commission (NRC)-suggested standard of >500 psi. From American National Standards Institute/American Nuclear Society 16.1-1986 leach testing, effective diffusivities for Cs were determined to be on the order of 10{sup {minus}11} to 10{sup {minus}10} cm{sup 2}/s and for Sr were 10{sup {minus}12} cm{sup 2}/s, which are four orders of magnitude less than diffusivities in some other radwaste materials. Average leach indices (LI) were 9.6 and 11.9 for Cs and Sr, respectively, meeting the NRC Standard of LI > 6. The 28-day Materials Characterization Center-1 leach testing resulted in normalized elemental mass losses between 0.63 and 28 g/(m{sup 2}{center_dot}day) for Cs and between 0.34 and 0.70 g/(m{sup 2}{center_dot}day) industry-accepted standard while Cs losses indicate a process sensitive parameter.

  2. Improvement of technology for reprocessing of low-level wastes with the use of ozone

    SciTech Connect

    Revenko, Y.A.; Manakov, S.A.; Petrov, A.I.

    1995-12-31

    An original flowsheet is proposed for reprocessing of low-level wastes (LLW) containing surfactants. The flowsheet involves the use of ozone for destruction of surfactants and clinoptilolite for purification from ozonolysis products. Testing of the process in a pilot facility has shown the possibility for reduction of radioactive slurries by a factor of 10--15 with increasing performance of one filter-cycle by a factor of 2--3.

  3. Conservatism in effective dose calculations for accident events involving fuel reprocessing waste tanks.

    PubMed

    Bevelacqua, J J

    2011-07-01

    Conservatism in the calculation of the effective dose following an airborne release from an accident involving a fuel reprocessing waste tank is examined. Within the regulatory constraints at the Hanford Site, deterministic effective dose calculations are conservative by at least an order of magnitude. Deterministic calculations should be used with caution in reaching decisions associated with required safety systems and mitigation philosophy related to the accidental release of airborne radioactive material to the environment.

  4. Prospects for using membrane distallation for reprocessing liquid radioactive wastes

    SciTech Connect

    Dytnerskii, Y.I.; Karlin, Y.V.; Kropotov, B.N.

    1994-05-01

    Membrane distillation is a promising method for deep desalinization and for removal of impurities of different nature from water. The crux of the method is as follows. The initial (hot) solution, heated up to 30-70{degrees}C, is fed into one side of a hydrophobic microporous membrane. A less heated (cold) distillate moves along the other. Since the membrane is hydrophobic and the pores are small ({approximately}1 {mu}m and less), the liquid phase does not penetrate into the pores in accordance with Kelvin`s law. The vapor evaporating from the surface of the hot solution (the evaporation surface in this case are solution meniscuses forming at the entrance into a pore) penetrates into the pores of the membrane, diffuses through the air layer in the pore, and condenses on the surface of the menisci of cold liquid. In the process rarefaction is produced in the pores, and this accelerates evaporation and therefore increases its efficiency.

  5. DOE Waste Treatability Group Guidance

    SciTech Connect

    Kirkpatrick, T.D.

    1995-01-01

    This guidance presents a method and definitions for aggregating U.S. Department of Energy (DOE) waste into streams and treatability groups based on characteristic parameters that influence waste management technology needs. Adaptable to all DOE waste types (i.e., radioactive waste, hazardous waste, mixed waste, sanitary waste), the guidance establishes categories and definitions that reflect variations within the radiological, matrix (e.g., bulk physical/chemical form), and regulated contaminant characteristics of DOE waste. Beginning at the waste container level, the guidance presents a logical approach to implementing the characteristic parameter categories as part of the basis for defining waste streams and as the sole basis for assigning streams to treatability groups. Implementation of this guidance at each DOE site will facilitate the development of technically defined, site-specific waste stream data sets to support waste management planning and reporting activities. Consistent implementation at all of the sites will enable aggregation of the site-specific waste stream data sets into comparable national data sets to support these activities at a DOE complex-wide level.

  6. Detection of Cm-224 in plutonium-bearing wastes at reprocessing facilities

    SciTech Connect

    Beddingfield, D. H.; Belian, A. P.

    2004-01-01

    There is concern that small amounts of curium present in reprocessing facility low-activity wastes can interfere with plutonium verification measurement of these wastes. In this paper we will present calculational data showing the effect of curium presence on verification measurements and methods of discriminating between curium and plutonium for certain cases that may be encountered during verification activity. For curium contamination greater than 1 atom per 10{sup 6} Pu atoms, the presence of Cm in low-activity wastes can be determined by examining the D/T ratio observed from the sample. In the range of 1 to 1000 ppm it is possible to independently quantify both Pu and Cm in sample. Above 1000 ppm Cm the sample is indistinguishable from pure Cm using only this analysis. The line labeled SNF shown in Figure 1 and 2 is the typical Cm/Pu ratio of spent fuel cooled for 10 years. The ability to quantifiably distinguish between Pu and Cm only exists below the SNF ratio. First separation cycle wastes will typically have Cm concentrations above the SNF ratio and thus will be indistinguishable from pure Cm. And quantification of Pu in these samples is not possible from this technique. However, in the reverse, the analysis method will allow the certification of the absence of Cm in a particular sample originating from the first separation cycle. Wastes generated after the first separation cycle will have Cm contamination below the SNF line and will lend themselves to independent determination of the Pu and Cm mass or the certification of the absence of Pu in the waste. Despite the limited range of curium contamination where Cm and Pu can be independently determined by this technique, the ability to certify the absence of Cm in a waste drum with only passive neutron measurement is very useful when performing waste measurements from reprocessing facilities. If the wastes are being assayed for 'measured discard' purposes the possibility of a significant bias in the reported

  7. Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

    DOE PAGES

    Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; ...

    2017-08-30

    Here, this paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy.more » These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability.« less

  8. Vitrifiable concrete for disposal of spent nuclear fuel reprocessing waste at I.N.E.L.

    SciTech Connect

    Gougar, M.L.D.; Scheetz, B.E.; Siemer, D.D.

    1996-08-01

    A cement capable of being Hot Isostatically Pressed (HIP`ed) into a glass-ceramic has been proposed for use as the waste form for SNF reprocessing wastes at the Idaho National Engineering Laboratories. Such an ``intermediate`` cement, with a composition based on that of common glasses, has been designed and tested. The cement formulations included mixed I.N.E.L. wastes, blast furnace flag, reactive silica, alumina, and I.N.E.L. soil or vermiculite, which was activated with potassium or sodium hydroxide. Modified FUETAP processing was performed and the cement was subsequently characterized. Results of compressive strength testing ranged from 1,452 psi to 4,163 psi, exceeding the NRC-suggested standard of >500 psi. Total dissolved solids concentrations in waste form leachates were calculated from a static leach test in which leachate conductivity was measured. Effective diffusivities for radioisotopes Cs and Sr were calculated from leachate analysis data. Diffusivity values were on the order of 10{sup {minus}15} to 10{sup {minus}10} cm{sup 2}/sec, which compare favorably with diffusivities in other materials.

  9. PRELIMINARY STUDY OF CERAMICS FOR IMMOBILIZATION OF ADVANCED FUEL CYCLE REPROCESSING WASTES

    SciTech Connect

    Fox, K.; Billings, A.; Brinkman, K.; Marra, J.

    2010-09-22

    The Savannah River National Laboratory (SRNL) developed a series of ceramic waste forms for the immobilization of Cesium/Lanthanide (CS/LN) and Cesium/Lanthanide/Transition Metal (CS/LN/TM) waste streams anticipated to result from nuclear fuel reprocessing. Simple raw materials, including Al{sub 2}O{sub 3}, CaO, and TiO{sub 2} were combined with simulated waste components to produce multiphase ceramics containing hollandite-type phases, perovskites (particularly BaTiO{sub 3}), pyrochlores, zirconolite, and other minor metal titanate phases. Identification of excess Al{sub 2}O{sub 3} via X-ray Diffraction (XRD) and Scanning Electron Microscopy with Energy Dispersive Spectroscopy (SEM/EDS) in the first series of compositions led to a Phase II study, with significantly reduced Al{sub 2}O{sub 3} concentrations and increased waste loadings. Three fabrication methodologies were used, including melting and crystallizing, pressing and sintering, and Spark Plasma Sintering (SPS), with the intent of studying phase evolution under various sintering conditions. XRD and SEM/EDS results showed that the partitioning of the waste elements in the sintered materials was very similar, despite varying stoichiometry of the phases formed. The Phase II compositions generally contained a reduced amount of unreacted Al{sub 2}O{sub 3} as identified by XRD, and had phase assemblages that were closer to the initial targets. Chemical composition measurements showed no significant issues with meeting the target compositions. However, volatilization of Cs and Mo was identified, particularly during melting, since sintering of the pressed pellets and SPS were performed at lower temperatures. Partitioning of some of the waste components was difficult to determine via XRD. SEM/EDS mapping showed that those elements, which were generally present in small concentrations, were well distributed throughout the waste forms. Initial studies of radiation damage tolerance using ion beam irradiation at Los

  10. Accumulation of COGEMA-La Hague-derived reprocessing wastes in French salt marsh sediments.

    PubMed

    Cundy, Andrew B; Croudace, Ian W; Warwick, Phillip E; Oh, Jung-Suk; Haslett, Simon K

    2002-12-01

    sedimentation in the central Channel. These marshes are not major sinks for discharged reprocessing wastes.

  11. Silica-based waste form for immobilization of iodine from reprocessing plant off-gas streams

    SciTech Connect

    Matyáš, Josef; Canfield, Nathan; Sulaiman, Sannoh; Zumhoff, Mac

    2016-08-01

    A high selectivity and sorption capacity for iodine and a feasible consolidation to a durable SiO2-based waste form makes silver-functionalized silica aerogel (Ag0-aerogel) an attractive choice for the removal and sequestration of iodine compounds from the off-gas of a nuclear fuel reprocessing plant. Hot uniaxial pressing of iodine-loaded Ag0-aerogel (20.2 mass% iodine) at 1200°C for 30 min under 29 MPa pressure provided a partially sintered product with residual open porosity of 16.9% that retained ~93% of sorbed iodine. Highly iodine-loaded Ag0-aerogel was successfully consolidated by hot isostatic pressing at 1200°C with a 30-min hold and under 207 MPa. The fully densified waste form had a bulk density of 3.3 g/cm3 and contained ~39 mass% iodine. The iodine was retained in the form of nano- and micro-particles of AgI that were uniformly distributed inside and along boundaries of fused silica grains.

  12. Silica-based waste form for immobilization of iodine from reprocessing plant off-gas streams

    NASA Astrophysics Data System (ADS)

    Matyáš, Josef; Canfield, Nathan; Sulaiman, Sannoh; Zumhoff, Mac

    2016-08-01

    A high selectivity and sorption capacity for iodine and a feasible consolidation to a durable SiO2-based waste form makes silver-functionalized silica aerogel (Ag0-aerogel) an attractive choice for the removal and sequestration of iodine compounds from the off-gas of a nuclear fuel reprocessing plant. Hot uniaxial pressing of iodine-loaded Ag0-aerogel (20.2 mass% iodine) at 1200 °C for 30 min under 29 MPa pressure provided a partially sintered product with residual open porosity of 16.9% that retained ∼93% of sorbed iodine. Highly iodine-loaded Ag0-aerogel was successfully consolidated by hot isostatic pressing at 1200 °C with a 30-min hold and under 207 MPa. The fully densified waste form had a bulk density of 3.3 × 103 kg/m3 and contained ∼39 mass% iodine. The iodine was retained in the form of nano- and micro-particles of AgI that were uniformly distributed inside and along boundaries of fused silica grains.

  13. Negotiating equity for management of DOE wastes

    SciTech Connect

    Carnes, S.A.

    1994-09-01

    One important factor frustrating optimal management of Department of Energy (DOE)-complex wastes is the inability to use licensed and permitted facilities systematically. Achieving the goal of optimal use of DOE`s waste management facilities is politically problematic for two reasons. First, no locale wants to bear a disproportionate burden from DOE wastes. Second, the burden imposed by additional wastes transported from one site to another is difficult to characterize. To develop a viable framework for equitably distributing these burdens while achieving efficient use of all DOE waste management facilities, several implementation and equity issues must be addressed and resolved. This paper discusses stakeholder and equity issues and proposes a framework for joint research and action that could facilitate equity negotiations among stakeholder and move toward a more optimal use of DOE`s waste management capabilities.

  14. Sensitization and Intergranular Corrosion Behavior of High Nitrogen Type 304LN Stainless Steels for Reprocessing and Waste Management Applications

    NASA Astrophysics Data System (ADS)

    Parvathavarthini, N.; Kamachi Mudali, U.; Nenova, Lilyana; Andreev, Chavdar; Raj, Baldev

    2012-06-01

    High nitrogen 304LN stainless steels (SS) intended for chloride and nitric acid environments in spent nuclear fuel reprocessing and waste management applications were evaluated for their sensitization and intergranular corrosion (IGC) resistance. For this purpose, high nitrogen (0.132 pct, 0.193 pct and 0.406 pct) containing, impurity-controlled, vanadium-added 304LN SS alloys were developed. For comparison, 304L SS, which is currently used in reprocessing plants, was also studied. These stainless steels were subjected to heat treatment at 948 K (675 °C) for various durations ranging from 1 to 1000 hours and tested for susceptibility to IGC as per ASTM A262 Practice A and E tests. The degree of sensitization was estimated with the double loop electrochemical potentiokinetic reactivation technique. The increase in nitrogen content resulted in higher hardness and finer grain size. Based on the detailed microstructural and corrosion studies, it was determined that an addition of 0.132 pct and 0.193 pct nitrogen showed better IGC resistance and an additional increase in nitrogen resulted in deterioration resulting from chromium nitride precipitation, which was confirmed by electrochemical phase separation and X-ray diffraction studies. The onset of desensitization was faster for the alloy with 0.132 pct nitrogen as well as 0.406 pct nitrogen because of the lower nitrogen content in the former case and the finer grain size in the latter case. The higher hardness and superior IGC resistance of 0.132 pct and 0.193 pct nitrogen containing Type 304LN SS suggests the suitability of this alloy for nitric acid- and chloride-containing environments of reprocessing and waste management plants.

  15. Solidification of DOE problem wastes

    SciTech Connect

    Franz, E.M.; Heiser, J.H. III; Colombo, P.

    1986-01-01

    Sodium nitrate waste has been successfully solidified in two types of polymeric materials: polyethylene, a thermoplastic material, and polyester styrene (PES), a thermosetting material. Waste form property evaluation tests such as ANS 16.1 leaching test and compressive strength measurements were performed on the waste forms containing various amounts of sodium nitrate. A single-screw extruder was employed for incorporating dry waste into polyethylene at its melt temperature of 120/sup 0/C to produce a homogenous mixture. Results of the leaching test for polyethylene waste forms containing 30, 50, 60 and 70 wt% sodium nitrate are presented as cumulative fraction leached and leaching indices ranging from 11 to 7.8. Two PES systems are discussed. The first is for solidification of dry salt wastes and the second is a water extendible system that is compatible with wet waste streams. Leaching data for PES and water extendible PES waste forms containing 30 wt% sodium nitrate are presented as cumulative fraction leached and leaching indices of approximately 9. Results from compressive strength measurements are also included.

  16. DOE complex buried waste characterization assessment

    SciTech Connect

    Kaae, P.S.; Holter, G.M.; Garrett, S.M.K.

    1993-01-01

    The work described in this report was conducted by Pacific Northwest Laboratory to provide information to the Buried Waste Integrated Demonstration (BWID) program. The information in this report is intended to provide a complex-wide planning base for th.e BWID to ensure that BWID activities are appropriately focused to address the range of remediation problems existing across the US Department of Energy (DOE) complex. This report contains information characterizing the 2.1 million m[sup 3] of buried and stored wastes and their associated sites at six major DOE facilities. Approximately 85% of this waste is low-level waste, with about 12% TRU or TRU mixed waste; the remaining 3% is low-level mixed waste. In addition, the report describes soil contamination sites across the complex. Some of the details that would be useful in further characterizing the buried wastes and contaminated soil sites across the DOE complex are either unavailable or difficult to locate. Several options for accessing this information and/or improving the information that is available are identified in the report. This document is a companion to Technology Needs for Remediation: Hanford and Other DOE Sites, PNL-8328 (Stapp 1993).

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

    SciTech Connect

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

    1996-03-01

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

  18. Full-scale experimental facility for the development technologies for the reprocessing of tritium contaminated light and heavy water wastes by CECE process and cryogenic distillation

    SciTech Connect

    Trenin, V.D.; Alekseev, I.A.; Karpov, S.P.; Bondarenko, S.D.; Vasyanina, T.V.; Konoplev, K.A.; Fedorchenko, O.A.; Uborski, V.V.; Voronina, T.

    1995-10-01

    The problem of the formation and accumulation of the tritiated heavy and light water wastes produced under operation of the various nuclear facilities is considered. It is shown that the tritium contaminated wastes may have a wide spectrum of isotope concentrations of H:D:T and correlation one with other. Reprocessing of these wastes is expensive matter due to the small tritium concentration respectfully to other hydrogen isotopes and as well as the small value of separation factor. It requires the development of the versatile technology. The description of the full scale experimental facility constructed at PNPI is given. 18 refs., 1 fig.

  19. Estimation of carbon 14 inventory in hull and end-piece wastes from Japanese commercial reprocessing operation

    SciTech Connect

    Tomofumi Sakuragi; Hiromi Tanabe; Emiko Hirose; Akira Sakashita; Tsutomu Nishimura

    2013-07-01

    Hull and end-piece wastes generated from reprocessing plant operations are expected to be disposed of in a deep underground repository as Group 2 TRU wastes under the Japanese classification system. The activated metals that compose the spent fuel assemblies such as Zircaloy claddings and stainless steel nozzles are mixed and compressed after fuel dissolution, and then stuffed into stainless steel canisters. Carbon 14 is a typical activated product in the hulls and end-pieces and is mainly generated by the {sup 14}N(n,p){sup 14}C reaction. In the previous safety assessment of the TRU waste in Japan, the radionuclides inventory was calculated by ORIGEN-2 code. Some conservative assumptions and preliminary estimates were used in this calculation. For example, total radionuclides generated from a single type of fuel assembly (45 GWd/tU for a PWR unit), and the thickness of the Zircaloy oxide film on the hulls (80 μm) were both overestimated. The second assumption in particular has a large effect on exposure dose evaluation. Therefore, it is essential to have a realistic source term evaluation regarding such items as the C-14 inventory and its distribution to waste parts. In the present study, a C-14 inventory of the hull and end-piece wastes from the operation of a commercial reprocessing plant in Japan corresponding to 32,000 tU (16,000 tU in each BWR and PWR) was calculated. Analysis using individual irradiation conditions and fuel characteristics was conducted on 6 types of fuel assemblies for BWRs and 12 types for PWRs (4 pile types x 3 burnup limits). The oxide film thickness data for each fuel type cladding were obtained from the published literature. Activation calculations were performed by using ORIGEN-2 code. For the amount of spent assembly and other waste characteristics, representative values were assumed based on the published literature. As a preliminary experiment, C-14 in irradiated BWR claddings was measured and found to be consistent with the

  20. DOE mixed waste treatment capacity analysis

    SciTech Connect

    Ross, W.A.; Wehrman, R.R.; Young, J.R.; Shaver, S.R.

    1994-06-01

    This initial DOE-wide analysis compares the reported national capacity for treatment of mixed wastes with the calculated need for treatment capacity based on both a full treatment of mixed low-level and transuranic wastes to the Land Disposal Restrictions and on treatment of transuranic wastes to the WIPP waste acceptance criteria. The status of treatment capacity is reported based on a fifty-element matrix of radiation-handling requirements and functional treatment technology categories. The report defines the classifications for the assessment, describes the models used for the calculations, provides results from the analysis, and includes appendices of the waste treatment facilities data and the waste stream data used in the analysis.

  1. Investigation of the possibility of using hydrogranulation in reprocessing radioactive wastes of radiochemical production facilities

    SciTech Connect

    Revyakin, V.; Borisov, L.M.

    1996-05-01

    Radio-chemical production facilities are constantly accumulating liquid radioactive wastes (still residues as the result of evaporation of extraction and adsorption solutions etc.) which are a complex multicomponent mixtures. The wastes are frequently stored for extended periods of time while awaiting disposition and in some cases, and this is much worse, they are released into the environment. In this report, I would like to draw your attention to some results we have obtained from investigations aimed at simplifying handing of such wastes by the precipitation of hard to dissolve metal hydroxides, the flocculation of the above into granules with the help of surface-active agents (in this case a polyacrylamide - PAA), quickly precipitated and easily filtered. The precipitate may be quickly dried and calcinated, if necessary, and transformed into a dense oxide sinter. In other words it may be transformed into a material convenient for storage or burial.

  2. Reprocessing of dental instruments in washer-disinfectors: does a representative test soil exist in dentistry?

    PubMed

    Franz, Alexander; Bristela, Margit; Stauffer, Fritz

    2012-01-01

    Reprocessing of medical devices, being classified as semi-critical B is recommended to be performed in a washer-disinfector. In order to estimate, whether the expected contaminants of the various medical disciplines can be effectively removed by this washer-disinfector, different so called "test soils" have been proposed to be tested as a marker of cleaning efficacy of the disinfector. Todays described test soils are optimised for the testing of contaminations occurring in surgical procedures, but not for dental procedures. In this study the test soils being proposed in the EN 15883-5 (e.g. KMNE soil, recipe by Koller and coagulated sheep's blood) were compared with 8 reference substances used in the conservative-prosthetic dental practice. The success of the cleaning efficacy in the washer-disinfector was checked visually and by determining the residual protein concentration on the contaminated instruments after the cleaning procedure. It could be shown that in contrast to the proposed test soils of the EN 15883-5, the used reference substances of the dental practice could not be removed by the washer-disinfector. Removal of these reference substances was only possible after manual or ultrasonic cleaning. Since blood plays a subordinate role as a contaminant of instruments during conservative-prosthetic dental treatments, testing of the cleaning efficacy of the washer-disinfector with test soils according to the proposals of the EN 15883-5 is not representative in this discipline of dentistry. Most of the materials used in dental practice can only be removed manually or with the help of the ultrasound bath.

  3. DEVELOPMENT OF CRYSTALLINE CERAMICS FOR IMMOBILIZATION OF ADVANCED FUEL CYCLE REPROCESSING WASTES

    SciTech Connect

    Fox, K.; Brinkman, K.

    2011-09-22

    The Savannah River National Laboratory (SRNL) is developing crystalline ceramic waste forms to incorporate CS/LN/TM high Mo waste streams consisting of perovskite, hollandite, pyrochlore, zirconolite, and powellite phase assemblages. Simple raw materials, including Al{sub 2}O{sub 3}, CaO, and TiO{sub 2} were combined with simulated waste components to produce multiphase crystalline ceramics. Fiscal Year 2011 (FY11) activities included (i) expanding the compositional range by varying waste loading and fabrication of compositions rich in TiO{sub 2}, (ii) exploring the processing parameters of ceramics produced by the melt and crystallize process, (iii) synthesis and characterization of select individual phases of powellite and hollandite that are the target hosts for radionuclides of Mo, Cs, and Rb, and (iv) evaluating the durability and radiation stability of single and multi-phase ceramic waste forms. Two fabrication methods, including melting and crystallizing, and pressing and sintering, were used with the intent of studying phase evolution under various sintering conditions. An analysis of the XRD and SEM/EDS results indicates that the targeted crystalline phases of the FY11 compositions consisting of pyrochlore, perovskite, hollandite, zirconolite, and powellite were formed by both press and sinter and melt and crystallize processing methods. An evaluation of crystalline phase formation versus melt processing conditions revealed that hollandite, perovskite, zirconolite, and residual TiO{sub 2} phases formed regardless of cooling rate, demonstrating the robust nature of this process for crystalline phase development. The multiphase ceramic composition CSLNTM-06 demonstrated good resistance to proton beam irradiation. Electron irradiation studies on the single phase CaMoO{sub 4} (a component of the multiphase waste form) suggested that this material exhibits stability to 1000 years at anticipated self-irradiation doses (2 x 10{sup 10}-2 x 10{sup 11} Gy), but that

  4. Recent Progress in DOE Waste Tank Closure

    SciTech Connect

    Langton, Ch.H.; Cook, J.R.

    2008-07-01

    The US DOE complex currently has over 330 underground storage tanks that have been used to process and store radioactive waste generated from the production of weapons materials. These tanks contain over 380 million liters of high-level and low-level radioactive waste. The waste consists of radioactively contaminated sludge, supernate, salt cake or calcine. Most of the waste exists at four US DOE locations, the Hanford Site, the Savannah River Site, the Idaho Nuclear Technology and Engineering Center and the West Valley Demonstration Project. A summary of the DOE tank closure activities was first issued in 2001. Since then, regulatory changes have taken place that affect some of the sites and considerable progress has been made in closing tanks. This paper presents an overview of the current regulatory changes and drivers and a summary of the progress in tank closures at the various sites over the intervening six years. A number of areas are addressed including closure strategies, characterization of bulk waste and residual heel material, waste removal technologies for bulk waste, heel residuals and annuli, tank fill materials, closure system modeling and performance assessment programs, lessons learned, and external reviews. (authors)

  5. Reprocessing and reuse of waste tire rubber to solve air-quality related problems

    USGS Publications Warehouse

    Lehmann, C.M.B.; Rostam-Abadi, M.; Rood, M.J.; Sun, Jielun

    1998-01-01

    There is a potential for using waste tire rubber to make activated-carbon adsorbents for air-quality control applications. Such an approach provides a recycling path for waste tires and the production of new adsorbents from a low-cost waste material. Tire-derived activated carbons (TDACs) were prepared from waste tires. The resulting products are generally mesoporous, with N2-BET specific surface areas ranging from 239 to 1031 m2/g. TDACs were tested for their ability to store natural gas and remove organic compounds and mercury species from gas streams. TDACs are able to achieve 36% of the recommended adsorbed natural gas (methane) storage capacity for natural-gas-fueled vehicles. Equilibrium adsorption capacities for CH4 achieved by TDACs are comparable to Calgon BPL, a commercially available activated-carbon adsorbent. The acetone adsorption capacity for a TDAC is 67% of the adsorption capacity achieved by BPL at 1 vol % acetone. Adsorption capacities of mercury in simulated flue-gas streams are, in general, larger than adsorption capacities achieved by coal-derived activated carbons (CDACs) and BPL. Although TDACs may not perform as well as commercial adsorbents in some air pollution control applications, the potential lower cost of TDACS should be considered when evaluating economics.

  6. Methodology of Qualification of CCIM Vitrification Process Applied to the High- Level Liquid Waste from Reprocessed Oxide Fuels - 12438

    SciTech Connect

    Lemonnier, S.; Labe, V.; Ledoux, A.; Nonnet, H.; Godon, N.

    2012-07-01

    The vitrification of high-level liquid waste from reprocessed oxide fuels (UOX fuels) by Cold Crucible Induction Melter is planed by AREVA in 2013 in a production line of the R7 facility at La Hague plant. Therefore, the switch of the vitrification technology from the Joule Heated Metal Melter required a complete process qualification study. It involves three specialties, namely the matrix formulation, the glass long-term behavior and the vitrification process development on full-scale pilot. A new glass frit has been elaborated in order to adapt the redox properties and the thermal conductivity of the glass suitable for being vitrified with the Cold Crucible Induction Melter. The role of cobalt oxide on the long term behavior of the glass has been described in the range of the tested concentrations. Concerning the process qualification, the nominal tests, the sensitivity tests and the study of the transient modes allowed to define the nominal operating conditions. Degraded operating conditions tests allowed to identify means of detecting incidents leading to these conditions and allowed to define the procedures to preserve the process equipments protection and the material quality. Finally, the endurance test validated the nominal operating conditions over an extended time period. This global study allowed to draft the package qualification file. The qualification file of the UOX package is currently under approval by the French Nuclear Safety Authority. (authors)

  7. 76 FR 13605 - Notice of Availability of Draft Waste Incidental to Reprocessing Evaluation for the Vitrification...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-14

    ... Web site listed above. SUPPLEMENTARY INFORMATION: The vitrification melter is a box structure, approximately 10 feet on each side, with a stainless steel outer structure and an interior lined with refractory..., pursuant to DOE's authority under the Atomic Energy Act of 1954, as amended, and in accordance with...

  8. On the possibility of reprocessing spent nuclear fuel and radioactive waste by plasma methods

    NASA Astrophysics Data System (ADS)

    Vorona, N. A.; Gavrikov, A. V.; Samokhin, A. A.; Smirnov, V. P.; Khomyakov, Yu. S.

    2015-12-01

    The concept of plasma separation of spent nuclear fuel and radioactive waste is presented. An approach that is based on using an accelerating potential to overcome the energy and angular spread of plasma ions at the separation region inlet and utilizing a potential well to separate spatially the ions of different masses is proposed. It is demonstrated that such separation may be performed at distances of about 1 m with electrical potentials of about 1 kV and a magnetic field of about 1 kG. The estimates of energy consumption and performance of the plasma separation method are presented. These estimates illustrate its potential for technological application. The results of development and construction of an experimental setup for testing the method of plasma separation are presented.

  9. On the possibility of reprocessing spent nuclear fuel and radioactive waste by plasma methods

    SciTech Connect

    Vorona, N. A.; Gavrikov, A. V. Samokhin, A. A.; Smirnov, V. P.; Khomyakov, Yu. S.

    2015-12-15

    The concept of plasma separation of spent nuclear fuel and radioactive waste is presented. An approach that is based on using an accelerating potential to overcome the energy and angular spread of plasma ions at the separation region inlet and utilizing a potential well to separate spatially the ions of different masses is proposed. It is demonstrated that such separation may be performed at distances of about 1 m with electrical potentials of about 1 kV and a magnetic field of about 1 kG. The estimates of energy consumption and performance of the plasma separation method are presented. These estimates illustrate its potential for technological application. The results of development and construction of an experimental setup for testing the method of plasma separation are presented.

  10. Glass ceramic obtained by tailings and tin mine waste reprocessing from Llallagua, Bolivia

    NASA Astrophysics Data System (ADS)

    Arancibia, Jony Roger Hans; Villarino, Cecilia; Alfonso, Pura; Garcia-Valles, Maite; Martinez, Salvador; Parcerisa, David

    2014-05-01

    In Bolivia Sn mining activity produces large tailings of SiO2-rich residues. These tailings contain potentially toxic elements that can be removed into the surface water and produce a high environmental pollution. This study determines the thermal behaviour and the viability of the manufacture of glass-ceramics from glass. The glass has been obtained from raw materials representative of the Sn mining activities from Llallagua (Bolivia). Temperatures of maximum nucleation rate (Tn) and crystallization (Tcr) were calculated from the differential thermal analyses. The final mineral phases were determined by X-ray diffraction and textures were observed by scanning electron microscopy. Crystalline phases are nefeline occurring with wollastonite or plagioclase. Tn for nepheline is between 680 ºC and 700 ºC, for wollastonite, 730 ºC and for plagioclase, 740 ºC. Tcr for nefeline is between 837 and 965 ºC; for wollastonite, 807 ºC and for plagioclase, 977 ºC. In order to establish the mechanical characteristics and efficiency of the vitrification process in the fixation of potentially toxic elements the resistance to leaching and micro-hardness were determined. The obtained contents of the elements leached from the glass ceramic are well below the limits established by the European legislation. So, these analyses confirm that potentially toxic elements remain fixed in the structure of mineral phases formed in the glass-ceramic process. Regarding the values of micro-hardness results show that they are above those of a commercial glass. The manufacture of glass-ceramics from mining waste reduces the volume of tailings produced for the mining industry and, in turn enhances the waste, transforming it into a product with industrial application. Acknowledgements: This work was partly financed by the project AECID: A3/042750/11, and the SGR 2009SGR-00444.

  11. EPA's Review of DOE's Inventory Tracking for TRU Wastes at Waste Control Specialists

    EPA Pesticide Factsheets

    On April 9, 2014, EPA's Waste Isolation Pilot Plant (WIPP) waste characterization team visited Waste Control Specialists (WCS) to determine whether DOE was meeting EPA's waste inventory tracking requirements at 40 CFR 194.24(c)(4).

  12. Re-Processing of Mining Waste: An Alternative Way to Secure Metal Supplies of European Union

    NASA Astrophysics Data System (ADS)

    Guézennec, Anne-Gwénaelle; Bodénan, Françoise; Bertrand, Guillaume; Fuentes, Annabelle; Bellenfant, Gael; Lemière, Bruno; d'Hugues, Patrick; Cassard, Daniel; Save, Maurice

    In France, a recently started project handled by the French geological survey (BRGM) is aimed at identifying interesting old mining wastes deposits at the national level and assessing the metal recovery potential of these dumps. By crossing several databases and information from BRGM archives, 95 old mining sites with sizeable tailings dumps were identified. Selection criteria used to draw up this list were chosen mainly on the basis of the "Criticality Report" compiled for the European Commission in 2010, in which 14 mineral raw materials — 12 critical metals- have been explicitly named as highly critical for the industrial development of the European Union. In most of these mines which date back hundreds of years or more, only a single or at best a couple of metals were extracted with processes whose performances were considerably lower than those used today. Knowing the type of ore commodities and the processes characteristics, it has been thus possible to assess the presence of valuable elements for each tailings dump. From this list an Ag-Pb French abandoned mine has then been selected as a case study to evaluate the potential of extraction of metals still remaining in the tailings with special focus on Ag and Sb. A global site characterization methodology is proposed which can be extrapolated to other sites according to key parameters.

  13. RECENT PROGRESS IN DOE WASTE TANK CLOSURE

    SciTech Connect

    Langton, C

    2008-02-01

    The USDOE complex currently has over 330 underground storage tanks that have been used to process and store radioactive waste generated from the production of weapons materials. These tanks contain over 380 million liters of high-level and low-level radioactive waste. The waste consists of radioactively contaminated sludge, supernate, salt cake or calcine. Most of the waste exists at four USDOE locations, the Hanford Site, the Savannah River Site, the Idaho Nuclear Technology and Engineering Center and the West Valley Demonstration Project. A summary of the DOE tank closure activities was first issued in 2001. Since then, regulatory changes have taken place that affect some of the sites and considerable progress has been made in closing tanks. This paper presents an overview of the current regulatory changes and drivers and a summary of the progress in tank closures at the various sites over the intervening six years. A number of areas are addressed including closure strategies, characterization of bulk waste and residual heel material, waste removal technologies for bulk waste, heel residuals and annuli, tank fill materials, closure system modeling and performance assessment programs, lessons learned, and external reviews.

  14. The Thermal Oxide Reprocessing Plant at Sellafield - Lessons Learned from 10 Years of Hot Operations and their Applicability to the DOE Environmental Management Program

    SciTech Connect

    Burrows, C.; Phillips, C.; Milliken, A.

    2006-07-01

    The Thermal Oxide Reprocessing Plant (Thorp) at Sellafield in northwest England is a $4 billion integrated plant that takes irradiated fuel from worldwide Light Water Reactors and UK Gas-Cooled Reactors and separates the uranium and plutonium from the fission products so that the latter can be vitrified and safely stored. The uranium and plutonium are further separated so that the uranium can be recycled as new reactor fuel, either by itself or in combination with the plutonium as Mixed Oxide (MOX) fuel. Thorp concentrates in excess of 99% of the radioactivity in the irradiated fuel into the vitrified waste product and produces a 40-fold reduction in high active waste volume to be stored, in comparison with direct disposal of the fuel. Thorp incorporates a range of design and operational principles developed over the 50 year history of the Sellafield site. These include the extensive use of no-moving part, no-maintenance equipment in contact with radioactive material, located in shielded 'dark cells' where entry is not expected through the life of the plant. These cells are nevertheless provided with comprehensive secondary containment, instruments, wash-down and recovery systems, and access for cameras and potentially also for remote repair equipment. These arrangements were found to be required during an unusual incident identified in early 2005 when approximately 83 m{sup 3} of highly active liquid escaped from primary tank containment into the secondary containment formed by the dark cell and its stainless steel liner. Although a serious incident, the secondary containment and the in-cell design provisions worked exactly as designed and the liquid was recovered by mid 2005 with no releases to the environment and no member of the workforce or public affected. The ability to access and clean dark cells means that repairs are possible and the large capital asset is thus protected. The enquiry that followed this incident identified issues with carry-through of late

  15. Risks of nuclear fuel reprocessing

    SciTech Connect

    Durant, W.S.

    1990-01-01

    The Savannah River Site's primary function is the production of weapons materials. It consists of four reactors, two fuel reprocessing facilities, a fuel fabrication facility, a nuclear fuel facility for the Navy and a heavy water recycle facility. Under construction is a facility to convert the site's liquid wastes into borosilicate glass. The topic of this paper is risks of nuclear fuel reprocessing. Also discussed are facility operations. 18 figs.

  16. Waste container fabrication from recycled DOE metal

    SciTech Connect

    Motl, G.P.; Burns, D.D.

    1994-02-15

    The Department of Energy (DOE) has more than 2.5 million tons of radioactive scrap metal (RSM) that is either in inventory or expected to be generated over the next 25 years as major facilities within the weapons complex are decommissioned. Much of this material cannot be surface decontaminated. In an attempt to conserve natural resources and to avoid burial of this material at DOE disposal sites, options are now being explored to {open_quotes}beneficially reuse{close_quotes} this material in applications where small amounts of radioactivity are not a detriment. One example is where RSM is currently being beneficially used to fabricate shield blocks for use in DOE medium energy physics programs. This paper describes other initiatives now underway within DOE to utilize RSM to fabricate other products, such as radioactive waste shipping, storage and disposal containers.

  17. Reprocessing of Shallow Seismic Reflection Data to Image Faults Near a Hazardous Waste Site on the Oak Ridge Reservation, Tennessee

    SciTech Connect

    Doll, W.E.

    1997-12-31

    Shallow seismic reflection data from Bear Creek Valley on the Oak Ridge Reservation demonstrates that spectral balancing and tomographic refraction statics can be important processing tools for shallow seismic data. At this site, reprocessing of data which had previously yielded no usable CMP stacked sections was successful after application of these processing techniques.

  18. REPROCESSING OF SHALLOW SEISMIC REFLECTION DATA TO IMAGE FAULTS NEAR A HAZARDOUS WASTE SITE ON THE OAK RIDGE RESERVATION, TENNESSEE

    SciTech Connect

    DOLL, W.E.

    1997-12-30

    Shallow seismic reflection data from Bear Creek Valley on the Oak Ridge Reservation demonstrates that spectral balancing and tomographic refraction statics can be important processing tools for shallow seismic data. At this site, reprocessing of data which had previously yielded no useable CMP stacked sections was successful after application of these processing techniques.

  19. Nuclear waste: Quarterly report on DOE`s nuclear waste program as of March 31, 1988

    SciTech Connect

    1988-12-31

    As part of the Department of Energy`s implementation of the Nuclear Waste Policy Act of 1982, DOE is required to investigate a site at Yucca Mountain, Nevada and, if it determines that the site is suitable, recommend to the President its selection for a nuclear waste repository. The Nuclear Regulatory Commission, in considering development of the plan, issued five objections, one of which is DOE`s failure to recognize the range of alternative conceptual models of the Yucca Mountain site that can be supported by the limited existing technical data. At the end of the quarter DOE directed its project offices in Washington and Texas to begin an orderly phase-out of all site-specific repository activities. Costs for this phase-out are $53 million of the Deaf Smith site and $85 million for the Hanford site.

  20. DOE acceptance of commercial mixed waste -- Studies are under way

    SciTech Connect

    Plummer, T.L.; Owens, C.M.

    1993-03-01

    The topic of the Department of Energy acceptance of commercial mixed waste at DOE facilities has been proposed by host States and compact regions that are developing low-level radioactive waste disposal facilities. States support the idea of DOE accepting commercial mixed waste because (a) very little commercial mixed waste is generated compared to generation by DOE facilities (Department of Energy--26,300 cubic meters annually vs. commercial--3400 cubic meters annually); (b) estimated costs for commercial disposal are estimated to be $15,000 to $40,000 per cubic foot; (c) once treatment capability becomes available, 70% of the current levels of commercial mixed waste will be eliminated, (d) some State laws prohibit the development of mixed waste disposal facilities in their States; (e) DOE is developing a nationwide strategy that will include treatment and disposal capacity for its own mixed waste and the incremental burden on the DOE facilities would be minuscule, and (6) no States are developing mixed waste disposal facilities. DOE senior management has repeatedly expressed willingness to consider investigating the feasibility of DOE accepting commercial mixed waste. In January 1991, Leo Duffy of the Department of energy met with members of the Low-Level Radioactive Waste Forum, which led to an agreement to explore such an arrangement. He stated that this seems like a cost-effective way to solve commercial mixed waste management problems.

  1. DOE methods for evaluating environmental and waste management samples.

    SciTech Connect

    Goheen, S C; McCulloch, M; Thomas, B L; Riley, R G; Sklarew, D S; Mong, G M; Fadeff, S K

    1994-04-01

    DOE Methods for Evaluating Environmental and Waste Management Samples (DOE Methods) provides applicable methods in use by. the US Department of Energy (DOE) laboratories for sampling and analyzing constituents of waste and environmental samples. The development of DOE Methods is supported by the Laboratory Management Division (LMD) of the DOE. This document contains chapters and methods that are proposed for use in evaluating components of DOE environmental and waste management samples. DOE Methods is a resource intended to support sampling and analytical activities that will aid in defining the type and breadth of contamination and thus determine the extent of environmental restoration or waste management actions needed, as defined by the DOE, the US Environmental Protection Agency (EPA), or others.

  2. Nuclear waste: Status of DOE`s nuclear waste site characterization activities

    SciTech Connect

    1987-12-31

    Three potential nuclear waste repository sites have been selected to carry out characterization activities-the detailed geological testing to determine the suitability of each site as a repository. The sites are Hanford in south-central Washington State, Yucca Mountain in southern Nevada, and Deaf Smith in the Texas Panhandle. Two key issues affecting the total program are the estimations of the site characterization completion data and costs and DOE`s relationship with the Nuclear Regulatory Commission which has been limited and its relations with affected states and Indian tribes which continue to be difficult.

  3. A glass-encapsulated calcium phosphate wasteform for the immobilization of actinide-, fluoride-, and chloride-containing radioactive wastes from the pyrochemical reprocessing of plutonium metal

    NASA Astrophysics Data System (ADS)

    Donald, I. W.; Metcalfe, B. L.; Fong, S. K.; Gerrard, L. A.; Strachan, D. M.; Scheele, R. D.

    2007-03-01

    Chloride-containing radioactive wastes are generated during the pyrochemical reprocessing of Pu metal. Immobilization of these wastes in borosilicate glass or Synroc-type ceramics is not feasible due to the very low solubility of chlorides in these hosts. Alternative candidates have therefore been sought including phosphate-based glasses, crystalline ceramics and hybrid glass/ceramic systems. These studies have shown that high losses of chloride or evolution of chlorine gas from the melt make vitrification an unacceptable solution unless suitable off-gas treatment facilities capable of dealing with these corrosive by-products are available. On the other hand, both sodium aluminosilicate and calcium phosphate ceramics are capable of retaining chloride in stable mineral phases, which include sodalite, Na 8(AlSiO 4) 6Cl 2, chlorapatite, Ca 5(PO 4) 3Cl, and spodiosite, Ca 2(PO 4)Cl. The immobilization process developed in this study involves a solid state process in which waste and precursor powders are mixed and reacted in air at temperatures in the range 700-800 °C. The ceramic products are non-hygroscopic free-flowing powders that only require encapsulation in a relatively low melting temperature phosphate-based glass to produce a monolithic wasteform suitable for storage and ultimate disposal.

  4. Geohydrologic conditions at the nuclear-fuels reprocessing plant and waste-management facilities at the Western New York Nuclear Service Center, Cattaraugus County, New York

    USGS Publications Warehouse

    Bergeron, M.P.; Kappel, W.M.; Yager, R.M.

    1987-01-01

    A nuclear-fuel reprocessing plant, a high-level radioactive liquid-waste tank complex, and related waste facilities occupy 100 hectares (ha) within the Western New York Nuclear Service Center near West Valley, N.Y. The facilities are underlain by glacial and postglacial deposits that fill an ancestrial bedrock valley. The main plant facilities are on an elevated plateau referred to as the north plateau. Groundwater on the north plateau moves laterally within a surficial sand and gravel from the main plant building to areas northeast, east, and southeast of the facilities. The sand and gravel ranges from 1 to 10 m thick and has a hydraulic conductivity ranging from 0.1 to 7.9 m/day. Two separate burial grounds, a 4-ha area for low-level radioactive waste disposal and a 2.9-ha area for disposal of higher-level waste are excavated into a clay-rich till that ranges from 22 to 28 m thick. Migration of an organic solvent from the area of higher level waste at shallow depth in the till suggests that a shallow, fractured, oxidized, and weathered till is a significant pathway for lateral movement of groundwater. Below this zone, groundwater moves vertically downward through the till to recharge a lacustrine silt and fine sand. Within the saturated parts of the lacustrine unit, groundwater moves laterally to the northeast toward Buttermilk Creek. Hydraulic conductivity of the till, based on field and laboratory analyses , ranges from 0.000018 to 0.000086 m/day. (USGS)

  5. Hazardous waste shipment data collection from DOE sites

    SciTech Connect

    Page, L.A.; Kirkpatrick, T.D.; Stevens, L.

    1992-12-31

    Past practices at the US Department of Energy (DOE) sites for offsite release of hazardous waste are being reviewed to determine if radioactively contaminated hazardous wastes were released to commercial treatment, storage, and disposal facilities. Records indicating the presence of radioactivity in waste shipped to and treated at a commercial incineration facility led to a ban on offsite hazardous waste shipments and investigation of past practices for offsite release of hazardous waste from the DOE sites. A House of Representatives Interior and Insular Affairs Committee oversight hearing on potentially contaminated waste shipments to commercial facilities concluded that the main issue was the lack of a uniform national standard to govern disposal of mixed waste.

  6. Hazardous waste shipment data collection from DOE sites

    SciTech Connect

    Page, L.A.; Kirkpatrick, T.D. ); Stevens, L. )

    1992-01-01

    Past practices at the US Department of Energy (DOE) sites for offsite release of hazardous waste are being reviewed to determine if radioactively contaminated hazardous wastes were released to commercial treatment, storage, and disposal facilities. Records indicating the presence of radioactivity in waste shipped to and treated at a commercial incineration facility led to a ban on offsite hazardous waste shipments and investigation of past practices for offsite release of hazardous waste from the DOE sites. A House of Representatives Interior and Insular Affairs Committee oversight hearing on potentially contaminated waste shipments to commercial facilities concluded that the main issue was the lack of a uniform national standard to govern disposal of mixed waste.

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

    SciTech Connect

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

    2012-07-01

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

  8. Airborne waste management technology applicable for use in reprocessing plants for control of iodine and other off-gas constituents

    SciTech Connect

    Jubin, R.T.

    1988-02-01

    Extensive work in the area of iodine removal from reprocessing plant off-gas streams using various types of solid sorbent materials has been conducted worldwide over the past two decades. This work has focused on the use of carbon filters, primarily for power plant applications. More recently, the use of silver-containing sorbents has been the subject of considerable research. The most recent work in the United States has addressed the use of silver-exchanged faujasites and mordenites. The chemical reactions of iodine with silver on the sorbent are not well defined, but it is generally believed that chemisorbed iodides and iodates are formed. The process for iodine recovery generally involves passage of the iodine-laden gas stream through a packed bed of the adsorbent material preheated to a temperature of about 150/degree/C. Most iodine removal system designs utilizing silver-containing solid sorbents assume only a 30 to 50% silver utilization. Based on laboratory tests, potentially 60 to 70% of the silver contained in the sorbents can be reacted with iodine. To overcome the high cost of silver associated with these materials, various approaches have been explored. Among these are the regeneration of the silver-containing sorbent by stripping the iodine and trapping the iodine on a sorbent that has undergone only partial silver exchange and is capable of attaining a much higher silver utilization. This summary report describes the US work in regeneration of iodine-loaded solid sorbent material. In addition, the report discusses the broader subject of plant off-gas treatment including system design. The off-gas technologies to recovery No/sub x/ and to recover and dispose of Kr, /sup 14/C, and I are described as to their impacts on the design of an integrated off-gas system. The effect of ventilation philosophy for the reprocessing plant is discussed as an integral part of the overall treatment philosophy of the plant off-gas. 103 refs., 5 figs., 8 tabs.

  9. Iron Phosphate Glasses for Vitrifying DOE High Priority Nuclear Wastes

    SciTech Connect

    Kim, C.W.; Day, D.E.

    2004-03-29

    Iron phosphate glasses have been studied as an alternative glass for vitrifying Department of Energy (DOE) high priority wastes. The high priority wastes were the Low Activity Waste (LAW) and the High Level Waste (HLW) with high chrome content stored at Hanford, WA, and the Sodium Bearing Waste (SBW) stored at the Idaho National Engineering and Environmental Laboratory. These wastes were recommended by Tanks Focus Area since they were expected to require special attention when vitrified in borosilicate glasses. All three of these wastes have been successfully vitrified in iron phosphate glasses at waste loadings ranging from a low of 32 wt% for the high sulfate LAW to 40 wt% for the SBW to a high of 75 wt% for the high chrome HLW. In addition to these desirable high waste loadings, the iron phosphate glasses were easily melted, typically between 950 and 1200 C, in less than 4 hours in commercial refractory oxide containers. It is noteworthy that the chemical durability of both glassy and deliberately crystallized iron phosphate wasteforms not only met, but significantly exceeded, all current DOE chemical durability requirements as measured by the Product Consistency Test (PCT) and Vapor Hydration Test (VHT). The high waste loading, low melting temperature, rapid furnace throughput (short melting time) and their outstanding chemical durability could significantly accelerate the clean up effort and reduce the time and cost of vitrifying these high priority wastes.

  10. EPA and DOE to Resolve Hanford Hazardous Waste Violations

    EPA Pesticide Factsheets

    (Seattle, WA - January 28, 2015) The U.S. Environmental Protection Agency (EPA) and the U.S. Department of Energy (US DOE) have resolved alleged violations of hazardous waste requirements at the Hanford Site near Richland, Washington.

  11. EPA/DOE joint efforts on mixed waste treatment

    SciTech Connect

    Lee, C.C.; Huffman, G.L.; Nalesnik, R.P.

    1995-12-31

    Under the requirements of the Federal Facility Compliance Act (FFCA), the Department of Energy (DOE) is directed to develop treatment plans for their stockpile of wastes generated at their various sites. As a result, DOE is facing the monumental problem associated with the treatment and ultimate disposal of their mixed (radioactive and hazardous) waste. Meanwhile, the Environmental Protection Agency (EPA) issued a final {open_quotes}Hazardous Waste Combustion Strategy{close_quotes} in November 1994. Under the Combustion Strategy, EPA permit writers have been given the authority to use the Omnibus Provision of the Resource Conservation and Recovery Act (RCRA) to impose more stringent emission limits for waste combustors prior to the development of new regulations. EPA and DOE established a multi-year Interagency Agreement (IAG) in 1991. The main objective of the IAG (and of the second IAG that was added in 1993) is to conduct a research program on thermal technologies for treating mixed waste and to establish permit procedures for these technologies particularly under the new requirements of the above-mentioned EPA Combustion Strategy. The objective of this Paper is to summarize the results of the EPA/DOE joint efforts on mixed waste treatment since the establishment of the original Interagency Agreement. Specifically, this Paper will discuss six activities that have been underway; namely: (1) National Technical Workgroup (NTW) on Mixed Waste Treatment, (2) State-of-the-Art Assessment of APC (Air Pollution Control) and Monitoring Technologies for the Rocky Flats Fluidized Bed Unit, (3) Initial Study of Permit {open_quotes}Roadmap{close_quotes} Development for Mixed Waste Treatment, (4) Risk Assessment Approach for a Mixed Waste Thermal Treatment Facility, (5) Development and Application of Technology Selection Criteria for Mixed Waste Thermal Treatment, and (6) Performance Testing of Mixed Waste Incineration: In-Situ Chlorine Capture in a Fluidized Bed Unit.

  12. DOE report assesses environmental impact of waste oil industry

    SciTech Connect

    Not Available

    1987-07-01

    Most current methods of used oil and unused waste oil utilization and disposal create risks of contaminating air, water, or soil with substances that pose hazards to human, animal, and plant life. Recent actions taken to regulate used oil may create severe constraints on those who generate, collect, and handle used oil, such that many of them may leave the market. This may lead to decreased availability of sound disposal options resulting in increased improper disposal of used oil. The U.S. EPA has tried several times to classify used oil as a hazardous waste. The rationale for proposing such a regulation was that its implementation would force more energy recovery through fuel reprocessing and lube oil re-refining. The release of 61-128 million gal/year of used oil into the environment would likely threaten ground and surface waters with oil contamination, thereby endangering drinking water supplies and aquatic life. Therefore, in its latest attempt to classify used oil as a hazardous waste, EPA concluded that such a listing would discourage recycling or reuse.

  13. R D activities at DOE applicable to mixed waste

    SciTech Connect

    Erickson, M.D.; Devgun, J.S.; Brown, J.J.; Beskid, N.J.

    1991-01-01

    The Department of Energy (DOE) has established the Office of Environmental Restoration and Waste Management. Within the new organization, the Office of Technology Development (OTD) is responsible for research, development, demonstration, testing and evaluation (RDDT E) activities aimed at meeting DOE cleanup goals, while minimizing cost and risk. Because of US governmental activities dating back to the Manhattan project, mixed radioactive and hazardous waste is an area of particular concern to DOE. The OTD is responsible for a number of R D activities aimed at improving capabilities to characterize, control, and properly dispose of mixed waste. These activities and their progress to date will be reviewed. In addition, needs for additional R D on managing mixed waste will be presented. 5 refs., 2 tabs.

  14. Waste Management Planned for the Advanced Fuel Cycle Facility

    SciTech Connect

    Soelberg

    2007-09-01

    The U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) program has been proposed to develop and employ advanced technologies to increase the proliferation resistance of spent nuclear fuels, recover and reuse nuclear fuel resources, and reduce the amount of wastes requiring permanent geological disposal. In the initial GNEP fuel cycle concept, spent nuclear fuel is to be reprocessed to separate re-useable transuranic elements and uranium from waste fission products, for fabricating new fuel for fast reactors. The separated wastes would be converted to robust waste forms for disposal. The Advanced Fuel Cycle Facility (AFCF) is proposed by DOE for developing and demonstrating spent nuclear fuel recycling technologies and systems. The AFCF will include capabilities for receiving and reprocessing spent fuel and fabricating new nuclear fuel from the reprocessed spent fuel. Reprocessing and fuel fabrication activities will generate a variety of radioactive and mixed waste streams. Some of these waste streams are unique and unprecedented. The GNEP vision challenges traditional U.S. radioactive waste policies and regulations. Product and waste streams have been identified during conceptual design. Waste treatment technologies have been proposed based on the characteristics of the waste streams and the expected requirements for the final waste forms. Results of AFCF operations will advance new technologies that will contribute to safe and economical commercial spent fuel reprocessing facilities needed to meet the GNEP vision. As conceptual design work and research and design continues, the waste management strategies for the AFCF are expected to also evolve.

  15. Using benchmarking to minimize common DOE waste streams. Volume 1, Methodology and liquid photographic waste

    SciTech Connect

    Levin, V.

    1994-04-01

    Finding innovative ways to reduce waste streams generated at Department of Energy (DOE) sites by 50% by the year 2000 is a challenge for DOE`s waste minimization efforts. This report examines the usefulness of benchmarking as a waste minimization tool, specifically regarding common waste streams at DOE sites. A team of process experts from a variety of sites, a project leader, and benchmarking consultants completed the project with management support provided by the Waste Minimization Division EM-352. Using a 12-step benchmarking process, the team examined current waste minimization processes for liquid photographic waste used at their sites and used telephone and written questionnaires to find ``best-in-class`` industrv partners willing to share information about their best waste minimization techniques and technologies through a site visit. Eastman Kodak Co., and Johnson Space Center/National Aeronautics and Space Administration (NASA) agreed to be partners. The site visits yielded strategies for source reduction, recycle/recovery of components, regeneration/reuse of solutions, and treatment of residuals, as well as best management practices. An additional benefit of the work was the opportunity for DOE process experts to network and exchange ideas with their peers at similar sites.

  16. NRC Waste Incidental to Reprocessing Program: Overview of Consultation and Monitoring Activities at the Idaho National Laboratory and the Savannah River Site - What We Have Learned - 12470

    SciTech Connect

    Suber, Gregory

    2012-07-01

    In 2005 the U.S. Nuclear Regulatory Commission (NRC) began to implement a new set of responsibilities under the Ronald W. Reagan National Defense Authorization Act (NDAA) of Fiscal Year 2005. Section 3116 of the NDAA requires the U.S. Department of Energy (DOE) to consult with the NRC for certain non-high level waste determinations and also requires NRC to monitor DOE's disposal actions related to those determinations. In Fiscal Year 2005, the NRC staff began consulting with DOE and completed reviews of draft waste determinations for salt waste at the Savannah River Site. In 2006, a second review was completed on tank waste residuals including sodium-bearing waste at the Idaho Nuclear Technology and Engineering Center Tank Farm at the Idaho National Laboratory. Monitoring Plans were developed for these activities and the NRC is actively monitoring disposal actions at both sites. NRC is currently in consultation with DOE on the F-Area Tank Farm closure and anticipates entering consultation on the H-Area Tank Farm at the Savannah River Site. This paper presents, from the NRC perspective, an overview of how the consultation and monitoring process has evolved since its conception in 2005. It addresses changes in methods and procedures used to collect and develop information used by the NRC in developing the technical evaluation report and monitoring plan under consultation and the implementation the plan under monitoring. It will address lessons learned and best practices developed throughout the process. The NDAA has presented significant challenges for the NRC and DOE. Past and current successes demonstrate that the NDAA can achieve its intended goal of facilitating tank closure at DOE legacy defense waste sites. The NRC believes many of the challenges in performing the WD reviews have been identified and addressed. Lessons learned have been collected and documented throughout the review process. Future success will be contingent on each agencies commitment to

  17. IR and Raman Spectroscopy of Sodium-Aluminophosphate Glasses for Immobilizing High-Level Wastes from Spent Nuclear Fuel Reprocessing

    NASA Astrophysics Data System (ADS)

    Stefanovsky, S. V.; Myasoedov, B. F.; Remizov, M. B.; Belanova, E. A.

    2014-09-01

    The structure of sodium-aluminophosphate glasses containing constituents of high-level wastes (cesium, magnesium, copper, and molybdenum oxides) from uranium-graphite reactors was studied by IR and Raman spectroscopy coupled with x-ray diffraction. The structural network was shown to be composed of short P-O chains with embedded AlO4 tetrahedra. Cross-linking by Mg2+ was possible in the Mg-bearing samples. The effect of the other oxides (Cs2O, MoO3, CuO) on the glass structure was negligible for the occurring amounts. The glasses devitrified partially upon quenching and more strongly upon annealing. This was reflected in splitting of the vibrational bands for bonds in the glass anionic structural motif.

  18. Waste separation: Does it influence municipal waste combustor emissions?

    SciTech Connect

    Chandler, A.J.; Rigo, H.G.

    1996-09-01

    It has been suggested that MSW incinerator emissions show significant variations because of the heterogeneous nature of the waste fed to the furnace. This argument has even been used to propose banning certain materials from incinerators. However, data previously reported by the authors suggests that a large portion of the trace metals come from natural sources. Furthermore, full scale incinerator spiking experiments suggest that certain forms of trace metals have minimal effects on stack emissions. Similar studies with chlorinated plastics have failed to identify a significant effect on incinerator dioxin emissions. The implication of segregating the lawn and garden waste and other fines from the furnace feed is explored using data from a 400 tpd mass burn facility equipped with a conditioning tower, dry reactor and fabric filter air pollution control system (APCS) preceded by an NRT separation system. The stack emissions have been tested periodically since commissioning to characterize emissions for various seasons using both processed fuel and raw MSW. Front end processing to remove selected portions of the waste stream based upon size or physical properties, i.e. fines, grass, or ferrous materials, did not result in a statistically significant difference in stack emissions. System operating regime, and in particular those that effect the effective air to cloth ratio in the fabric filter, appear to be the principal influence on emission levels.

  19. DOE`s integrated low-level waste management program and strategic planning

    SciTech Connect

    Duggan, G.; Hwang, J.

    1993-03-01

    To meet the DOE`s commitment to operate its facilities in a safe, economic, and environmentally sound manner, and to comply with all applicable federal, state, and local rules, regulations, and agreements, DOE created the Office of Environmental Restoration and Waste Management (EM) in 1989 to focus efforts on controlling waste management and cleaning up contaminated sites. In the first few years of its existence, the Office of Waste Management (EM-30) has concentrated on operational and corrective activities at the sites. In 1992, the Office of Waste Management began to apply an integrated approach to managing its various waste types. Consequently, DOE established the Low-Level Waste Management Program (LLWMP) to properly manage its complex-wide LLW in a consistent manner. The objective of the LLWMP is to build and operate an integrated, safe, and cost-effective program to meet the needs of waste generators. The program will be based on acceptable risk and sound planning, resulting in public confidence and support. Strategic planning of the program is under way and is expected to take two to three years before implementation of the integrated waste management approach.

  20. DOE methods for evaluating environmental and waste management samples

    SciTech Connect

    Goheen, S.C.; McCulloch, M.; Thomas, B.L.; Riley, R.G.; Sklarew, D.S.; Mong, G.M.; Fadeff, S.K.

    1994-10-01

    DOE Methods for Evaluating Environmental and Waste Management Samples (DOE Methods) is a resource intended to support sampling and analytical activities for the evaluation of environmental and waste management samples from U.S. Department of Energy (DOE) sites. DOE Methods is the result of extensive cooperation from all DOE analytical laboratories. All of these laboratories have contributed key information and provided technical reviews as well as significant moral support leading to the success of this document. DOE Methods is designed to encompass methods for collecting representative samples and for determining the radioisotope activity and organic and inorganic composition of a sample. These determinations will aid in defining the type and breadth of contamination and thus determine the extent of environmental restoration or waste management actions needed, as defined by the DOE, the U.S. Environmental Protection Agency, or others. The development of DOE Methods is supported by the Analytical Services Division of DOE. Unique methods or methods consolidated from similar procedures in the DOE Procedures Database are selected for potential inclusion in this document. Initial selection is based largely on DOE needs and procedure applicability and completeness. Methods appearing in this document are one of two types, {open_quotes}Draft{close_quotes} or {open_quotes}Verified{close_quotes}. {open_quotes}Draft{close_quotes} methods that have been reviewed internally and show potential for eventual verification are included in this document, but they have not been reviewed externally, and their precision and bias may not be known. {open_quotes}Verified{close_quotes} methods in DOE Methods have been reviewed by volunteers from various DOE sites and private corporations. These methods have delineated measures of precision and accuracy.

  1. The behaviour of (137)Cs in the North Atlantic Ocean assessed from numerical modelling: Releases from nuclear fuel reprocessing factories, redissolution from contaminated sediments and leakage from dumped nuclear wastes.

    PubMed

    Periáñez, R; Suh, Kyung-Suk; Min, Byung-Il

    2016-12-15

    A Lagrangian model which simulates the dispersion of (137)Cs in the North Atlantic has been developed. The model includes water/sediment interactions. It has been tested comparing calculated and measured (137)Cs concentrations in water and sediments of the European Shelf resulting after the releases from the nuclear fuel reprocessing plants of Sellafield and La Hague. Some additional numerical experiments have been carried out. First, the redissolution of (137)Cs from contaminated sediments after the reduction in releases from the reprocessing plants has been studied. This allowed to calculate effective half-lives of (137)Cs in several sub-basins. Later, potential leakage of (137)Cs from dumped nuclear wastes in several locations of the Atlantic has been investigated. Even in worst-case scenarios, these leakages should not lead to any radiological implications.

  2. Analysis of waste treatment requirements for DOE mixed wastes: Technical basis

    SciTech Connect

    1995-02-01

    The risks and costs of managing DOE wastes are a direct function of the total quantities of 3wastes that are handled at each step of the management process. As part of the analysis of the management of DOE low-level mixed wastes (LLMW), a reference scheme has been developed for the treatment of these wastes to meet EPA criteria. The treatment analysis in a limited form was also applied to one option for treatment of transuranic wastes. The treatment requirements in all cases analyzed are based on a reference flowsheet which provides high level treatment trains for all LLMW. This report explains the background and basis for that treatment scheme. Reference waste stream chemical compositions and physical properties including densities were established for each stream in the data base. These compositions are used to define the expected behavior for wastes as they pass through the treatment train. Each EPA RCRA waste code was reviewed, the properties, chemical composition, or characteristics which are of importance to waste behavior in treatment were designated. Properties that dictate treatment requirements were then used to develop the treatment trains and identify the unit operations that would be included in these trains. A table was prepared showing a correlation of the waste physical matrix and the waste treatment requirements as a guide to the treatment analysis. The analysis of waste treatment loads is done by assigning wastes to treatment steps which would achieve RCRA compliant treatment. These correlation`s allow one to examine the treatment requirements in a condensed manner and to see that all wastes and contaminant sets are fully considered.

  3. Using benchmarking to minimize common DOE waste streams: Volume 5. Office paper waste

    SciTech Connect

    Levin, V.

    1995-10-01

    Finding innovative ways to reduce waste streams generated at US Department of Energy (DOE) sites by 50% by the year 2000 is a challenge for DOE`s waste minimization efforts. A team composed of members from several DOE facilities used the quality tool known as benchmarking to improve waste minimization efforts. First the team examined office waste generation and handling processes at their sites. Then team members developed telephone and written questionnaires to help identify potential ``best-in-class`` industry partners willing to share information about their best waste minimization techniques and technologies. The team identified two benchmarking partners, NIKE, Inc., in Beaverton, Oregon, and Microsoft, Inc., in Redmond, Washington. Both companies have proactive, employee-driven environmental issues programs. Both companies report strong employee involvement, management commitment, and readily available markets for recyclable materials such as white paper and nonwhite assorted paper. The availability of markets, the initiative and cooperation of employees, and management support are the main enablers for their programs. At both companies, recycling and waste reduction programs often cut across traditional corporate divisions such as procurement, janitorial services, environmental compliance, grounds maintenance, cafeteria operations, surplus sales, and shipping and receiving. These companies exhibited good cooperation between these functions to design and implement recycling and waste reduction programs.

  4. DOE complex buried waste characterization assessment. Buried Waste Integrated Demonstration Program

    SciTech Connect

    Kaae, P.S.; Holter, G.M.; Garrett, S.M.K.

    1993-01-01

    The work described in this report was conducted by Pacific Northwest Laboratory to provide information to the Buried Waste Integrated Demonstration (BWID) program. The information in this report is intended to provide a complex-wide planning base for th.e BWID to ensure that BWID activities are appropriately focused to address the range of remediation problems existing across the US Department of Energy (DOE) complex. This report contains information characterizing the 2.1 million m{sup 3} of buried and stored wastes and their associated sites at six major DOE facilities. Approximately 85% of this waste is low-level waste, with about 12% TRU or TRU mixed waste; the remaining 3% is low-level mixed waste. In addition, the report describes soil contamination sites across the complex. Some of the details that would be useful in further characterizing the buried wastes and contaminated soil sites across the DOE complex are either unavailable or difficult to locate. Several options for accessing this information and/or improving the information that is available are identified in the report. This document is a companion to Technology Needs for Remediation: Hanford and Other DOE Sites, PNL-8328 (Stapp 1993).

  5. High-level waste program integration within the DOE complex

    SciTech Connect

    Valentine, J.H.; Davis, N.R.; Malone, K.; Schaus, P.S.

    1998-03-01

    Eleven major Department of Energy (DOE) site contractors were chartered by the Assistant Secretary to use a systems engineering approach to develop and evaluate technically defensible cost savings opportunities across the complex. Known as the complex-wide Environmental Management Integration (EMI), this process evaluated all the major DOE waste streams including high level waste (HLW). Across the DOE complex, this waste stream has the highest life cycle cost and is scheduled to take until at least 2035 before all HLW is processed for disposal. Technical contract experts from the four DOE sites that manage high level waste participated in the integration analysis: Hanford, Savannah River Site (SRS), Idaho National Engineering and Environmental Laboratory (INEEL), and West Valley Demonstration Project (WVDP). In addition, subject matter experts from the Yucca Mountain Project and the Tanks Focus Area participated in the analysis. Also, departmental representatives from the US Department of Energy Headquarters (DOE-HQ) monitored the analysis and results. Workouts were held throughout the year to develop recommendations to achieve a complex-wide integrated program. From this effort, the HLW Environmental Management (EM) Team identified a set of programmatic and technical opportunities that could result in potential cost savings and avoidance in excess of $18 billion and an accelerated completion of the HLW mission by seven years. The cost savings, schedule improvements, and volume reduction are attributed to a multifaceted HLW treatment disposal strategy which involves waste pretreatment, standardized waste matrices, risk-based retrieval, early development and deployment of a shipping system for glass canisters, and reasonable, low cost tank closure.

  6. [Does cerebral salt wasting syndrome exist?].

    PubMed

    Leblanc, P-E; Cheisson, G; Geeraerts, T; Tazarourte, K; Duranteau, J; Vigué, B

    2007-11-01

    Increased natriuresis is a frequent situation after subarachnoid haemorrhage (SAH). It may be responsible for hyponatremia, which can be dangerous in case of severe hypo-osmolarity or hypovolemia. Inappropriate secretion of antidiuretic hormone or cerebral salt wasting syndrome (CSWS) have been incriminated for hyponatremia after SAH, but it remains difficult to distinguish between both syndromes. There are many explanations for increased natriuresis after SAH, depending on the level of blood pressure, the volemia, and the presence or not of natriuretic peptides. The cerebral insult and the treatments, which are done to fight against elevated intracranial pressure or vasospasm, can modify any of these parameters. So it appears that the word "cerebral" in CSWS is probably not a good term and it would be better to talk about appropriate or non-appropriate natriuretic response. Corticoïds or urea can be useful for controlling hypernatriuresis.

  7. Characteristics of potential repository wastes. Volume 2

    SciTech Connect

    Not Available

    1992-07-01

    The LWR spent fuels discussed in Volume 1 of this report comprise about 99% of all domestic non-reprocessed spent fuel. In this report we discuss other types of spent fuels which, although small in relative quantity, consist of a number of diverse types, sizes, and compositions. Many of these fuels are candidates for repository disposal. Some non-LWR spent fuels are currently reprocessed or are scheduled for reprocessing in DOE facilities at the Savannah River Site, Hanford Site, and the Idaho National Engineering Laboratory. It appears likely that the reprocessing of fuels that have been reprocessed in the past will continue and that the resulting high-level wastes will become part of defense HLW. However, it is not entirely clear in some cases whether a given fuel will be reprocessed, especially in cases where pretreatment may be needed before reprocessing, or where the enrichment is not high enough to make reprocessing attractive. Some fuels may be canistered, while others may require special means of disposal. The major categories covered in this chapter include HTGR spent fuel from the Fort St. Vrain and Peach Bottom-1 reactors, research and test reactor fuels, and miscellaneous fuels, and wastes generated from the decommissioning of facilities.

  8. A DOE manual: DOE methods for evaluating environmental and waste management samples

    SciTech Connect

    Goheen, S.C.; Fadeff, S.K.; Sklarew, D.S.; McCulloch, M.; Mong, G.M.; Riley, R.G.; Thomas, B.L.

    1994-08-01

    DOE Methods for Evaluating Environmental and Waste Management Samples (DOE Methods) is a guidance/methods document supporting environmental restoration (ER) and waste management (WM) (collectively referred to as EM) sampling and analysis activities at US Department of Energy (DOE) sites. DOE Methods is intended to supplement existing guidance documents (e.g., the US Environmental Protection Agency`s Test Methods for Evaluating Solid Waste, SW-846), which apply to low-level or non-radioactive samples, and the complexities of waste and environmental samples encountered at DOE sites. The document contains quality assurance (QA), quality control (QC), safety, sampling, organic analysis, inorganic analysis, and radio-analytical guidance as well as sampling and analytical methods. It is updated every six months (April and October) with additional methods. As of April 1994, DOE methods contained 3 sampling and 39 analytical methods. It is anticipated that between 10 and 20 new methods will be added in October 1994. All methods are either peer reviewed and contain performance data, or are included as draft methods.

  9. Generic Degraded Congiguration Probability Analysis for DOE Codisposal Waste Package

    SciTech Connect

    S.F.A. Deng; M. Saglam; L.J. Gratton

    2001-05-23

    In accordance with the technical work plan, ''Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages'' (CRWMS M&O 2000c), this Analysis/Model Report (AMR) is developed for the purpose of screening out degraded configurations for U.S. Department of Energy (DOE) spent nuclear fuel (SNF) types. It performs the degraded configuration parameter and probability evaluations of the overall methodology specified in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000, Section 3) to qualifying configurations. Degradation analyses are performed to assess realizable parameter ranges and physical regimes for configurations. Probability calculations are then performed for configurations characterized by k{sub eff} in excess of the Critical Limit (CL). The scope of this document is to develop a generic set of screening criteria or models to screen out degraded configurations having potential for exceeding a criticality limit. The developed screening criteria include arguments based on physical/chemical processes and probability calculations and apply to DOE SNF types when codisposed with the high-level waste (HLW) glass inside a waste package. The degradation takes place inside the waste package and is long after repository licensing has expired. The emphasis of this AMR is on degraded configuration screening and the probability analysis is one of the approaches used for screening. The intended use of the model is to apply the developed screening criteria to each DOE SNF type following the completion of the degraded mode criticality analysis internal to the waste package.

  10. Characterization of hazardous waste residuals from Environmental Restoration Program activities at DOE installations: Waste management implications

    SciTech Connect

    Lazaro, M.A.; Esposito, M.P.

    1995-06-01

    Investigators at Argonne National Laboratory (ANL), with support from associates at the Pacific Northwest Laboratory (PNL), have assembled an inventory of the types and volumes of radioactive, toxic or hazardous, and mixed waste likely to be generated over the next 30 years as the US Department of Energy (DOE) implements its nationwide Environmental Restoration (ER) Program. The inventory and related analyses are being considered for integration into DOE`s Programmatic Environmental Impact Statement (PEIS) covering the potential environmental impacts and risks associated with alternative management practices and programs for wastes generated from routine operations. If this happens, the ER-generated waste could be managed under a set of alternatives considered under the PEIS and selected at the end of the current National Environmental Policy Act process.

  11. Development of consistent hazard controls for DOE transuranic waste operations

    SciTech Connect

    Woody, W.J.

    2007-07-01

    This paper describes the results of a re-engineering initiative undertaken with the Department of Energy's (DOE) Office of Environmental Management (EM) in order to standardize hazard analysis assumptions and methods and resulting safety controls applied to multiple transuranic (TRU) waste operations located across the United States. A wide range of safety controls are historically applied to transuranic waste operations, in spite of the fact that these operations have similar operational characteristics and hazard/accident potential. The re-engineering effort supported the development of a DOE technical standard with specific safety controls designated for accidents postulated during waste container retrieval, staging/storage, venting, onsite movements, and characterization activities. Controls cover preventive and mitigative measures; include both hardware and specific administrative controls; and provide protection to the facility worker, onsite co-located workers and the general public located outside of facility boundaries. The Standard development involved participation from all major DOE sites conducting TRU waste operations. Both safety analysts and operations personnel contributed to the re-engineering effort. Acknowledgment is given in particular to the following individuals who formed a core working group: Brenda Hawks, (DOE Oak Ridge Office), Patrice McEahern (CWI-Idaho), Jofu Mishima (Consultant), Louis Restrepo (Omicron), Jay Mullis (DOE-ORO), Mike Hitchler (WSMS), John Menna (WSMS), Jackie East (WSMS), Terry Foppe (CTAC), Carla Mewhinney (WIPP-SNL), Stephie Jennings (WIPP-LANL), Michael Mikolanis (DOESRS), Kraig Wendt (BBWI-Idaho), Lee Roberts (Fluor Hanford), and Jim Blankenhorn (WSRC). Additional acknowledgment is given to Dae Chung (EM) and Ines Triay (EM) for leadership and management of the re-engineering effort. (authors)

  12. Toward a Greenish Nuclear Fuel Cycle: Ionic Liquids as Solvents for Spent Nuclear Fuel Reprocessing and Other Decontamination Processes for Contaminated Metal Waste

    NASA Astrophysics Data System (ADS)

    Straka, Martin

    2016-12-01

    The final disposition of spent nuclear fuel (SNF) is an area that requires innovative solutions. The use of ionic liquids (ILs) has been examined as one means to remediate SNF in a variety of different chemical environments and with different chemical starting materials. The effectiveness of various ILs for SNF reprocessing, as well as the reaction chemistry that occurs in them, is discussed.

  13. Impact assessment of draft DOE Order 5820.2B. Radioactive Waste Technical Support Program

    SciTech Connect

    1995-04-01

    The Department of Energy (DOE) has prepared a revision to DOE Order 5820.2A, entitled ``Radioactive Waste Management.`` DOE issued DOE Order 5820.2A in September 1988 and, as the title implies, it covered only radioactive waste forms. The proposed draft order, entitled ``Waste Management,`` addresses the management of both radioactive and nonradioactive waste forms. It also includes spent nuclear fuel, which DOE does not consider a waste. Waste forms covered include hazardous waste, high-level waste, transuranic (TRU) waste, low-level radioactive waste, uranium and thorium mill tailings, mixed waste, and sanitary waste. The Radioactive Waste Technical Support Program (TSP) of Leached Idaho Technologies Company (LITCO) is facilitating the revision of this order. The EM Regulatory Compliance Division (EM-331) has requested that TSP estimate the impacts and costs of compliance with the revised order. TSP requested Dames & Moore to aid in this assessment by comparing requirements in Draft Order 5820.2B to ones in DOE Order 5820.2A and other DOE orders and Federal regulations. The assessment started with a draft version of 5820.2B dated January 14, 1994. DOE has released three updated versions of the draft order since then (dated May 20, 1994; August 26, 1994; and January 23, 1995). Each time DOE revised the order, Dames and Moore updated the assessment work to reflect the text changes. This report reflects the January 23, 1995 version of the draft order.

  14. High-level wastes: DOE names three sites for characterization

    SciTech Connect

    1986-07-01

    DOE announced in May 1986 that there will be there site characterization studies made to determine suitability for a high-level radioactive waste repository. The studies will include several test drillings to the proposed disposal depths. Yucca Mountain, Nevada; Deaf Smith Country, Texas, and Hanford, Washington were identified as the study sites, and further studies for a second repository site in the East were postponed. The affected states all filed suits in federal circuit courts because they were given no advance warning of the announcement of their selection or the decision to suspend work on a second repository. Criticisms of the selection process include the narrowing or DOE options.

  15. PYRO, a system for modeling fuel reprocessing

    SciTech Connect

    Ackerman, J.P.

    1989-01-01

    Compact, on-site fuel reprocessing and waste management for the Integral Fast Reactor are based on the pyrochemical reprocessing of metal fuel. In that process, uranium and plutonium in spent fuel are separated from fission products in an electrorefiner using liquid cadmium and molten salt solvents. Quantitative estimates of the distribution of the chemical elements among the metal and salt phases are essential for development of both individual pyrochemical process steps and the complete process. This paper describes the PYRO system of programs used to generate reliable mass flows and compositions.

  16. Nuclear Fuel Reprocessing

    SciTech Connect

    Michael F. Simpson; Jack D. Law

    2010-02-01

    This is an a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. No formal abstract was required for the article. The full article will be attached.

  17. DOE site performance assessment activities. Radioactive Waste Technical Support Program

    SciTech Connect

    Not Available

    1990-07-01

    Information on performance assessment capabilities and activities was collected from eight DOE sites. All eight sites either currently dispose of low-level radioactive waste (LLW) or plan to dispose of LLW in the near future. A survey questionnaire was developed and sent to key individuals involved in DOE Order 5820.2A performance assessment activities at each site. The sites surveyed included: Hanford Site (Hanford), Idaho National Engineering Laboratory (INEL), Los Alamos National Laboratory (LANL), Nevada Test Site (NTS), Oak Ridge National Laboratory (ORNL), Paducah Gaseous Diffusion Plant (Paducah), Portsmouth Gaseous Diffusion Plant (Portsmouth), and Savannah River Site (SRS). The questionnaire addressed all aspects of the performance assessment process; from waste source term to dose conversion factors. This report presents the information developed from the site questionnaire and provides a comparison of site-specific performance assessment approaches, data needs, and ongoing and planned activities. All sites are engaged in completing the radioactive waste disposal facility performance assessment required by DOE Order 5820.2A. Each site has achieved various degrees of progress and have identified a set of critical needs. Within several areas, however, the sites identified common needs and questions.

  18. THE ROLE OF LIQUID WASTE PRETREATMENT TECHNOLOGIES IN SOLVING THE DOE CLEAN-UP MISSION

    SciTech Connect

    Wilmarth, B; Sheryl Bush, S

    2008-10-31

    The objective of this report is to describe the pretreatment solutions that allow treatment to be tailored to specific wastes, processing ahead of the completion schedules for the main treatment facilities, and reduction of technical risks associated with future processing schedules. Wastes stored at Hanford and Savannah River offer challenging scientific and engineering tasks. At both sites, space limitations confound the ability to effectively retrieve and treat the wastes. Additionally, the radiation dose to the worker operating and maintaining the radiochemical plants has a large role in establishing the desired radioactivity removal. However, the regulatory requirements to treat supernatant and saltcake tank wastes differ at the two sites. Hanford must treat and remove radioactivity from the tanks based on the TriParty Agreement and Waste Incidental to Reprocessing (WIR) documentation. These authorizing documents do not specify treatment technologies; rather, they specify endstate conditions. Dissimilarly, Waste Determinations prepared at SRS in accordance with Section 3116 of the 2005 National Defense Authorization Act along with state operating permits establish the methodology and amounts of radioactivity that must be removed and may be disposed of in South Carolina. After removal of entrained solids and site-specific radionuclides, supernatant and saltcake wastes are considered to be low activity waste (LAW) and are immobilized in glass and disposed of at the Hanford Site Integrated Disposal Facility (IDF) or formulated into a grout for disposal at the Savannah River Site Saltstone Disposal Facility. Wastes stored at the Hanford Site or SRS comprise saltcake, supernate, and sludges. The supernatant and saltcake waste fractions contain primarily sodium salts, metals (e.g., Al, Cr), cesium-137 (Cs-137), technetium-99 (Tc-99) and entrained solids containing radionuclides such as strontium-90 (Sr-90) and transuranic elements. The sludges contain many of the

  19. [DOE method for evaluating environmental and waste management samples: Revision 1, Addendum 1

    SciTech Connect

    Goheen, S.C.

    1995-04-01

    The US Dapartment of Energy`s (DOE`s) environmental and waste management (EM) sampling and analysis activities require that large numbers of samples be analyzed for materials characterization, environmental surveillance, and site-remediation programs. The present document, DOE Methods for Evaluating Environmental and Waste Management Samples (DOE Methods), is a supplemental resource for analyzing many of these samples.

  20. High level nuclear waste

    SciTech Connect

    Crandall, J L

    1980-01-01

    The DOE Division of Waste Products through a lead office at Savannah River is developing a program to immobilize all US high-level nuclear waste for terminal disposal. DOE high-level wastes include those at the Hanford Plant, the Idaho Chemical Processing Plant, and the Savannah River Plant. Commercial high-level wastes, for which DOE is also developing immobilization technology, include those at the Nuclear Fuel Services Plant and any future commercial fuels reprocessing plants. The first immobilization plant is to be the Defense Waste Processing Facility at Savannah River, scheduled for 1983 project submission to Congress and 1989 operation. Waste forms are still being selected for this plant. Borosilicate glass is currently the reference form, but alternate candidates include concretes, calcines, other glasses, ceramics, and matrix forms.

  1. Release protocol to address DOE moratorium on shipments of waste generated in radiologically controlled areas

    SciTech Connect

    Rathbun, L A; Boothe, G F

    1992-10-01

    On May 17, 1991 the US DOE Office of Waste Operations issued a moratorium on the shipment of hazardous waste from radiologically contaminated or potentially contaminated areas on DOE sites to offsite facilities not licensed for radiological material. This document describes a release protocol generated by Westinghouse Hanford submitted for US DOE approval. Topics considered include designating Radiological Materials Management Areas (RMMAs), classification of wastes, handling of mixed wastes, detection limits.

  2. Mixed waste characterization, treatment & disposal focus area

    SciTech Connect

    1996-08-01

    The mission of the Mixed Waste Characterization, Treatment, and Disposal Focus Area (referred to as the Mixed Waste Focus Area or MWFA) is to provide treatment systems capable of treating DOE`s mixed waste in partnership with users, and with continual participation of stakeholders, tribal governments, and regulators. The MWFA deals with the problem of eliminating mixed waste from current and future storage in the DOE complex. Mixed waste is waste that contains both hazardous chemical components, subject to the requirements of the Resource Conservation and Recovery Act (RCRA), and radioactive components, subject to the requirements of the Atomic Energy Act. The radioactive components include transuranic (TRU) and low-level waste (LLW). TRU waste primarily comes from the reprocessing of spent fuel and the use of plutonium in the fabrication of nuclear weapons. LLW includes radioactive waste other than uranium mill tailings, TRU, and high-level waste, including spent fuel.

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

    SciTech Connect

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

    1995-05-01

    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.

  4. 40 CFR 266.235 - What waste treatment does the storage and treatment conditional exemption allow?

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) STANDARDS FOR THE MANAGEMENT OF SPECIFIC HAZARDOUS WASTES AND SPECIFIC TYPES OF HAZARDOUS WASTE MANAGEMENT FACILITIES Conditional Exemption for Low... 40 Protection of Environment 27 2011-07-01 2011-07-01 false What waste treatment does the...

  5. 40 CFR 60.1010 - Does this subpart apply to my municipal waste combustion unit?

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... waste combustion unit? 60.1010 Section 60.1010 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY... Performance for Small Municipal Waste Combustion Units for Which Construction is Commenced After August 30....1010 Does this subpart apply to my municipal waste combustion unit? Yes, if your municipal waste...

  6. 40 CFR 60.1010 - Does this subpart apply to my municipal waste combustion unit?

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... waste combustion unit? 60.1010 Section 60.1010 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY... Performance for Small Municipal Waste Combustion Units for Which Construction is Commenced After August 30....1010 Does this subpart apply to my municipal waste combustion unit? Yes, if your municipal waste...

  7. 40 CFR 60.1010 - Does this subpart apply to my municipal waste combustion unit?

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... waste combustion unit? 60.1010 Section 60.1010 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY... Performance for Small Municipal Waste Combustion Units for Which Construction is Commenced After August 30....1010 Does this subpart apply to my municipal waste combustion unit? Yes, if your municipal waste...

  8. Technical aspects of fuel reprocessing

    SciTech Connect

    Groenier, W.S.

    1982-02-01

    The purpose of this paper is to present a brief description of fuel reprocessing and some present developments which show the reliability of nuclear energy as a long-term supply. The following topics are discussed: technical reasons for reprocessing; economic reasons for reprocessing; past experience; justification for advanced reprocessing R and D; technical aspects of current reprocessing development. The present developments are mainly directed at the reprocessing of breeder reactor fuels but there are also many applications to light-water reactor fuel reprocessing. These new developments involve totally remote operation, and maintenance. To demonstrate this advanced reprocessing concept, pilot-scale demonstration facilities are planned with commercial application occurring sometime after the year 2000. (ATT)

  9. Reprocessing of research reactor fuel the Dounreay option

    SciTech Connect

    Cartwright, P.

    1997-08-01

    Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storage tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.

  10. EXTERNAL CRITICALITY CALCULATION FOR DOE SNF CODISPOSAL WASTE PACKAGES

    SciTech Connect

    H. Radulescu

    2002-10-18

    The purpose of this document is to evaluate the potential for criticality for the fissile material that could accumulate in the near-field (invert) and in the far-field (host rock) beneath the U.S. Department of Energy (DOE) spent nuclear fuel (SNF) codisposal waste packages (WPs) as they degrade in the proposed monitored geologic repository at Yucca Mountain. The scope of this calculation is limited to the following DOE SNF types: Shippingport Pressurized Water Reactor (PWR), Enrico Fermi, Fast Flux Test Facility (FFTF), Fort St. Vrain, Melt and Dilute, Shippingport Light Water Breeder Reactor (LWBR), N-Reactor, and Training, Research, Isotope, General Atomics reactor (TRIGA). The results of this calculation are intended to be used for estimating the probability of criticality in the near-field and in the far-field. There are no limitations on use of the results of this calculation. The calculation is associated with the waste package design and was developed in accordance with the technical work plan, ''Technical Work Plan for: Department of Energy Spent Nuclear Fuel and Plutonium Disposition Work Packages'' (Bechtel SAIC Company, LLC [BSC], 2002a). This calculation is subject to the Quality Assurance Requirements and Description (QARD) per the activity evaluation under work package number P6212310Ml in the technical work plan TWP-MGR-MD-0000 10 REV 01 (BSC 2002a).

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

    SciTech Connect

    Not Available

    1990-01-01

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

  12. Overview of the DOE Waste Facilities Operations Robotics Technology Development Program

    SciTech Connect

    Ward, C.R.

    1992-01-01

    The Department of Energy's Office of Technology Development has initiated a Robotics Technology Development Program that includes technology aimed at DOE Waste Facilities Operations (WFO). Much of this technology may also be applicable to waste facilities outside of DOE and will be available for use. This is a team effort of several DOE Laboratories and Sites. The WFO team includes the Savannah River Technology Center, Sandia National Laboratories, Oak Ridge National Laboratory, Idaho National Engineering Laboratory, Lawrence Livermore National Laboratory, the Fernald Environmental Management Project and the Waste Isolation Pilot Plant. DOE has waste facilities currently in operation at many sites and many more of these facilities are planned as part of the DOE site cleanup effort. As the cleanup continues, existing facilities and new facilities will be taxed with an increasing volume of waste, more varied waste streams, more complex processing requirements, more challenging waste characteristics, more stringent waste form criteria, and stricter regulations. The WFO Robotics Technology Development Program will address these challenges by developing robotic systems technology that will be safer, better, faster, and cheaper than existing technology. The goals of this technology development are to remove humans from both radiologically and physically dangerous environments in waste facilities, improve the quality and quality assurance of operations, increase the throughput of operations, increase the flexibility of facilities, reduce the manpower requirements for operations and meet federal, state and local regulations. There are four areas within the WFO program; the Mixed Waste Treatment Project, Stored Waste, the Waste Isolation Pilot Plant and Remote Size Reduction.

  13. Nuclear Fuel Reprocessing

    SciTech Connect

    Harold F. McFarlane; Terry Todd

    2013-11-01

    Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore. Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor

  14. 10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... solvent extraction system, or equivalent, and the concentrated wastes from subsequent extraction cycles... on unexpended balances will be designed to defray all costs of disposal and perpetual surveillance.... Disposal of high-level radioactive fission product waste material will not be permitted on any land...

  15. 10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... solvent extraction system, or equivalent, and the concentrated wastes from subsequent extraction cycles... on unexpended balances will be designed to defray all costs of disposal and perpetual surveillance.... Disposal of high-level radioactive fission product waste material will not be permitted on any land...

  16. 10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... solvent extraction system, or equivalent, and the concentrated wastes from subsequent extraction cycles... on unexpended balances will be designed to defray all costs of disposal and perpetual surveillance.... Disposal of high-level radioactive fission product waste material will not be permitted on any land...

  17. Nuclear fuel reprocessing deactivation plan for the Idaho Chemical Processing Plant, Revision 1

    SciTech Connect

    Patterson, M.W.

    1994-10-01

    The decision was announced on April 28, 1992 to cease all United States Department of Energy (DOE) reprocessing of nuclear fuels. This decision leads to the deactivation of all fuels dissolution, solvent extraction, krypton gas recovery operations, and product denitration at the Idaho Chemical Processing Plant (ICPP). The reprocessing facilities will be converted to a safe and stable shutdown condition awaiting future alternate uses or decontamination and decommissioning (D&D). This ICPP Deactivation Plan includes the scope of work, schedule, costs, and associated staffing levels necessary to achieve a safe and orderly deactivation of reprocessing activities and the Waste Calcining Facility (WCF). Deactivation activities primarily involve shutdown of operating systems and buildings, fissile and hazardous material removal, and related activities. A minimum required level of continued surveillance and maintenance is planned for each facility/process system to ensure necessary environmental, health, and safety margins are maintained and to support ongoing operations for ICPP facilities that are not being deactivated. Management of the ICPP was transferred from Westinghouse Idaho Nuclear Company, Inc. (WINCO) to Lockheed Idaho Technologies Company (LITCO) on October 1, 1994 as part of the INEL consolidated contract. This revision of the deactivation plan (formerly the Nuclear Fuel Reprocessing Phaseout Plan for the ICPP) is being published during the consolidation of the INEL site-wide contract and the information presented here is current as of October 31, 1994. LITCO has adopted the existing plans for the deactivation of ICPP reprocessing facilities and the plans developed under WINCO are still being actively pursued, although the change in management may result in changes which have not yet been identified. Accordingly, the contents of this plan are subject to revision.

  18. 40 CFR 266.235 - What waste treatment does the storage and treatment conditional exemption allow?

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 26 2010-07-01 2010-07-01 false What waste treatment does the storage and treatment conditional exemption allow? 266.235 Section 266.235 Protection of Environment...-Level Mixed Waste Storage, Treatment, Transportation and Disposal. Treatment § 266.235 What waste...

  19. 40 CFR 60.1010 - Does this subpart apply to my municipal waste combustion unit?

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... waste combustion unit? 60.1010 Section 60.1010 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY... Performance for Small Municipal Waste Combustion Units for Which Construction is Commenced After August 30....1010 Does this subpart apply to my municipal waste combustion unit? Yes, if your municipal...

  20. 40 CFR 60.1010 - Does this subpart apply to my municipal waste combustion unit?

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... waste combustion unit? 60.1010 Section 60.1010 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY... Performance for Small Municipal Waste Combustion Units for Which Construction is Commenced After August 30....1010 Does this subpart apply to my municipal waste combustion unit? Yes, if your municipal...

  1. Reprocessing in Luminous Disks

    NASA Technical Reports Server (NTRS)

    Bell, K. Robbins; DeVincenzi, Donald L. (Technical Monitor)

    1999-01-01

    We develop and investigate a procedure that accounts for disk reprocessing of photons that originate in the disk itself. Surface temperatures and simple, black body spectral energy distributions (SEDs) of protostellar disks are calculated. In disks that flare with radius, reprocessing of stellar photons results in temperature profiles considerably shallower than r(sup -3/4). Including the disk as a radiation source (as in the case of actively secreting disks) along with the stellar source further flattens the temperature profile. Disks that flare strongly near the star and then smoothly curve over and become shadowed at some distance ("decreasing curvature" disks) exhibit nearly power-law temperature profiles which result in power-law infrared SEDs with slopes in agreement with typical observations of young stellar objects. Disk models in which the photospheric thickness is controlled by the local opacity and in which the temperature decreases with radius naturally show this shape. Uniformly flaring models do not match observations as well; progressively stronger reprocessing at larger radii leads to SEDs that flatten toward the infrared or even have a second peak at the wavelength corresponding (through the Wien law) to the temperature of the outer edge of the disk. In FU Orionis outbursting systems, the dominant source of energy is the disk itself. The details of the reprocessing depend sensitively on the assumed disk shape and emitted temperature profile. The thermal instability outburst models of Bell Lin reproduce trends in the observed SEDs of Fuors with T varies as r(sup -3/4) in the inner disk (r approx. less than 0.25au corresponding to lambda approx. less than 10 microns) and T varies as r(sup -1/2) in the outer disk. Surface irradiation during outburst and quiescence is compared in the region of planet formation (1 - 10 au). The contrast between the two phases is diminished by the importance of the reprocessing of photons from the relatively high mass

  2. Reprocessing in Luminous Disks

    NASA Technical Reports Server (NTRS)

    Bell, K. Robbins; DeVincenzi, Donald L. (Technical Monitor)

    1999-01-01

    We develop and investigate a procedure that accounts for disk reprocessing of photons that originate in the disk itself. Surface temperatures and simple, black body spectral energy distributions (SEDs) of protostellar disks are calculated. In disks that flare with radius, reprocessing of stellar photons results in temperature profiles considerably shallower than r(sup -3/4). Including the disk as a radiation source (as in the case of actively secreting disks) along with the stellar source further flattens the temperature profile. Disks that flare strongly near the star and then smoothly curve over and become shadowed at some distance ("decreasing curvature" disks) exhibit nearly power-law temperature profiles which result in power-law infrared SEDs with slopes in agreement with typical observations of young stellar objects. Disk models in which the photospheric thickness is controlled by the local opacity and in which the temperature decreases with radius naturally show this shape. Uniformly flaring models do not match observations as well; progressively stronger reprocessing at larger radii leads to SEDs that flatten toward the infrared or even have a second peak at the wavelength corresponding (through the Wien law) to the temperature of the outer edge of the disk. In FU Orionis outbursting systems, the dominant source of energy is the disk itself. The details of the reprocessing depend sensitively on the assumed disk shape and emitted temperature profile. The thermal instability outburst models of Bell Lin reproduce trends in the observed SEDs of Fuors with T varies as r(sup -3/4) in the inner disk (r approx. less than 0.25au corresponding to lambda approx. less than 10 microns) and T varies as r(sup -1/2) in the outer disk. Surface irradiation during outburst and quiescence is compared in the region of planet formation (1 - 10 au). The contrast between the two phases is diminished by the importance of the reprocessing of photons from the relatively high mass

  3. Combined Waste Form Cost Trade Study

    SciTech Connect

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

    2008-11-01

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

  4. Planned revision to DOE Order 5820.2A, Radioactive Waste Management

    SciTech Connect

    Duggan, G.J.; Williams, R.E.; Kudera, D.E.; Bailey, D.E.

    1993-03-01

    US Department of Energy Headquarters initiated efforts to revise DOE Order 5820.2A, ``Radioactive Waste Management``. The purpose of the revision is to enhance DOE waste management requirements, reflect new DOE organizational responsibilities, and consolidate requirements for management of all waste, under the responsibility of Environmental Restoration and Waste Management, into a single order. This paper discusses the revision philosophy, objectives of the revision, and strategy for the revision. Issues being considered for inclusion in the revision and recommended methods of resolving each issue are also discussed.

  5. Risk averse` DOE is wasting time, money in cleanup effort-GAO

    SciTech Connect

    Newman, P.

    1994-09-01

    According to an August 1994 GAO report, internal strife, poor decisionmaking and conflicting stakeholder interests have plague the cleanup effort and prevented DOE from taking advantages of what its won technology program call the best hope for ensuring a substantive waste reduction. This article details the problems effecting radioactive waste cleanup at DOE facilities, and lists the five technology priorities which have been established.

  6. Technology needs for remediation: Hanford and other DOE sites. Buried Waste Integrated Demonstration Program

    SciTech Connect

    Stapp, D.C.

    1993-01-01

    Technologies are being developed under the Buried Waste Integrated Demonstration (BWID) program to facilitate remediation of the US Department of Energy`s (DOE) buried and stored low-level radioactive, transuranic (TRU), and mixed radioactive and hazardous buried wastes. The BWID program is being coordinated by the Idaho National Engineering Laboratory (INEL) in southeastern Idaho, a DOE site that has large volumes of buried radioactive wastes. The program is currently focusing its efforts on the problems at INEL`s Subsurface Disposal Area (SDA) of the Radioactive Waste Management Complex (RWMC). As specific technologies are successfully demonstrated, they will be available for transfer to applications at other DOE buried waste sites. The purpose of this study is to present buried waste technology needs that have been identified for DOE sites other than INEL.

  7. THE NGA-DOE GRANT TO EXAMINE CRITICAL ISSUES RELATED TO RADIOACTIVE WASTE AND MATERIALS DISPOSITION INVOLVING DOE FACILITIES

    SciTech Connect

    Ann M. Beauchesne

    1999-01-31

    Through the National Governors' Association (NGA) project ''Critical Issues Related to Radioactive Waste and Materials Disposition Involving DOE Facilities'' NGA brings together Governors' policy advisors, state regulators, and DOE officials to examine critical issues related to the cleanup and operation of DOE nuclear weapons and research facilities. Topics explored through this project include: (1) Decisions involving disposal of mixed, low-level, and transuranic (TRU) waste and disposition of nuclear materials; (2) Decisions involving DOE budget requests and their effect on environmental cleanup and compliance at DOE facilities; (3) Strategies to treat mixed, low-level, and transuranic (TRU) waste and their effect on individual sites in the complex; (4) Changes to the FFCA site treatment plans as a result of proposals in the Department's Accelerating Cleanup: Paths to Closure plan and contractor integration analysis; (5) Interstate waste and materials shipments; and (6) Reforms to existing RCRA and CERCLA regulations/guidance to address regulatory overlap and risks posed by DOE wastes. The overarching theme of this project is to help the Department improve coordination of its major program decisions with Governors' offices and state regulators and to ensure such decisions reflect input from these key state officials and stakeholders. This report summarizes activities conducted during the quarter from October 1, 1998 through January 31, 1999, under the NGA grant. The work accomplished by the NGA project team during the past four months can be categorized as follows: (1) maintained open communication with DOE on a variety of activities and issues within the DOE environmental management complex; (2) maintained communication with NGA Federal Facilities Compliance Task Force members regarding DOE efforts to formulate a configuration for mixed low-level waste and low-level treatment and disposal, external regulation of DOE; and EM Integration activities; and (3

  8. THE NGA-DOE GRANT TO EXAMINE CRITICAL ISSUES RELATED TO RADIOACTIVE WASTE AND MATERIALS DISPOSITION INVOLVING DOE FACILITIES

    SciTech Connect

    Ann B. Beauchesne

    1998-09-30

    Through the National Governors' Association (NGA) project ''Critical Issues Related to Radioactive Waste and Materials Disposition Involving DOE Facilities'' NGA brings together Governors' policy advisors, state regulators, and DOE officials to examine critical issues related to the cleanup and operation of DOE nuclear weapons and research facilities. Topics explored through this project include: (1) Decisions involving disposal of mixed, low-level, and transuranic (TRU) waste and disposition of nuclear materials; (2) Decisions involving DOE budget requests and their effect on environmental cleanup and compliance at DOE facilities; (3) Strategies to treat mixed, low-level, and transuranic (TRU) waste and their effect on individual sites in the complex; (4) Changes to the FFCA site treatment plans as a result of proposals in the Department's Accelerating Cleanup: Paths to Closure plan and contractor integration analysis; (5) Interstate waste and materials shipments; and (6) Reforms to existing RCRA and CERCLA regulations/guidance to address regulatory overlap and risks posed by DOE wastes. The overarching theme of this project is to help the Department improve coordination of its major program decisions with Governors' offices and state regulators and to ensure such decisions reflect input from these key state officials and stakeholders. This report summarizes activities conducted during the quarter from June 1, 1998 through September 30, 1998, under the NGA grant. The work accomplished by the NGA project team during the past four months can be categorized as follows: (1) maintained open communication with DOE on a variety of activities and issues within the DOE environmental management complex; (2) maintained communication with NGA Federal Facilities Compliance Task Force members regarding DOE efforts to formulate a configuration for mixed low-level waste and low-level treatment and disposal, external regulation of DOE; and EM Integration activities; and (3

  9. ICPP Waste Management Technology Development Program

    SciTech Connect

    Hogg, G.W.; Olson, A.L.; Knecht, D.A.; Bonkoski, M.J.

    1993-01-01

    As a result of the decision to curtail reprocessing at the Idaho Chemical Processing Plant (ICPP), a Spent fuel and Waste Management Technology Development plan has been implemented to identify acceptable options for disposing of the (1) sodium-bearing liquid radioactive waste, (2) radioactive calcine, and (3) irradiated spent fuel stored at the Idaho National Engineering Laboratory (INEL). The plan was developed jointly by DOE and WINCO.

  10. DOE's Remote-Handled TRU Waste Characterization Program: Implementation Plan

    EPA Pesticide Factsheets

    Remote-handled (RH) transuranic (TRU) waste characterization, which involves obtaining chemical, radiological, and physical data, is a primary component of ensuring compliance of the Waste Isolation Pilot Plant (WIPP) with regulatory requirements.

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

    SciTech Connect

    Not Available

    1985-08-01

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

  12. Long-term high-level waste technology

    NASA Astrophysics Data System (ADS)

    Cornman, W. R.

    1980-07-01

    This series of reports summarizes research and development studies on the immobilization of high level wastes from the chemical reprocessing of nuclear reactor fuels. Immobilization of the wastes (defense and commercial) consists of placing them in a high integrity form with a very low potential for radionuclide release. Immobilization of commercial wastes is being considered on a contingency basis in the event that reprocessing is resumed. The basic plan for meeting the goal of immobilization of the DOE high level wastes is: (1) to develop technology to support a realistic choice of waste form alternatives for each of the three DOE sites; (2) to develop product and processing technology with sufficient scaleup to provide design data for full scale facilities; and (3) to construct and operate the facilities.

  13. 10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... and Related Waste Management Facilities F Appendix F to Part 50 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. F Appendix F to Part 50—Policy... qualifications as required by § 50.33(f), shall include information enabling the Commission to determine...

  14. 10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... and Related Waste Management Facilities F Appendix F to Part 50 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. F Appendix F to Part 50—Policy... qualifications as required by § 50.33(f), shall include information enabling the Commission to determine...

  15. Comparison of selected DOE and non-DOE requirements, standards, and practices for Low-Level Radioactive Waste Disposal

    SciTech Connect

    Cole, L.; Kudera, D.; Newberry, W.

    1995-12-01

    This document results from the Secretary of Energy`s response to Defense Nuclear Facilities Safety Board Recommendation 94--2. The Secretary stated that the US Department of Energy (DOE) would ``address such issues as...the need for additional requirements, standards, and guidance on low-level radioactive waste management. `` The authors gathered information and compared DOE requirements and standards for the safety aspects Of low-level disposal with similar requirements and standards of non-DOE entities.

  16. Development of nuclear analysis capabilities for DOE waste management activities

    SciTech Connect

    Parks, C.V.; DeHart, M.D.; Broadhead, B.L.; Hopper, C.M.; Petrie, L.M.

    1998-05-01

    The objective of this project is to develop and demonstrate prototypic analysis capabilities that can be used by the nuclear safety analysis practitioners to: (1) demonstrate a more thorough understanding of the underlying physics phenomena that can lead to improved reliability and defensibility of safety evaluations; and (2) optimize operations related to the handling, storage, transportation, and disposal of fissile material and DOE spent fuel. To address these problems, the project will investigate the implementation of sensitivity and uncertainty methods within existing Monte Carlo codes used for criticality safety analyses, as well as within a new deterministic code that allows specification of arbitrary grids to accurately model the geometry details required in a criticality safety analysis. This capability can facilitate improved estimations of the required subcritical margin and potentially enable the use of a broader range of experiments in the validation process. The new arbitrary-grid radiation transport code will also enable detailed geometric modeling valuable for improved accuracy in application to a myriad of other problems related to waste characterization. Application to these problems will also be explored.

  17. Crystal chemistry of sodium zirconium phosphate based simulated ceramic waste forms of effluent cations (Ba(2+), Sn(4+), Fe(3+), Cr(3+), Ni(2+) and Si(4+)) from light water reactor fuel reprocessing plants.

    PubMed

    Shrivastava, O P; Chourasia, Rashmi

    2008-05-01

    A novel concept of immobilization of light water reactor (LWR) fuel reprocessing waste effluent through interaction with sodium zirconium phosphate (NZP) has been established. Such conversion utilizes waste materials like zirconium and nickel alloys, stainless steel, spent solvent tri-butyl phosphate and concentrated solution of NaNO(3). The resultant multi component NZP material is a physically and chemically stable single phase crystalline product having good mechanical strength. The NZP matrix can also incorporate all types of fission product cations in a stable crystalline lattice structure; therefore, the resultant solid solutions deserve quantification of crystallographic data. In this communication, crystal chemistry of the two types of simulated waste forms (type I-Na(1.49)Zr(1.56)Sn(0.02)Fe(0).(28)Cr(0.07)Ni(0.07)P(3)O(12) and type II-Na(1.35)Ba(0.14)Zr(1.56)Sn(0.02)Fe(0).(28)Cr(0.07)Ni(0.07)P(2.86)Si(0.14)O(12)) has been investigated using General Structure Analysis System (GSAS) programming of the X-ray powder diffraction data. About 4001 data points of each have been subjected to Rietveld analysis to arrive at a satisfactory structural convergence of Rietveld parameters; R-pattern (R(p))=0.0821, R-weighted pattern (R(wp))=0.1266 for type I and R(p)=0.0686, R(wp)=0.0910 for type II. The structure of type I and type II waste forms consist of ZrO(6) octahedra and PO(4) tetrahedra linked by the corners to form a three-dimensional network. Each phosphate group is on a two-fold rotation axis and is linked to four ZrO(6) octahedra while zirconium octahedra lies on a three-fold rotation axis and is connected to six PO(4) tetrahedra. Though the expansion along c-axis and shrinkage along a-axis with slight distortion of bond angles in the synthesized crystal indicate the flexibility of the structure, the waste forms are basically of NZP structure. Morphological examination by SEM reveals that the size of almost rectangular parallelepiped crystallites varies

  18. Complex-wide review of DOE`s management of low-level radioactive waste - progress to date

    SciTech Connect

    Letourneau, M.J.

    1995-12-31

    The Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 94-2 includes a recommendation that the Department of Energy (DOE) conduct a comprehensive, complex-wide review of the low-level waste issue to establish the dimensions of the low-level waste problem and to identify necessary corrective actions to address the safe disposition of past, present, and future volumes. DOE`s Implementation Plan calls for the conduct of a complex-wide review of low-level radioactive waste treatment, storage, and disposal sites to identify environmental, safety, and health vulnerabilities. The complex-wide review focuses on low-level waste disposal facilities through a site evaluation survey, reviews of existing documentation, and onsite observations. Low-level waste treatment and storage facilities will be assessed for their ability to meet waste acceptance criteria for disposal. Results from the complex-wide review will be used to form the basis for an integrated and planned set of actions to correct the identified vulnerabilities and to prompt development of new requirements for managing low-level waste.

  19. Commentary: a call to go green in health care by reprocessing medical equipment.

    PubMed

    Kwakye, Gifty; Pronovost, Peter J; Makary, Martin A

    2010-03-01

    Health care is one of the largest contributors to waste production in the United States. Given increased awareness of the environmental and financial costs associated with waste disposal and its public health impact, many hospitals are adopting environmentally friendly practices that reduce waste production and offer equally effective, yet less expensive alternatives. Reprocessing of medical equipment is one such practice that has gained popularity in recent years and has led to major cost savings across several medical disciplines. In this commentary, we seek to take a closer look at the practice of reprocessing, explore the evidence surrounding its safety, and suggest implications of reprocessing for medical centers.

  20. Requirements for shipment of DOE radioactive mixed waste

    SciTech Connect

    Gablin, K.; No, Hyo; Herman, J.

    1993-08-01

    There are several sources of radioactive mixed waste (RMW) at Argonne National Laboratory which, in the past, were collected at waste tanks and/or sludge tanks. They were eventually pumped out by special pumps and processed in an evaporator located in the waste operations area in Building No. 306. Some of this radioactive mixed waste represents pure elementary mercury. These cleaning tanks must be manually cleaned up because the RMW material was too dense to pump with the equipment in use. The four tanks being discussed in this report are located in Building No. 306. They are the Acid Waste Tank, IMOX/FLOC Tanks, Evaporation Feed Tanks, and Waste Storage Tanks. All of these tanks are characterized and handled separately. This paper discusses the process and the requirements for characterization and the associated paperwork for Argonne Waste to be shipped to Westinghouse Hanford Company for storage.

  1. Overview of the DOE Waste Facilities Operations Robotics Technology Development Program

    SciTech Connect

    Ward, C.R.

    1992-07-01

    The Department of Energy`s Office of Technology Development has initiated a Robotics Technology Development Program that includes technology aimed at DOE Waste Facilities Operations (WFO). Much of this technology may also be applicable to waste facilities outside of DOE and will be available for use. This is a team effort of several DOE Laboratories and Sites. The WFO team includes the Savannah River Technology Center, Sandia National Laboratories, Oak Ridge National Laboratory, Idaho National Engineering Laboratory, Lawrence Livermore National Laboratory, the Fernald Environmental Management Project and the Waste Isolation Pilot Plant. DOE has waste facilities currently in operation at many sites and many more of these facilities are planned as part of the DOE site cleanup effort. As the cleanup continues, existing facilities and new facilities will be taxed with an increasing volume of waste, more varied waste streams, more complex processing requirements, more challenging waste characteristics, more stringent waste form criteria, and stricter regulations. The WFO Robotics Technology Development Program will address these challenges by developing robotic systems technology that will be safer, better, faster, and cheaper than existing technology. The goals of this technology development are to remove humans from both radiologically and physically dangerous environments in waste facilities, improve the quality and quality assurance of operations, increase the throughput of operations, increase the flexibility of facilities, reduce the manpower requirements for operations and meet federal, state and local regulations. There are four areas within the WFO program; the Mixed Waste Treatment Project, Stored Waste, the Waste Isolation Pilot Plant and Remote Size Reduction.

  2. Low-level waste management alternatives and analysis in DOE`s programmatic environmental impact statement

    SciTech Connect

    Gerstein, J.S.

    1993-03-01

    The Department of Energy is preparing a Programmatic Environmental Impact Statement (PEIS) for the Environmental Restoration and Waste Management Program. The PEIS has been divided into an Environmental Restoration section and a Waste Management section. Each section has a unique set of alternatives. This paper will focus on the waste management alternatives and analysis. The set of alternatives for waste management has been divided into waste categories. These categories are: high-level waste, transuranic waste, low-level waste, low-level mixed waste, greater-than-class C and low-level waste from commercial sources, hazardous waste, and spent nuclear fuel. This paper will discuss the alternatives and analytical approach that will be used to evaluate these alternatives for the low-level waste section. Although the same alternatives will be considered for all waste types, the analysis will be performed separately for each waste type. In the sections that follow, information will be provided on waste management configurations, the analysis of waste management alternatives, waste types and locations, facility and transportation activities, the facility and transportation impacts assessment, and the compilation of impacts.

  3. Reprocessing of nonoptimally exposed holograms

    SciTech Connect

    Phipps, G.S.; Robertson, C.E.; Tamashiro, F.M.

    1980-03-01

    Two reprocessing techniques have been investigated that are capable of correcting the effects of nonoptimum optical density of photographic amplitude holograms recorded on Agfa-Gevaert type 10E75 plates. In some cases a reprocessed hologram will exhibit a diffraction efficiency even higher than that obtainable from a hologram exposed and processed to the optimum density. The SNR of the reprocessed holograms is much higher than that of the same holograms belached with cupric bromide. In some cases the SNR approaches the optimum value for a properly exposed amplitude hologram. Subjective image quality and resolution of reprocessed hologram reconstructins appear to be no different than for normal single-development holograms. Repeated reprocessing is feasible and in some cases desirable as a means of increasing diffraction efficiency.

  4. 1994 annual report on waste generation and waste minimization progress as required by DOE Order 5400.1

    SciTech Connect

    Irwin, E.F.; Poligone, S.E.

    1995-10-16

    The Y-12 Plant serves as a key manufacturing technology center for the development and demonstration of unique materials, components, and services of importance to the Department of Energy (DOE) and the nation. This is accomplished through the reclamation and storage of nuclear materials, manufacture of nuclear materials, manufacture of components for the nation`s defense capabilities, support to national security programs, and services provided to other customers as approved by DOE. We are recognized by our people, the community, and our customers as innovative, responsive, and responsible. We are a leader in worker health and safety, environmental protection, and stewardship of our national resources. As a DOE facility, Y-12 also supports DOE`s waste minimization mission. Data contained in this report represents waste generation in Tennessee.

  5. DOE model conference on waste management and environmental restoration

    SciTech Connect

    Not Available

    1990-01-01

    Reports dealing with current topics in waste management and environmental restoration were presented at this conference in six sessions. Session 1 covered the Hot Topics'' including regulations and risk assessment. Session 2 dealt with waste reduction and minimization; session 3 dealt with waste treatment and disposal. Session 4 covered site characterization and analysis. Environmental restoration and associated technologies wee discussed in session 5 and 6. Individual papers have been cataloged separately.

  6. Mercury and tritium removal from DOE waste oils

    SciTech Connect

    Klasson, E.T.

    1997-10-01

    This work covers the investigation of vacuum extraction as a means to remove tritiated contamination as well as the removal via sorption of dissolved mercury from contaminated oils. The radiation damage in oils from tritium causes production of hydrogen, methane, and low-molecular-weight hydrocarbons. When tritium gas is present in the oil, the tritium atom is incorporated into the formed hydrocarbons. The transformer industry measures gas content/composition of transformer oils as a diagnostic tool for the transformers` condition. The analytical approach (ASTM D3612-90) used for these measurements is vacuum extraction of all gases (H{sub 2}, N{sub 2}, O{sub 2}, CO, CO{sub 2}, etc.) followed by analysis of the evolved gas mixture. This extraction method will be adapted to remove dissolved gases (including tritium) from the SRS vacuum pump oil. It may be necessary to heat (60{degrees}C to 70{degrees}C) the oil during vacuum extraction to remove tritiated water. A method described in the procedures is a stripper column extraction, in which a carrier gas (argon) is used to remove dissolved gases from oil that is dispersed on high surface area beads. This method appears promising for scale-up as a treatment process, and a modified process is also being used as a dewatering technique by SD Myers, Inc. (a transformer consulting company) for transformers in the field by a mobile unit. Although some mercury may be removed during the vacuum extraction, the most common technique for removing mercury from oil is by using sulfur-impregnated activated carbon (SIAC). SIAC is currently being used by the petroleum industry to remove mercury from hydrocarbon mixtures, but the sorbent has not been previously tested on DOE vacuum oil waste. It is anticipated that a final process will be similar to technologies used by the petroleum industry and is comparable to ion exchange operations in large column-type reactors.

  7. Prospects for pyrolysis technologies in managing municipal, industrial, and DOE cleanup wastes

    SciTech Connect

    Reaven, S.J.

    1994-12-01

    Pyrolysis converts portions of municipal solid wastes, hazardous wastes, and special wastes such as tires, medical wastes, and even old landfills into solid carbon and a liquid or gaseous hydrocarbon stream. Pyrolysis heats a carbonaceous waste stream typically to 290--900 C in the absence of oxygen, and reduces the volume of waste by 90% and its weight by 75%. The solid carbon char has existing markets as an ingredient in many manufactured goods, and as an adsorbent or filter to sequester certain hazardous wastes. Pyrolytic gases may be burned as fuel by utilities, or liquefied for use as chemical feedstocks, or low-pollution motor vehicle fuels and fuel additives. This report analyzes the potential applications of pyrolysis in the Long Island region and evaluates for the four most promising pyrolytic systems their technological and commercial readiness, their applicability to regional waste management needs, and their conformity with DOE requirements for environmental restoration and waste management. This summary characterizes their engineering performance, environmental effects, costs, product applications, and markets. Because it can effectively treat those wastes that are inadequately addressed by current systems, pyrolysis can play an important complementing role in the region`s existing waste management strategy. Its role could be even more significant if the region moves away from existing commitments to incineration and MSW composting. Either way, Long Island could become the center for a pyrolysis-based recovery services industry serving global markets in municipal solid waste treatment and hazardous waste cleanup. 162 refs.

  8. Progress toward maturity of DOE methods for evaluating environmental and waste management samples

    SciTech Connect

    Cosby, W.C.; Goheen, S.C.; McCulloch, M.

    1994-07-01

    The document DOE Methods for Evaluating Environmental and Waste Management Samples (DOE Methods) has been in circulation since October 1992. DOE Methods is a living document, being updated twice each year. It contains both sampling and analytical methods in support of US Department of Energy/environmental restoration and waste management (DOE/EM) activities. Guidance on how to carry out sampling and analysis activities, focusing of EM needs, is also included in DOE Methods. This guidance applies to all aspects of sampling and analysis for EM. Methods from traditional standard methods documents often cannot provide needed characterization data because of radioactivity or complexity of the matrix. The intent of DOE Methods is to provide an alternative source of methods to meet this need. Efforts are underway to expand the use of DOE Methods throughout all DOE/EM programs. Copies of DOE Methods are available free of charge. The April 1994 update of the document includes 42 methods, of which 13 are new. In October 1994, Revision 2 of DOE Methods will be distributed. It will include additional guidance on how to plan sampling and analysis activities and will also include several new methods. DOE Methods is supported by the Laboratory Management Division of DOE. It is a vehicle for transgressing new technology for characterization capability within the environmental restoration (ER) and/or waste management (WM) community. As DOE Methods continues to evolve, its use and application will continue to grow.

  9. Cost and availability of gadolinium for nuclear fuel reprocessing plants

    SciTech Connect

    Klepper, O.H.

    1985-06-01

    Gadolinium is currently planned for use as a soluble neutron poison in nuclear fuel reprocessing plants to prevent criticality of solutions of spent fuel. Gadolinium is relatively rare and expensive. The present study was undertaken therefore to estimate whether this material is likely to be available in quantities sufficient for fuel reprocessing and at reasonable prices. It was found that gadolinium, one of 16 rare earth elements, appears in the marketplace as a by-product and that its present supply is a function of the production rate of other more prevalent rare earths. The potential demand for gadolinium in a fuel reprocessing facility serving a future fast reactor industry amounts to only a small fraction of the supply. At the present rate of consumption, domestic supplies of rare earths containing gadolinium are adequate to meet national needs (including fuel reprocessing) for over 100 years. With access to foreign sources, US demands can be met well beyond the 21st century. It is concluded therefore that the supply of gadolinium will quite likely be more than adequate for reprocessing spent fuel for the early generation of fast reactors. The current price of 99.99% pure gadolinium oxide lies in the range $50/lb to $65/lb (1984 dollars). By the year 2020, in time for reprocessing spent fuel from an early generation of large fast reactors, the corresponding values are expected to lie in the $60/lb to $75/lb (1984 dollars) price range. This increase is modest and its economic impact on nuclear fuel reprocessing would be minor. The economic potential for recovering gadolinium from the wastes of nuclear fuel reprocessing plants (which use gadolinium neutron poison) was also investigated. The cost of recycled gadolinium was estimated at over twelve times the cost of fresh gadolinium, and thus recycle using current recovery technology is not economical. 15 refs., 4 figs., 11 tabs.

  10. Criticality Potential of Waste Packages Containing DOE SNF Affected by Igneous Intrusion

    SciTech Connect

    D.S. Kimball; C.E. Sanders

    2006-02-07

    The Department of Energy (DOE) is currently preparing an application to submit to the U.S. Nuclear Regulatory Commission for a construction authorization for a monitored geologic repository. The repository will contain spent nuclear fuel (SNF) and defense high-level waste (DHLW) in waste packages placed in underground tunnels, or drifts. The primary objective of this paper is to perform a criticality analysis for waste packages containing DOE SNF affected by a disruptive igneous intrusion event in the emplacement drifts. The waste packages feature one DOE SNF canister placed in the center and surrounded by five High-Level Waste (HLW) glass canisters. The effective neutron multiplication factor (k{sub eff}) is determined for potential configurations of the waste package during and after an intrusive igneous event. Due to the complexity of the potential scenarios following an igneous intrusion, finding conservative and bounding configurations with respect to criticality requires some additional considerations. In particular, the geometry of a slumped and damaged waste package must be examined, drift conditions must be modeled over a range of parameters, and the chemical degradation of DOE SNF and waste package materials must be considered for the expected high temperatures. The secondary intent of this calculation is to present a method for selecting conservative and bounding configurations for a wide range of end conditions.

  11. High level radioactive waste processing experience in the US (an overview of the West Valley Demonstration Project)

    SciTech Connect

    Vance, R.F.; Borisch, R.R.

    1993-12-31

    The West Valley Nuclear Fuel Reprocessing Plant was constructed in 1966. Operations were stopped in 1972 after reprocessing 640 Mg (700 tons) of spent fuel. About 560,000 gallons of high-level radioactive liquid wastes from the Purex Process and 8,000 gallons of fuel containing thorium from the THOREX process were stored in underground steel tanks. The DOE contracted with West Valley Nuclear Services to operate the West Valley Demonstration Project for the processing of the radioactive wastes into a borosilicate waste form. This report provides a process overview and status report.

  12. Flowsheets and source terms for radioactive waste projections

    SciTech Connect

    Forsberg, C.W.

    1985-03-01

    Flowsheets and source terms used to generate radioactive waste projections in the Integrated Data Base (IDB) Program are given. Volumes of each waste type generated per unit product throughput have been determined for the following facilities: uranium mining, UF/sub 6/ conversion, uranium enrichment, fuel fabrication, boiling-water reactors (BWRs), pressurized-water reactors (PWRs), and fuel reprocessing. Source terms for DOE/defense wastes have been developed. Expected wastes from typical decommissioning operations for each facility type have been determined. All wastes are also characterized by isotopic composition at time of generation and by general chemical composition. 70 references, 21 figures, 53 tables.

  13. Composite quarterly technical report long-term high-level-waste technology, October-December 1981

    SciTech Connect

    Cornman, W.R.

    1982-06-01

    This document summarizes work performed at participating sites on the immobilization of high-level wastes from the chemical reprocessing of reactor fuels. The plan is to develop waste form alternatives for each of the three DOE sites (SRP, ICPP, and Hanford). Progress is reported in the following areas: waste preparation; fixation in glass, concrete, tailored ceramics, and coated particles; process and equipment development; and final handling. 12 figures, 19 tables. (DLC)

  14. Selection of melter systems for the DOE/Industrial Center for Waste Vitrification Research

    SciTech Connect

    Bickford, D.F.

    1993-12-31

    The EPA has designated vitrification as the best developed available technology for immobilization of High-Level Nuclear Waste. In a recent federal facilities compliance agreement between the EPA, the State of Washington, and the DOE, the DOE agreed to vitrify all of the Low Level Radioactive Waste resulting from processing of High Level Radioactive Waste stored at the Hanford Site. This is expected to result in the requirement of 100 ton per day Low Level Radioactive Waste melters. Thus, there is increased need for the rapid adaptation of commercial melter equipment to DOE`s needs. DOE has needed a facility where commercial pilot scale equipment could be operated on surrogate (non-radioactive) simulations of typical DOE waste streams. The DOE/Industry Center for Vitrification Research (Center) was established in 1992 at the Clemson University Department of Environmental Systems Engineering, Clemson, SC, to address that need. This report discusses some of the characteristics of the melter types selected for installation of the Center. An overall objective of the Center has been to provide the broadest possible treatment capability with the minimum number of melter units. Thus, units have been sought which have broad potential application, and which had construction characteristics which would allow their adaptation to various waste compositions, and various operating conditions, including extreme variations in throughput, and widely differing radiological control requirements. The report discusses waste types suitable for vitrification; technical requirements for the application of vitrification to low level mixed wastes; available melters and systems; and selection of melter systems. An annotated bibliography is included.

  15. Waste form product characteristics

    SciTech Connect

    Taylor, L.L.; Shikashio, R.

    1995-01-01

    The Department of Energy has operated nuclear facilities at the Idaho National Engineering Laboratory (INEL) to support national interests for several decades. Since 1953, it has supported the development of technologies for the storage and reprocessing of spent nuclear fuels (SNF) and the resultant wastes. However, the 1992 decision to discontinue reprocessing of SNF has left nearly 768 MT of SNF in storage at the INEL with unspecified plans for future dispositioning. Past reprocessing of these fuels for uranium and other resource recovery has resulted in the production of 3800 M{sup 3} calcine and a total inventory of 7600 M{sup 3} of radioactive liquids (1900 M{sup 3} destined for immediate calcination and the remaining sodium-bearing waste requiring further treatment before calcination). These issues, along with increased environmental compliance within DOE and its contractors, mandate operation of current and future facilities in an environmentally responsible manner. This will require satisfactory resolution of spent fuel and waste disposal issues resulting from the past activities. A national policy which identifies requirements for the disposal of SNF and high level wastes (HLW) has been established by the Nuclear Waste Policy Act (NWPA) Sec.8,(b) para(3)) [1982]. The materials have to be conditioned or treated, then packaged for disposal while meeting US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. The spent fuel and HLW located at the INEL will have to be put into a form and package that meets these regulatory criteria. The emphasis of Idaho Chemical Processing Plant (ICPP) future operations has shifted toward investigating, testing, and selecting technologies to prepare current and future spent fuels and waste for final disposal. This preparation for disposal may include mechanical, physical and/or chemical processes, and may differ for each of the various fuels and wastes.

  16. Waste Management Facilities Cost Information report for Greater-Than-Class C and DOE equivalent special case waste

    SciTech Connect

    Feizollahi, F.; Shropshire, D.

    1993-07-01

    This Waste Management Facility Cost Information (WMFCI) report for Greater-Than-Class C low-level waste (GTCC LLW) and DOE equivalent special case waste contains preconceptual designs and planning level life-cycle cost (PLCC) estimates for treatment, storage, and disposal facilities needed for management of GTCC LLW and DOE equivalent waste. The report contains information on 16 facilities (referred to as cost modules). These facilities are treatment facility front-end and back-end support functions (administration support, and receiving, preparation, and shipping cost modules); seven treatment concepts (incineration, metal melting, shredding/compaction, solidification, vitrification, metal sizing and decontamination, and wet/air oxidation cost modules); two storage concepts (enclosed vault and silo); disposal facility front-end functions (disposal receiving and inspection cost module); and four disposal concepts (shallow-land, engineered shallow-land, intermediate depth, and deep geological cost modules). Data in this report allow the user to develop PLCC estimates for various waste management options. A procedure to guide the U.S. Department of Energy (DOE) and its contractor personnel in the use of estimating data is also included in this report.

  17. Waste not - want not. DOE appropriate technology small grants program

    SciTech Connect

    Not Available

    1981-01-01

    The work reported was to look at various alternatives for local solid waste management and develop an implementation strategy for a resource conservation and recovery plan for the community of Berea, Kentucky. A library on recycling and conservation of resources was compiled, and state and local plans were examined. To get a better understanding of how the community would respond to a waste reduction and recycling program, a series of surveys was conducted. A community recycling project plan is proposed. (LEW)

  18. Properties and behavior of the platinum group metals in the glass resulting from the vitrification of simulated nuclear fuel reprocessing waste

    SciTech Connect

    Krause, C. ); Luckscheiter, B. )

    1991-12-01

    Two types of platinum group metal particles were found in borosilicate nuclear waste glasses: needle-shaped RuO{sub 2} particles and spherical PdRh{sub {ital x}}Te{sub {ital y}} alloys. They form a dense sediment of high electrical conductivity and relatively high viscosity at the bottom of the ceramic melting furnace. The sludge shows a non-Newtonian flow behavior. The viscosity and conductivity of the sludge depend not only on the platinum group metal content but also on the texture and morphology of the RuO{sub 2} particles. RuO{sub 2} forms long, needle-shaped crystals which are caused by alkalimolybdate salt melts that formed in the calcine layer. The salt melts oxidize the Ru present as small RuO{sub 2} particles after calcination to higher oxidation states. Ruthenium (VI) compounds are formed, presumably, which are not stable with respect to RuO{sub 2} under the melting conditions. RuO{sub 2} precipitates and crystallizes into long, needle-like particles.

  19. Plasma coal reprocessing

    NASA Astrophysics Data System (ADS)

    Messerle, V. E.; Ustimenko, A. B.

    2013-12-01

    Results of many years of investigations of plasma-chemical technologies for pyrolysis, hydrogenation, thermochemical preparation for combustion, gasification, and complex reprocessing of solid fuels and hydrocarbon gas cracking are represented. Application of these technologies for obtaining the desired products (hydrogen, industrial carbon, synthesis gas, valuable components of the mineral mass of coal) corresponds to modern ecological and economical requirements to the power engineering, metallurgy, and chemical industry. Plasma fuel utilization technologies are characterized by the short-term residence of reagents within a reactor and the high degree of the conversion of source substances into the desired products without catalyst application. The thermochemical preparation of the fuel to combustion is realized in a plasma-fuel system presenting a reaction chamber with a plasmatron; and the remaining plasma fuel utilization technologies, in a combined plasma-chemical reactor with a nominal power of 100 kW, whose zone of the heat release from an electric arc is joined with the chemical reaction zone.

  20. Proceedings of the tenth annual DOE low-level waste management conference: Session 4: Waste treatment minimization

    SciTech Connect

    Not Available

    1988-12-01

    This document contains eleven papers on various aspects of low-level radioactive waste management. Topics in this volume include: volume reduction plans; incentitives; and cost proposals; acid detoxification and reclamation; decontamination of lead; leach tests; West Valley demonstration project status report; and DOE's regional management strategies. Individual papers were processed separately for the data base. (TEM)

  1. Steam Reforming Technology for Denitration and Immobilization of DOE Tank Wastes

    SciTech Connect

    Mason, J. B.; McKibbin, J.; Ryan, K.; Schmoker, D.

    2003-02-26

    THOR Treatment Technologies, LLC (THOR) is a joint venture formed in June 2002 by Studsvik, Inc. (Studsvik) and Westinghouse Government Environmental Services Company LLC to further develop, market, and deploy Studsvik's patented THORSM non-incineration, steam reforming waste treatment technology. This paper provides an overview of the THORSM steam reforming process as applied to the denitration and conversion of Department of Energy (DOE) tank wastes to an immobilized mineral form. Using the THORSM steam reforming technology to treat nitrate containing tank wastes could significantly benefit the DOE by reducing capital and life-cycle costs, reducing processing and programmatic risks, and positioning the DOE to meet or exceed its stakeholder commitments for tank closure. Specifically, use of the THORSM technology can facilitate processing of up to 75% of tank wastes without the use of vitrification, yielding substantial life-cycle cost savings.

  2. Reprocessing and reuse of urological armamentarium: How correct are we!

    PubMed Central

    Raval, Krutik Vipulbhai; Chaudhari, Rajeev; Khant, Shahil Rameshbhai; Joglekar, Omkar; Patel, Dipen

    2017-01-01

    Healthcare is expensive for a large proportion of the population in spite of high per capita income and good health insurance penetration. In an effort to reduce cost of the procedure, reprocessing of devices was started in the late 1970s. Reprocessing practice includes various measures such as proper cleaning, disinfection, and sterilization procedures. As reprocessing is aimed at reducing cost, there is a potential risk of compromising patient safety due to cross contamination after inadequate sterilization. There is also risk of performance alteration of urological reprocessed devices during sterilization/disinfection processing. Therefore, there is a need for formulating proper guidelines to decide methods of reprocessing for various urological equipment. There is also need to discuss the problematic areas that urologists face and to find their solutions. A PubMed search was made in September 2016, using key words “reprocessing of medical devices,” “Single Use Devices,” “methods of reprocessing of devices in clinical practice,” “use of formalin chamber,” “urological disposable sterilization,” etc., After excluding duplicates, all English articles were reviewed by title and abstract. Full texts of selected articles were obtained, and these articles were cross-referenced to find any other related articles. All the articles were reviewed. A product can be reused if it can be economically reprocessed with validated protocols with preservation of its function. There is no reason to discard it after one use. This practice is useful for controlling economics of a urological case and to reduce the financial burden. Current Food and Drug Administration guidelines are stringent. The contamination described to test the sterilization process in the suggested guidelines actually does never exist in clinical practice. Therefore, new guidelines considering the clinical practice scenario are desirable. PMID:28479760

  3. Toxics turf fight: DOE, EPA battling over waste rules at nuke plants

    SciTech Connect

    Not Available

    1984-01-05

    The US DOE's contention that its nuclear weapons and research plants across the country are not subject to the hazardous waste disposal requirements of the Resource Conservation and Recovery Act will be tested in federal court if the Justice Dept. agrees with DOE. The pollution problems in Oak Ridge, TN are discussed.

  4. Review of DOE Waste Package Program. Semiannual report, October 1984-March 1985. Volume 8

    SciTech Connect

    Davis, M.S.

    1985-12-01

    A large number of technical reports on waste package component performance were reviewed over the last year in support of the NRC`s review of the Department of Energy`s (DOE`s) Environmental Assessment reports. The intent was to assess in some detail the quantity and quality of the DOE data and their relevance to the high-level waste repository site selection process. A representative selection of the reviews is presented for the salt, basalt, and tuff repository projects. Areas for future research have been outlined. 141 refs.

  5. ICPP radioactive liquid and calcine waste technologies evaluation. Interim report

    SciTech Connect

    Murphy, J.A.; Pincock, L.F.; Christiansen, I.N.

    1994-06-01

    The Department of Energy (DOE) has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage since 1951 and reprocessing since 1953. Until recently, the major activity of the ICPP has been the reprocessing of SNF to recover fissile uranium; however, changing world events have raised questions concerning the need to recover and recycle this material. In April 1992, DOE chose to discontinue reprocessing SNF for uranium recovery and shifted its focus toward the management and disposition of radioactive wastes accumulated through reprocessing activities. Currently, 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste) and 3,800 cubic meters (m{sup 3}) of calcine waste are in inventory at the ICPP. Legal drivers and agreements exist obligating the INEL to develop, demonstrate, and implement technologies for safe and environmentally sound treatment and interim storage of radioactive liquid and calcine waste. Candidate treatment processes and waste forms are being evaluated using the Technology Evaluation and Analysis Methodology (TEAM) Model. This process allows decision makers to (1) identify optimum radioactive waste treatment and disposal form alternatives; (2) assess tradeoffs between various optimization criteria; (3) identify uncertainties in performance parameters; and (4) focus development efforts on options that best satisfy stakeholder concerns. The Systems Analysis technology evaluation presented in this document supports the DOE in selecting the most effective radioactive liquid and calcine waste management plan to implement in compliance with established regulations, court orders, and agreements.

  6. Reprocessing Microflare Data

    NASA Technical Reports Server (NTRS)

    Ryan, James M.

    1999-01-01

    The report concerns work on detecting and cataloging solar microflares using an automated. An accompanying figure represents the solar microflare distribution during the period of April 1991 to November 1992, the height of solar activity after the launch of CGRO. It also shows the distribution extending below the distribution obtained at GSFC by manual means. We have implemented significant refinements in the search algorithm. The algorithm in its simplest form searches for transient events and based upon the distribution of the signal among the different BATSE detectors, we can assign it to be of solar origin if the signal distribution conforms to what one expects from a burst or transient from that direction. One of the major problems in an earlier effort was to search for microflares and large flares simultaneously. The requirement for a dynamic range of almost 10(exp 4) resulted in ambiguous identifications at the low side of the distribution. We have since restricted the search to events with peak count rates under 2000/s. Larger events are easily identified in the manual search, so we have chosen not to duplicate that work. The second problem was that missing counts existed below channel 0 in the BATSE Large Area Detector (LAD) data. These have been recovered and are now included in the search process. This provides data below 20 keV, and as we get closer to the thermal part of the spectrum, it provides greater sensitivity. The third problem was that too many BATSE detectors were used in the search. Detectors with pointing directions far from the Sun, although detecting the event, had poorly known responses. Detectors greater than approximately 60 degrees off the Sun are no longer included in the search process. By reducing the systematic errors with the large off-axis detectors we can conduct more rigorous statistical tests of a candidate event to ascertain whether it originated from the solar direction. We have reprocessed the period in the early mission

  7. Treatment of radioactive wastes from DOE underground storage tanks

    SciTech Connect

    Collins, J.L.; Egan, B.Z.; Spencer, B.B.; Chase, C.W.; Anderson, K.K.; Bell, J.T.

    1994-06-01

    Bench-scale batch tests have been conducted with sludge and supernate tank waste from the Melton Valley Storage Tank (MVST) Facility at Oak Ridge National Laboratory (ORNL) to evaluate separation technology process for use in a comprehensive sludge processing flow sheet as a means of concentrating the radionuclides and reducing the volumes of storage tank waste at national sites for final disposal. This paper discusses the separation of the sludge solids and supernate, the basic washing of the sludge solids, the acidic dissolution of the sludge solids, and the removal of the radionuclides from the supernate.

  8. THE EPA/DOE MINE WASTE TECHNOLOGY PROGRAM

    EPA Science Inventory

    Mining activities in the US (not counting coal) produce between 1-2B tons of mine waste annually. Since many of the ore mines involve sulfide minerals, the production of acid mine drainage (AMD) is a common problem from these abandoned mine sites. The combination of acidity, heav...

  9. THE EPA/DOE MINE WASTE TECHNOLOGY PROGRAM

    EPA Science Inventory

    Mining activities in the US (not counting coal) produce between 1-2B tons of mine waste annually. Since many of the ore mines involve sulfide minerals, the production of acid mine drainage (AMD) is a common problem from these abandoned mine sites. The combination of acidity, heav...

  10. Evaluation and selection of candidate high-level waste forms

    SciTech Connect

    Bernadzikowski, T. A.; Allender, J. S.; Butler, J. L.; Gordon, D. E.; Gould, Jr., T. H.; Stone, J. A.

    1982-03-01

    Seven candidate waste forms being developed under the direction of the Department of Energy's National High-Level Waste (HLW) Technology Program, were evaluated as potential media for the immobilization and geologic disposal of high-level nuclear wastes. The evaluation combined preliminary waste form evaluations conducted at DOE defense waste-sites and independent laboratories, peer review assessments, a product performance evaluation, and a processability analysis. Based on the combined results of these four inputs, two of the seven forms, borosilicate glass and a titanate based ceramic, SYNROC, were selected as the reference and alternative forms for continued development and evaluation in the National HLW Program. Both the glass and ceramic forms are viable candidates for use at each of the DOE defense waste-sites; they are also potential candidates for immobilization of commercial reprocessing wastes. This report describes the waste form screening process, and discusses each of the four major inputs considered in the selection of the two forms.

  11. Support for DOE program in mineral waste-form development

    SciTech Connect

    Palmour, H. III; Hare, T.M.; Russ, J.C.; Batchelor, A.D.; Paisley, M.J.; Freed, L.E.

    1982-09-01

    This research investigation relates to sintered simulation ceramic waste forms of the generic SYNROC compositional type. Though they have been formulated with simulated wastes only, they serve as prototypes for potential hot, processed, crystalline waste forms whose combined thermodynamic stability and physical integrity are considered to render them capable of long-term imobilization of high-level radwastes under deep geologic disposal conditions. The problems involved are nontrivial, largely because of the very complex nature of the radwastes: a typical waste stream would contain more than 31 cation species. When the stabilizing matrix constituents are included, the final batch composition must successfully account (and find substitutional homes for some 35 different cation species. One of the important objectives of this study thus has been to develop a computer-based method for simulating these complex ion substitutions, and for calculating the resultant phase demands and batch formulations. Primary goals of the study have been (1) use of that computer simulation capability to incorporate rationally the radwaste ions from a specific waste stream (PW-7a) into the available SYNROC lattice sites and (2) utilization of existing ceramic processing and sintering methodologies to assure (and to understand) the attainment of high density, fine microstructure, full phase development and other features of the sintered product which are known to relate directly to its integrity and leach resistance. Though improved resistance to leaching has been a continuing goal, time and budget constraints have precluded initiation of any leachability studies of these new compositions during this contract period. 27 references, 15 figures, 6 tables.

  12. Does industrial waste taxation contribute to reduction of landfilled waste? Dynamic panel analysis considering industrial waste category in Japan.

    PubMed

    Sasao, Toshiaki

    2014-11-01

    Waste taxes, such as landfill and incineration taxes, have emerged as a popular option in developed countries to promote the 3Rs (reduce, reuse, and recycle). However, few studies have examined the effectiveness of waste taxes. In addition, quite a few studies have considered both dynamic relationships among dependent variables and unobserved individual heterogeneity among the jurisdictions. If dependent variables are persistent, omitted variables cause a bias, or common characteristics exist across the jurisdictions that have introduced waste taxes, the standard fixed effects model may lead to biased estimation results and misunderstood causal relationships. In addition, most existing studies have examined waste in terms of total amounts rather than by categories. Even if significant reductions in total waste amounts are not observed, some reduction within each category may, nevertheless, become evident. Therefore, this study analyzes the effects of industrial waste taxation on quantities of waste in landfill in Japan by applying the bias-corrected least-squares dummy variable (LSDVC) estimators; the general method of moments (difference GMM); and the system GMM. In addition, the study investigates effect differences attributable to industrial waste categories and taxation types. This paper shows that industrial waste taxes in Japan have minimal, significant effects on the reduction of final disposal amounts thus far, considering dynamic relationships and waste categories. Copyright © 2014 Elsevier Ltd. All rights reserved.

  13. Application of a Plasma Mass Separator to Advanced LWR Spent Fuel Reprocessing

    SciTech Connect

    Freeman, Richard; Miller, Robert; Papay, Larry; Wagoner, John; Ahlfeld, Charles; Czerwinski, Ken

    2006-07-01

    The US Department of Energy (DOE) is investigating spent fuel reprocessing for the purposes of increasing the effective capacity of a deep geological repository, reducing the radiotoxicity of waste placed in the repository and conserving nuclear fuel resources. DOE is considering hydro-chemical processing of the spent fuel after cutting the fuel cladding and fuel dissolution in nitric acid. The front end process, known as UREX, is largely based on the PUREX process and extracts U, Tc as well as fission product gases. A number of additional processing steps have become known as UREX+. One of the steps includes a further chemical treatment of remove Cs and Sr to reduce repository heat load. Other steps include successive extraction of the actinides from residual fission products, including the lanthanides. The additional UREX+ processing renders the actinides suitable for burning as reactor fuel in an advanced reactor to convert actinides to shorter-lived fission products and to produce power. New methods for separating groups of elements by their atomic mass have been developed and can be exploited to enhance spent fuel reprocessing. These physical processes dry the waste streams so that they can be vaporized and singly ionized in plasma that is contained in longitudinal magnetic and perpendicular electric fields. Proper configuration of the fields causes the plasma to rapidly rotate and expel heavier mass ions at the center of the machine. Lower mass ions form closed orbits within the cylindrical plasma column and are transported to either end of the machine. This plasma mass separator was originally developed to reduce the mass of material that must be immobilized in borosilicate glass from DOE defense waste at former weapons production facilities. The plasma mass separator appears to be well-suited for processing the UREX raffinate and solids streams by exploiting the large atomic mass gap that exists between lanthanides (< {approx}180 amu) and actinides

  14. Waste Cleanup: Status and Implications of DOE’s Compliance Agreements

    DTIC Science & Technology

    2002-05-01

    Ridge, Tennessee; the Nevada Test Site, Nevada; and Los Alamos National Laboratory, New Mexico . Many other stakeholders are involved in the cleanup...requirement to transport a portion of its transuranic waste to New Mexico , allowing instead for disposal on-site. While these options could reduce treatment...treat, and dispose of covered mixed wastes at the laboratory (incorporates Site Treatment Plan) DOE; New Mexico Environment Department

  15. TECHNICAL ASSESSMENT OF FRACTIONAL CRYSTALLIZATION FOR TANK WASTE PRETREATMENT AT THE DOE HANFORD SITE

    SciTech Connect

    HAMILTON, D.W.

    2006-01-03

    Radioactive wastes from one hundred seventy-seven underground storage tanks in the 200 Area of the Department of Energy (DOE) Hanford Site in Washington State will be retrieved, treated and stored either on site or at an approved off-site repository. DOE is currently planning to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions, which would be treated and permanently disposed in separate facilities. A significant volume of the wastes in the Hanford tanks is currently classified as medium Curie waste, which will require separation and treatment at the Waste Treatment Plant (WTP). Because of the specific challenges associated with treating this waste stream, DOE EM-21 funded a project to investigate the feasibility of using fractional crystallization as a supplemental pretreatment technology. The two process requirements for fractional crystallization to be successfully applied to Hanford waste include: (1) evaporation of water from the aqueous solution to enrich the activity of soluble {sup 137}Cs, resulting in a higher activity stream to be sent to the WTP, and (2) separation of the crystalline salts that are enriched in sodium, carbonate, sulfate, and phosphate and sufficiently depleted in {sup 137}Cs, to produce a second stream to be sent to Bulk Vitrification. Phase I of this project has just been completed by COGEMA/Georgia Institute of Technology. The purpose of this report is to document an independent expert review of the Phase I results with recommendations for future testing. A team of experts with significant experience at both the Hanford and Savannah River Sites was convened to conduct the review at Richland, Washington the week of November 14, 2005.

  16. A Survey of Mixed-Waste HEPA Filters in the DOE Complex

    SciTech Connect

    Felicione, F. S.; Barber, D. B.; Carney, K. P.

    2002-02-28

    A brief investigation was made to determine the quantities of spent, mixed-waste HEPA filters within the DOE Complex. The quantities of both the mixed-waste filters that are currently being generated, as well as the legacy mixed-waste filters being stored and awaiting disposition were evaluated. Seven DOE sites representing over 89% of the recent HEPA filter usage were identified. These sites were then contacted to determine the number of these filters that were likely destined to become mixed waste and to survey the legacy-filter quantities. Inquiries into the disposition plans for the filters were also made. It was determined that the seven sites surveyed possess approximately 500 m3 of legacy mixed-waste HEPA filters that will require processing, with an annual generation rate of approximately 25 m3. No attempt was made to extrapolate the results of this survey to the entire DOE Complex. These results were simply considered to be the lower bound of the totality of mixed-waste HEPA filters throughout the Complex. The quantities determined encourage the development of new treatment technologies for these filters, and provide initial data on which an appropriate capacity for a treatment process may be based.

  17. Development of Advanced Electrochemical Emission Spectroscopy for Monitoring Corrosion in Simulated DOE Liquid Waste

    SciTech Connect

    MacDonal, Digby D.; Marx, Brian M.; Ahn, Sejin; Ruiz, Julio de; Soundararajan, Balaji; Smith, Morgan; Coulson, Wendy

    2005-06-15

    Various forms of general and localized corrosion represent principal threats to the integrity of DOE liquid waste storage tanks. These tanks, which are of a single wall or double wall design, depending upon their age, are fabricated from welded carbon steel and contain a complex waste-form comprised of NaOH and NaNO3, along with trace amounts of phosphate, sulfate, carbonate, and chloride. Because waste leakage can have a profound environmental impact, considerable interest exists in predicting the accumulation of corrosion damage, so as to more effectively schedule maintenance and repair.

  18. DOE assay methods used for characterization of contact-handled transuranic waste

    SciTech Connect

    Schultz, F.J. ); Caldwell, J.T. )

    1991-08-01

    US Department of Energy methods used for characterization of contact-handled transuranic (CH-TRU) waste prior to shipment to the Waste Isolation Pilot Plant (WIPP) are described and listed by contractor site. The methods described are part of the certification process. All CH-TRU waste must be assayed for determination of fissile material content and decay heat values prior to shipment and prior to storage on-site. Both nondestructive assay (NDA) and destructive assay methods are discussed, and new NDA developments such as passive-action neutron (PAN) crate counter improvements and neutron imaging are detailed. Specifically addressed are assay method physics; applicability to CH-TRU wastes; calibration standards and implementation; operator training requirements and practices; assay procedures; assay precision, bias, and limit of detection; and assay limitation. While PAN is a new technique and does not yet have established American Society for Testing and Materials. American National Standards Institute, or Nuclear Regulatory Commission guidelines or methods describing proper calibration procedures, equipment setup, etc., comparisons of PAN data with the more established assay methods (e.g., segmented gamma scanning) have demonstrated its reliability and accuracy. Assay methods employed by DOE have been shown to reliable and accurate in determining fissile, radionuclide, alpha-curie content, and decay heat values of CH-TRU wastes. These parameters are therefore used to characterize packaged waste for use in certification programs such as that used in shipment of CH-TRU waste to the WIPP. 36 refs., 10 figs., 7 tabs.

  19. Operating cost guidelines for benchmarking DOE thermal treatment systems for low-level mixed waste

    SciTech Connect

    Salmon, R.; Loghry, S.L.; Hermes, W.H.

    1994-11-01

    This report presents guidelines for estimating operating costs for use in benchmarking US Department of Energy (DOE) low-level mixed waste thermal treatment systems. The guidelines are based on operating cost experience at the DOE Toxic Substances Control Act (TSCA) mixed waste incinerator at the K-25 Site at Oak Ridge. In presenting these guidelines, it should be made clear at the outset that it is not the intention of this report to present operating cost estimates for new technologies, but only guidelines for estimating such costs.

  20. Steps of Reprocessing and Equipments

    PubMed Central

    Lee, Yong Kook

    2013-01-01

    With the increasing interest in endoscopy and the rising number of endoscopic examinations in hospitals, the importance of endoscopic reprocessing is also increasing. Cure facilities that are understaffed and ill-equipped are trying to cope with the problems of insufficient cleaning and high infection risks. To prevent endoscopy-associated infection, the endoscope cleaning, and disinfection guidelines prepared by the Korean Society of Gastrointestinal Endoscopy must be followed. In this review, the steps of endoscopic reprocessing and the equipments required in each step are discussed. PMID:23767039

  1. What life-cycle assessment does and does not do in assessments of waste management

    SciTech Connect

    Ekvall, Tomas Assefa, Getachew; Bjoerklund, Anna; Eriksson, Ola; Finnveden, Goeran

    2007-07-01

    In assessments of the environmental impacts of waste management, life-cycle assessment (LCA) helps expanding the perspective beyond the waste management system. This is important, since the indirect environmental impacts caused by surrounding systems, such as energy and material production, often override the direct impacts of the waste management system itself. However, the applicability of LCA for waste management planning and policy-making is restricted by certain limitations, some of which are characteristics inherent to LCA methodology as such, and some of which are relevant specifically in the context of waste management. Several of them are relevant also for other types of systems analysis. We have identified and discussed such characteristics with regard to how they may restrict the applicability of LCA in the context of waste management. Efforts to improve LCA with regard to these aspects are also described. We also identify what other tools are available for investigating issues that cannot be adequately dealt with by traditional LCA models, and discuss whether LCA methodology should be expanded rather than complemented by other tools to increase its scope and applicability.

  2. What life-cycle assessment does and does not do in assessments of waste management.

    PubMed

    Ekvall, Tomas; Assefa, Getachew; Björklund, Anna; Eriksson, Ola; Finnveden, Göran

    2007-01-01

    In assessments of the environmental impacts of waste management, life-cycle assessment (LCA) helps expanding the perspective beyond the waste management system. This is important, since the indirect environmental impacts caused by surrounding systems, such as energy and material production, often override the direct impacts of the waste management system itself. However, the applicability of LCA for waste management planning and policy-making is restricted by certain limitations, some of which are characteristics inherent to LCA methodology as such, and some of which are relevant specifically in the context of waste management. Several of them are relevant also for other types of systems analysis. We have identified and discussed such characteristics with regard to how they may restrict the applicability of LCA in the context of waste management. Efforts to improve LCA with regard to these aspects are also described. We also identify what other tools are available for investigating issues that cannot be adequately dealt with by traditional LCA models, and discuss whether LCA methodology should be expanded rather than complemented by other tools to increase its scope and applicability.

  3. Waste management policy development from the AEC to the DOE

    SciTech Connect

    Dieckhoner, J.E.

    1996-03-01

    The goal of the federal government`s policy for the management of radioactive waste from its nuclear materials and weapon programs was, and is, to protect the health and safety of those working in these programs, the public and the environment. What has changed since the days of the Manhattan Project are the specific technical solutions pursued, the implementation schedules, and the projected cost to the American taxpayer. Although some of these changes are due to nontechnical and political issues, many have been the result of knowledge gained from research and development programs, scientific over-optimism, and an unrealistic appreciation of the cost and schedule impacts of required stakeholder involvement. The goal of this article is to review the development of waste management policy from the early days of the Atomic Energy Commission to the present. The current generation of policymakers, and their critics, will be greatly aided in their tasks if they have a general understanding of what events have occurred over the past half century that have led us to the current situation.

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

    SciTech Connect

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

    1994-11-01

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

  5. 40 CFR 60.1545 - Does this subpart directly affect municipal waste combustion unit owners and operators in my State?

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... municipal waste combustion unit owners and operators in my State? 60.1545 Section 60.1545 Protection of... NEW STATIONARY SOURCES Emission Guidelines and Compliance Times for Small Municipal Waste Combustion... municipal waste combustion unit owners and operators in my State? (a) No, this subpart does not...

  6. 40 CFR 60.1545 - Does this subpart directly affect municipal waste combustion unit owners and operators in my State?

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... municipal waste combustion unit owners and operators in my State? 60.1545 Section 60.1545 Protection of... NEW STATIONARY SOURCES Emission Guidelines and Compliance Times for Small Municipal Waste Combustion... municipal waste combustion unit owners and operators in my State? (a) No, this subpart does not...

  7. 40 CFR 60.1545 - Does this subpart directly affect municipal waste combustion unit owners and operators in my State?

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... municipal waste combustion unit owners and operators in my State? 60.1545 Section 60.1545 Protection of... NEW STATIONARY SOURCES Emission Guidelines and Compliance Times for Small Municipal Waste Combustion... municipal waste combustion unit owners and operators in my State? (a) No, this subpart does not directly...

  8. 40 CFR 60.1545 - Does this subpart directly affect municipal waste combustion unit owners and operators in my State?

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... municipal waste combustion unit owners and operators in my State? 60.1545 Section 60.1545 Protection of... NEW STATIONARY SOURCES Emission Guidelines and Compliance Times for Small Municipal Waste Combustion... municipal waste combustion unit owners and operators in my State? (a) No, this subpart does not directly...

  9. 40 CFR 60.1545 - Does this subpart directly affect municipal waste combustion unit owners and operators in my State?

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... municipal waste combustion unit owners and operators in my State? 60.1545 Section 60.1545 Protection of... NEW STATIONARY SOURCES Emission Guidelines and Compliance Times for Small Municipal Waste Combustion... municipal waste combustion unit owners and operators in my State? (a) No, this subpart does not directly...

  10. Incineration of DOE offsite mixed waste at the Idaho National Engineering and Environmental Laboratory

    SciTech Connect

    Harris, J.D.; Harvego, L.A.; Jacobs, A.M.; Willcox, M.V.

    1998-01-01

    The Waste Experimental Reduction Facility (WERF) incinerator at the Idaho National Engineering and Environmental Laboratory (INEEL) is one of three incinerators in the US Department of Energy (DOE) Complex capable of incinerating mixed low-level waste (MLLW). WERF has received MLLW from offsite generators and is scheduled to receive more. The State of Idaho supports receipt of offsite MLLW waste at the WERF incinerator within the requirements established in the (INEEL) Site Treatment Plan (STP). The incinerator is operating as a Resource Conservation and Recovery Act (RCRA) Interim Status Facility, with a RCRA Part B permit application currently being reviewed by the State of Idaho. Offsite MLLW received from other DOE facilities are currently being incinerated at WERF at no charge to the generator. Residues associated with the incineration of offsite MLLW waste that meet the Envirocare of Utah waste acceptance criteria are sent to that facility for treatment and/or disposal. WERF is contributing to the treatment and reduction of MLLW in the DOE Complex.

  11. Classic Nuclear Fuel Reprocessing Flowsheet

    SciTech Connect

    Fallgren, Andrew James

    2015-02-13

    This is a flowsheet as well as a series of subsheets to be used for discussion on the standard design of a reprocessing plant. This flowsheet consists of four main sections: offgas handling, separations, solvent wash, and acid recycle. As well as having the main flowsheet, subsections have been broken off into their own sheets to provide for larger font and ease of printing.

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

    SciTech Connect

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

    1987-10-01

    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.

  13. Does improved waste treatment have demonstrable biological benefits?

    NASA Astrophysics Data System (ADS)

    Seagle, Henry H.; Hendricks, Albert C.; Cairns, John

    1980-01-01

    Since 1972, 10 benthic surveys and 9 static fish bioassays have been conducted to assess the impact of AVTEX Fibers, Inc. effluent on the lower South Fork of the Shenandoah River. AVTEX (formerly FMC Corp.) is a rayon and polyester fibers plant located in Front Royal, Virginia. Benthic samples were taken at four stations, one above and three below the plant discharges. Single surveys in 1972 and 1973 indicated a severe impact on the benthic community along the right side of the river, below the plant, as a result of the channelized effluent. Diversity values (¯ d) were low (0 2.42) and numbers of taxa and organisms were reduced. A fish bioassay in 1973 indicated the effluent to be acutely toxic at the 34.5% level (mixture of effluent and river water). In early 1974, FMC Corp. constructed an activated sludge treatment system to reduce BOD and supplement the neutralization and chemical precipitation (zinc hydroxide and liquid-solid separation) facilities that had been used to treat waste waters since 1948. After the new equipment was placed in operation, the previously stressed area became more stable. In 1975 and 1976 the stressed area exhibited greater ¯ d values (1.19 3.39) and an increased number of taxa and organisms. Bioassays showed the effluent to be acutely toxic to fish only once since 1973. The major improvements in the effluent were a 70% reduction in BOD5 and a 60% reduction in the amount of zinc entering the river. Community conditions in 1977 indicated a partial remission of improvement, probably due to drought conditions. The rehabilitation of damaged ecosystems is a process important to all biologists. An important factor in encouraging industry to participate in this activity is evidence that improved waste treatment will often have demonstrable biological benefits rather soon. As data accumulate on the recovery process it may be possible to predict the degree of rehabilitation and time required more precisely.

  14. Review of DOE Planned Change Request for Shielded Containers for Remote-Handled Transuranic Waste

    EPA Pesticide Factsheets

    This report summarizes SC&A's review of the planned change request (PCR) submitted by DOE to EPA proposing the disposal of some remote-handled (RH) transuranic (TRU) waste in shielded containers on the floor of the disposal rooms at WIPP

  15. Radioactive Iodine and Krypton Control for Nuclear Fuel Reprocessing Facilities

    SciTech Connect

    Soelberg, Nicolas R.; Garn, Troy; Greenhalgh, Mitchell; Law, Jack; Jubin, Robert T.; Strachan, Denis M.; Thallapally, Praveen K.

    2013-07-22

    Nuclear fission results in the production of fission products and activation products, some of which tend to be volatile during used fuel reprocessing. These can evolve in volatile species in the reprocessing facility off-gas streams, depending on the separations and reprocessing technologies that are used. Radionuclides that have been identified as “volatile radionuclides” are noble gases (most notably isotopes of Kr and Xe); 3H; 14C; and 129I. Radionuclides that tend to form volatile species that evolve into reprocessing facility off-gas systems are more challenging to efficiently control compared to radionuclides that tend to stay in solid or liquid phases. Future used fuel reprocessing facilities in the United States can require efficient capture of some volatile radionuclides in their off-gas streams to meet regulatory emission requirements. In aqueous reprocessing, these radionuclides are most commonly expected to evolve into off-gas streams in tritiated water [3H2O (T2O) and 3HHO (THO)], radioactive CO2, noble gases, and gaseous HI, I2, or volatile organic iodides. The fate and speciation of these radionuclides from a non-aqueous fuel reprocessing facility is less well known at this time, but active investigations are in progress. An Off-Gas Sigma Team was formed in late FY 2009 to integrate and coordinate the Fuel Cycle Research and Development (FCR&D) activities directed towards the capture and sequestration of the these volatile radionuclides (Jubin 2012a). The Sigma Team concept was envisioned to bring together multidisciplinary teams from across the DOE complex that would work collaboratively to solve the technical challenges and to develop the scientific basis for the capture and immobilization technologies such that the sum of the efforts was greater than the individual parts. The Laboratories currently participating in this effort are Argonne National Laboratory (ANL), Idaho National Laboratory (INL), Oak Ridge National Laboratory (ORNL), Pacific

  16. Consolidated Fuel-Reprocessing Program. Progress report, April 1 to June 30, 1983

    SciTech Connect

    Not Available

    1983-08-01

    All research and development on fuel reprocessing in the United States is managed under the Consolidated Fuel Reprocessing Program. Technical progress is reported in overview fashion. Conceptual studies for the proposed Breeder Reprocessing Engineering Test (BRET) have continued. Studies to date have confirmed the feasibility of modifying an existing DOE facility at Hanford, Washington. A study to measure the extent of plutonium polymerization during steam-jet transfers of nitric acid solutions indicated polymer would appear only after several successive transfers at temperatures of 75/sup 0/C or higher. Fast-Flux Test Facility fuel was processed for the first time in the Solvent Extraction Test Facility. Studies of krypton release from pulverized sputter-deposited Ni-Y-Kr matrices have shown that the release rate is inversely proportional to the particle radius at 200/sup 0/C. Preparation of the initial 500-g batch of mixed oxide gel-spheres was completed. Fabrication processing at HEDL of mixed oxide gel-spheres (DIPRES process) was initiated. Operational testing of both 8 packs of the centrifugal contactor has been completed. Fabrication of both the prototypical disassembly system and the prototypical shear system has been initiated. Planning for FY 1984 installation and modification work in the integrated equipment list facility was completed. Acceptance tests of the original Integrated Process Demonstration system have been completed. Instrumentation and controls work with the prototype multiwavelength uranium photometer was successful and has been expanded to continuously and simultaneously monitor three process streams (raffinate, aqueous feed, and organic strip) in the secondary extraction cycle. Major efforts of the environmental, safeguards, and waste management areas were directed toward providing data for BRET.

  17. SHARING AND DEPLOYING INNOVATIVE INFORMATION TECHNOLOGY SOLUTIONS TO MANAGE WASTE ACROSS THE DOE COMPLEX

    SciTech Connect

    Crolley, R.; Thompson, M.

    2011-01-31

    There has been a need for a faster and cheaper deployment model for information technology (IT) solutions to address waste management needs at US Department of Energy (DOE) complex sites for years. Budget constraints, challenges in deploying new technologies, frequent travel, and increased job demands for existing employees have prevented IT organizations from staying abreast of new technologies or deploying them quickly. Despite such challenges, IT organizations have added significant value to waste management handling through better worker safety, tracking, characterization, and disposition at DOE complex sites. Systems developed for site-specific missions have broad applicability to waste management challenges and in many cases have been expanded to meet other waste missions. Radio frequency identification (RFID) and global positioning satellite (GPS)-enabled solutions have reduced the risk of radiation exposure and safety risks. New web-based and mobile applications have enabled precision characterization and control of nuclear materials. These solutions have also improved operational efficiencies and shortened schedules, reduced cost, and improved regulatory compliance. Collaboration between US Department of Energy (DOE) complex sites is improving time to delivery and cost efficiencies for waste management missions with new information technologies (IT) such as wireless computing, global positioning satellite (GPS), and radio frequency identification (RFID). Integrated solutions developed at separate DOE complex sites by new technology Centers of Excellence (CoE) have increased material control and accountability, worker safety, and environmental sustainability. CoEs offer other DOE sister sites significant cost and time savings by leveraging their technology expertise in project scoping, implementation, and ongoing operations.

  18. Nuclear Waste Management - A Need to Ensure that the Waste Decays While the Knowledge Does Not

    SciTech Connect

    Biedscheid, J.; Devarakonda, M.

    2006-07-01

    The unique time scales associated with nuclear waste management require active work to preserve the knowledge associated with programs that can often span long time frames. For example, repository programs are designed and developed for the safe containment of radionuclides with half-lives on the order of 24,000 years (for Pu-239). Performance assessment studies are required to show, by predictive modeling, that these repositories can safely contain the waste for tens of thousands of years, with one million years a reality in the debate. Development of successful repository programs can span several decades. Once operational, a repository is expected to function until closure for a period of 30 to 40 years. Yet, the decision makers at all levels in such a repository program exert influence and authority over much shorter periods of time (e.g., four years for executive appointments). A discontinuous decision-making process and the associated potential for loss of critical program knowledge are challenges that nuclear waste management programs face in various countries. Due to the disparity of the time frames for the project and associated decision-making, a proactive program to preserve original and developing knowledge for technical and programmatic decisions is vital to the success of any nuclear waste management program. While the mechanisms for knowledge preservation may be program specific, the need for programs to acknowledge and implement a methodology for maintaining program intelligence is collectively shared. (authors)

  19. Consolidated Fuel Reprocessing Program. Progress report, January 1-March 31, 1985

    SciTech Connect

    Not Available

    1985-04-01

    The DOE has concentrated all US research and development on fuel reprocessing into one major program - the Consolidated Fuel Reprocessing Program (CFRP) - under the management of the Oak Ridge National Laboratory and the Oak Ridge Operations Office. Other major program participants are GA Technologies, Inc., where reprocessing research and development on the HTGR fuel cycle are done, and the Hanford Engineering Development Laboratory (HEDL). The coverage is generally overview in nature. Experimental details and data have been limited to (1) make the report more concise and (2) meet the requirements which would qualify the report for unrestricted distribution in the open literature. All research and development on civilian power reactor fuel reprocessing in the United States is managed under the Consolidated Fuel Reprocessing Program (CFRP) centered at Oak Ridge National Laboratory (ORNL). Technical progress is reported in overview fashion in this series of quarterly progress reports.

  20. DOE's Notification of Planned Change to the EPA 40 CFR Part 194 Certification of the Waste Isolation Pilot Plant: Remote-Handled Transuranic Waste Characterization Plan

    EPA Pesticide Factsheets

    The U.S. Department of Energy's Carlsbad Field Office (DOE/CBFO) provided the U.S. Environmental Protection Agency (EPA) this Notification of Planned Change to accept remote-handled (RH) transuranic (TRU) waste at the Waste Isolation Pilot Plant (WIPP).

  1. Evaluation and compilation of DOE waste package test data; Biannual report, August 1988--January 1989: Volume 6

    SciTech Connect

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

    1990-11-01

    This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six-month period August 1988 through January 1989. Included are reviews of related materials research and plans, activities for the DOE Materials Characterization Center, information on the Yucca Mountain Project, and other information regarding supporting research and special assistance. NIST comments are given on the Yucca Mountain Consultation Draft Site Characterization Plan (CDSCP) and on the Waste Compliance Plan for the West Valley Demonstration Project (WVDP) High-Level Waste (HLW) Form. 3 figs.

  2. Review of Corrosion Inhibition in High Level Radioactive Waste Tanks in the DOE Complex

    SciTech Connect

    Subramanian, K.H.

    2004-03-08

    Radioactive waste is stored in underground storage tanks at the Department of Energy (DOE) Savannah River Site (SRS). The waste tanks store supernatant liquid salts, consisting primarily of sodium nitrate, sodium nitrite, sodium hydroxide, and sludge. An assessment of the potential degradation mechanisms of the high level waste (HLW) tanks determined that nitrate- induced pitting corrosion and stress corrosion cracking were the two most significant degradation mechanisms. Controls on the solution chemistry (minimum nitrite and hydroxide concentrations) are in place to prevent the initiation and propagation of pitting and stress corrosion cracking in the tanks. These controls are based upon a series of experiments performed using simulated solutions on materials used for construction of the tanks. The technical bases and evolution of these controls is presented.

  3. Potential dispositioning flowsheets for ICPP SNF and wastes

    SciTech Connect

    Olson, A.L.; Anderson, P.A.; Bendixsen, C.L.

    1995-11-01

    The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1953. This activity resulted mainly in the recovery of uranium and the management of the resulting wastes. The acidic radioactive high-level liquid waste was routinely stored in stainless steel tanks and then calcined to form a dry granular solid. The calcine is stored in stainless steel bins that are housed in underground concrete vaults. In April 1992, the DOE discontinued the practice of reprocessing irradiated nuclear fuels. This decision has left a legacy of 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3800 cubic meters of calcine waste, and 289 metric tons of heavy metal within unprocessed spent nuclear fuel (SNF) left in inventory at the ICPP. The nation`s radioactive waste policy has been established by the Nuclear Waste Policy Act (NWPA), which requires the final disposal of SNF and radioactive waste in accordance with US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) standards. In accordance with these regulations and other legal agreements between the State of Idaho and the DOE, the DOE must, among other requirements, (1) complete a final Environmental Impact Statement by April 30, 1995, (2) evaluate and test sodium-bearing waste pre-treatment technologies, (3) select the sodium-bearing and calcine waste pre-treatment technology, if necessary, by June 1, 1995, and (4) select a technology for converting calcined waste into an appropriate disposal form by June 1, 1995.

  4. Chemical basis for pyrochemical reprocessing of nuclear fuel

    SciTech Connect

    Ackerman, J.P. )

    1991-01-01

    The integral fast reactor (IFR) is an advanced breeder reactor concept that includes on-site reprocessing of spent fuel and wastes. Spent metallic fuel from the IFR is separated from fission products and cladding, and wastes are put into acceptable forms by use of a compact pyrochemical process based on partition of fuel and wastes between molten salt and liquid metal. To minimize reagent usage and, consequently, waste volume, electrotransport between metal phases is used extensively for feed dissolution and product recovery, but chemical oxidation and reduction are required for some operations. This paper describes the processes that are used and presents the chemical theory that was developed for quantitatively predicting the results of both chemical and electrotransport operations.

  5. The Effect of Reprocessing on the Tensile Properties of Composites

    NASA Astrophysics Data System (ADS)

    Bodur, Mehmet Safa; Bakkal, Mustafa; Berkalp, Omer Berk; Sadikoglu, Telem Gok

    2011-01-01

    In this study, waste cotton fabric reinforced polymer matrix composite material has been manufactured by a custom made recycling extruder. Composites with different reinforcement ratios as 12,5%wt ( 12,5%wtRPE ) and 25%wt ( 25%wtRPE ) were tested for their mechanical properties such as tensile strength and young's modulus. The material was then granulated down to the size enough to be used in the extrusion process in order to observe the effects of reprocessing. Reprocessing leads to improve Tensile Strength of composite materials and slows down the reduction of tensile strength of polyethylene. It was observed that composite materials were highly affected by the fiber orientation and acts as anisotropic material under the load.

  6. DEMONSTRATiON OF A SUBSURFACE CONTAINMENT SYSTEM FOR INSTALLATION AT DOE WASTE SITES

    SciTech Connect

    Thomas J. Crocker; Verna M. Carpenter

    2003-05-21

    Between 1952 and 1970, DOE buried mixed waste in pits and trenches that now have special cleanup needs. The disposal practices used decades ago left these landfills and other trenches, pits, and disposal sites filled with three million cubic meters of buried waste. This waste is becoming harmful to human safety and health. Today's cleanup and waste removal is time-consuming and expensive with some sites scheduled to complete cleanup by 2006 or later. An interim solution to the DOE buried waste problem is to encapsulate and hydraulically isolate the waste with a geomembrane barrier and monitor the performance of the barrier over its 50-yr lifetime. The installed containment barriers would isolate the buried waste and protect groundwater from pollutants until final remediations are completed. The DOE has awarded a contract to RAHCO International, Inc.; of Spokane, Washington; to design, develop, and test a novel subsurface barrier installation system, referred to as a Subsurface Containment System (SCS). The installed containment barrier consists of commercially available geomembrane materials that isolates the underground waste, similar to the way a swimming pools hold water, without disrupting hazardous material that was buried decades ago. The barrier protects soil and groundwater from contamination and effectively meets environmental cleanup standards while reducing risks, schedules, and costs. Constructing the subsurface containment barrier uses a combination of conventional and specialized equipment and a unique continuous construction process. This innovative equipment and construction method can construct a 1000-ft-long X 34-ft-wide X 30-ft-deep barrier at construction rates to 12 Wday (8 hr/day operation). Life cycle costs including RCRA cover and long-term monitoring range from approximately $380 to $590/cu yd of waste contained or $100 to $160/sq ft of placed barrier based upon the subsurface geology surrounding the waste. Project objectives for Phase I

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

    SciTech Connect

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

    1988-08-01

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

  8. Does recyclable separation reduce the cost of municipal waste management in Japan?

    PubMed

    Chifari, Rosaria; Lo Piano, Samuele; Matsumoto, Shigeru; Tasaki, Tomohiro

    2017-02-01

    Municipal solid waste (MSW) management is a system involving multiple sub-systems that typically require demanding inputs, materials and resources to properly process generated waste throughput. For this reason, MSW management is generally one of the most expensive services provided by municipalities. In this paper, we analyze the Japanese MSW management system and estimate the cost elasticity with respect to the waste volumes at three treatment stages: collection, processing, and disposal. Although we observe economies of scale at all three stages, the collection cost is less elastic than the disposal cost. We also examine whether source separation at home affects the cost of MSW management. The empirical results show that the separate collection of the recyclable fraction leads to reduced processing costs at intermediate treatment facilities, but does not change the overall waste management cost. Our analysis also reveals that the cost of waste management systems decreases when the service is provided by private companies through a public tender. The cost decreases even more when the service is performed under the coordination of adjacent municipalities.

  9. Process waste assessment approach, training and technical assistance for DOE contractors; FY93 report, ADS {number_sign}35303C

    SciTech Connect

    Pemberton, S

    1994-03-01

    The Department of Energy (DOE) and its contractors are faced with a large waste management problem as are other industries. One of the tools used in a successful waste minimization pollution prevention (WMin/P2) program is a process waste assessment (PWA). The purpose of this project was to share the Kansas City Plant`s (KCP`s) PWA expertise with other DOE personnel and DOE contractors. This consisted of two major activities: (1) The KCP`s PWA graded approach methodology was modified with the assistance of DOE/Defense Program`s laboratories, and (2) PWA training and technical assistance were provided to interested DOE personnel and DOE contractors. This report documents the FY93 efforts, lesson learned, and future plans for both PWA-related activities.

  10. The WISE (Waste-free, Intrinsically Safe, and Efficient) Nuclear Plant Concept

    SciTech Connect

    Slessarev, Igor; Palmiotti, G.; Salvatores, M.; Berthou, V.

    2002-07-01

    A new concept WISE (Waste-free, Intrinsically Safe, and Efficient) for a 'clean' nuclear plant is proposed. WISE does not perturb the radiological equilibrium minimizing TRU and LLFP wastes. WISE is based on mobile fuel systems and on a feed-and-bleed fuel cycle strategy. Irradiated fuel together with fission products always remains inside of WISE core. This implies no release of waste until fuel reserves are available. After the fuel reserves are exhausted, an on-line reprocessing technology can be applied in order to minimize waste. (authors)

  11. Commercial Nuclear Reprocessing in the United States

    SciTech Connect

    Sherrill, Charles Leland; Balatsky, Galya Ivanovna

    2015-09-09

    The short presentation outline: Reprocessing Overview; Events leading up to Carter’s Policy; Results of the decision; Policy since Nuclear Nonproliferation Act. Conclusions reached: Reprocessing ban has become an easy and visible fix to the public concern about proliferation, but has not completely stopped proliferation; and, Reprocessing needs to become detached from political considerations, so technical research can continue, regardless of the policy decisions we decide to take.

  12. Clinical Practice Guidelines for Endoscope Reprocessing

    PubMed Central

    Oh, Hyun Jin

    2015-01-01

    Gastrointestinal endoscopy is effective and safe for the screening, diagnosis, and treatment of gastrointestinal disease. However, issues regarding endoscope-transmitted infections are emerging. Many countries have established and continuously revise guidelines for endoscope reprocessing in order to prevent infections. While there are common processes used in endoscope reprocessing, differences exist among these guidelines. It is important that the reprocessing of gastrointestinal endoscopes be carried out in accordance with the recommendations for each step of the process. PMID:26473117

  13. Consolidated Fuel Reprocessing Program. National Program Plan FY 1984

    SciTech Connect

    Not Available

    1984-01-01

    This FY 1984 National Program Plan for the Consolidated Fuel Reprocessing Program (CFRP) provides specific guidance from the Department of Energy (DOE) for FY 1984 CFRP activities and a baseline for activities in future years. This is the second issue of the Plan, which is updated anually and summarizes program objectives, summary plans and schedules, budget allocations, contractor involvement, and interfaces with other research programs. The National Program Plan is a controlling document for the Technical Program Plan, which is prepared annually by the CFRP at Oak Ridge National Laboratory (ORNL) and is one of the hierarchical group of planning documents. The CFRP is a part of the DOE's program of research and development (R and D) on nuclear fission systems and is directed by the LMFBR Fuel Cycle Projects Division, Office of Spent Fuel Management and Reprocessing Systems at DOE-Headquarters through the Oak Ridge Operations Office (ORO). The strategy of the program is to maintain the capability to commit to a breeder option through a strong R and D program on breeder reprocessing and alternate fuels and fuel cycles in order to achieve operating and economic advantages.

  14. Consolidated Fuel Reprocessing Program: National Program Plan, FY 1983

    SciTech Connect

    Not Available

    1983-01-01

    This FY 1983 National Program Plan for the Consolidated Fuel Reprocessing Program (CFRP) provides specific guidance from the Department of Energy (DOE) for FY 1983 CFRP activities and a baseline for future year activities. This initial issue of the Plan, which will be updated annually, summarizes program objectives, summary plans and schedules, budget allocations, contractor involvement, and interfaces with other research programs. The National Program Plan is a controlling document for the Technical Program Plan, which is prepared annually by the CFRP at ORNL and is one of a hierarchical group of planning documents. The CFRP is a part of the DOE's program of research and development (R and D) on nuclear fission systems and is directed by the LMFBR Fuel Cycle Projects Division, Office of Spent Fuel Management and Reprocessing Systems at DOE-Headquarters through the Oak Ridge Operations Office (ORO). The strategy of the program is to maintain the capability to commit to a breeder option through a strong R and D program on breeder reprocessing and alternate fuels and fuel cycles to achieve operating and economic advantages.

  15. Recent Update of Gastrointestinal Endoscope Reprocessing

    PubMed Central

    Hong, Kyong Hee

    2013-01-01

    As infection-related issues have become one of the most important concerns in endoscopy centers, proper reprocessing of endoscopes has attracted great interest. Compliance with established guidelines for reprocessing is critical to prevent pathogen transmission. However, hospital compliance with guidelines has not been satisfactory. To increase compliance, efforts have focused on developing new and more innovative disinfectants and an automated endoscope reprocessor. Reprocessing must be performed by appropriately trained personnel and regular monitoring of reprocessing is essential for quality assurance to improve compliance. PMID:23767038

  16. Endoscope Reprocessing: Update on Controversial Issues

    PubMed Central

    Choi, Hyun Ho

    2015-01-01

    Several issues concerning endoscope reprocessing remain unresolved based on currently available data. Thus, further studies are required to confirm standard practices including safe endoscope shelf life, proper frequency of replacement of some accessories including water bottles and connecting tubes, and microbiological surveillance testing of endoscopes after reprocessing. The efficacy and cost-effectiveness of newer technology that allows automated cleaning and disinfection is one such controversial issue. In addition, there are no guidelines on whether delayed reprocessing and extended soaking may harm endoscope integrity or increase the bioburden on the external or internal device surfaces. In this review, we discuss the unresolved and controversial issues regarding endoscope reprocessing. PMID:26473115

  17. Performance Assessment of a Generic Repository in Bedded Salt for DOE-Managed Nuclear Waste

    NASA Astrophysics Data System (ADS)

    Stein, E. R.; Sevougian, S. D.; Hammond, G. E.; Frederick, J. M.; Mariner, P. E.

    2016-12-01

    A mined repository in salt is one of the concepts under consideration for disposal of DOE-managed defense-related spent nuclear fuel (SNF) and high level waste (HLW). Bedded salt is a favorable medium for disposal of nuclear waste due to its low permeability, high thermal conductivity, and ability to self-heal. Sandia's Generic Disposal System Analysis framework is used to assess the ability of a generic repository in bedded salt to isolate radionuclides from the biosphere. The performance assessment considers multiple waste types of varying thermal load and radionuclide inventory, the engineered barrier system comprising the waste packages, backfill, and emplacement drifts, and the natural barrier system formed by a bedded salt deposit and the overlying sedimentary sequence (including an aquifer). The model simulates disposal of nearly the entire inventory of DOE-managed, defense-related SNF (excluding Naval SNF) and HLW in a half-symmetry domain containing approximately 6 million grid cells. Grid refinement captures the detail of 25,200 individual waste packages in 180 disposal panels, associated access halls, and 4 shafts connecting the land surface to the repository. Equations describing coupled heat and fluid flow and reactive transport are solved numerically with PFLOTRAN, a massively parallel flow and transport code. Simulated processes include heat conduction and convection, waste package failure, waste form dissolution, radioactive decay and ingrowth, sorption, solubility limits, advection, dispersion, and diffusion. Simulations are run to 1 million years, and radionuclide concentrations are observed within an aquifer at a point approximately 4 kilometers downgradient of the repository. The software package DAKOTA is used to sample likely ranges of input parameters including waste form dissolution rates and properties of engineered and natural materials in order to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Sandia

  18. Technology development program for Idaho Chemical Processing Plant spent fuel and waste management

    SciTech Connect

    Ermold, L.F.; Knecht, D.A.; Hogg, G.W.; Olson, A.L.

    1993-08-01

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

  19. 76 FR 45268 - Reprocessing of Reusable Medical Devices

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-28

    ... HUMAN SERVICES Food and Drug Administration Reprocessing of Reusable Medical Devices AGENCY: Food and... Administration (FDA) is considering factors affecting the reprocessing of reusable medical devices, including reprocessing quality, device design as it relates to the reprocessing of reusable medical devices, reprocessing...

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

    SciTech Connect

    Fellinger, A.

    2009-12-08

    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

  1. New challenges in the safety analysis of DOE`s high-level waste tanks

    SciTech Connect

    Edwards, J.N.; Pasamehmetoglu, K.O.; White, J.R.

    1994-09-01

    Tank 241-SY-101, located at the Department of Energy Hanford Site, has periodically released up to 283 m{sup 3} (10,000 ft{sup 3}) of flammable gas. This release has been one of the highest priority DOE operational safety problems because of potential consequences if the gas were ignited during one of these releases. The gases include hydrogen and ammonia (fuels) and nitrous oxide (oxidizer). There have been many opinions regarding the controlling mechanisms for these releases, but demonstrating an adequate understanding of the problem, selecting a mitigation methodology, and preparing the safety analysis have presented numerous new challenges. The purpose of this report is to present an overview of the problem, the main issues, the method selected to mitigate this hazard, and the results of the mitigation program.

  2. Transformative monitoring approaches for reprocessing.

    SciTech Connect

    Cipiti, Benjamin B.

    2011-09-01

    The future of reprocessing in the United States is strongly driven by plant economics. With increasing safeguards, security, and safety requirements, future plant monitoring systems must be able to demonstrate more efficient operations while improving the current state of the art. The goal of this work was to design and examine the incorporation of advanced plant monitoring technologies into safeguards systems with attention to the burden on the operator. The technologies examined include micro-fluidic sampling for more rapid analytical measurements and spectroscopy-based techniques for on-line process monitoring. The Separations and Safeguards Performance Model was used to design the layout and test the effect of adding these technologies to reprocessing. The results here show that both technologies fill key gaps in existing materials accountability that provide detection of diversion events that may not be detected in a timely manner in existing plants. The plant architecture and results under diversion scenarios are described. As a tangent to this work, both the AMUSE and SEPHIS solvent extraction codes were examined for integration in the model to improve the reality of diversion scenarios. The AMUSE integration was found to be the most successful and provided useful results. The SEPHIS integration is still a work in progress and may provide an alternative option.

  3. RESULTS OF THE FY09 ENHANCED DOE HIGH LEVEL WASTE MELTER THROUGHPUT STUDIES AT SRNL

    SciTech Connect

    Johnson, F.; Edwards, T.

    2010-06-23

    High-level waste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). The objective of this task is to develop data, assess property models, and refine or develop the necessary models to support increased WL of HLW at SRS. It is a continuation of the studies initiated in FY07, but is under the specific guidance of a Task Change Request (TCR)/Work Authorization received from DOE headquarters (Project Number RV071301). Using the data generated in FY07, FY08 and historical data, two test matrices (60 glasses total) were developed at the Savannah River National Laboratory (SRNL) in order to generate data in broader compositional regions. These glasses were fabricated and characterized using chemical composition analysis, X-ray Diffraction (XRD), viscosity, liquidus temperature (TL) measurement and durability as defined by the Product Consistency Test (PCT). The results of this study are summarized below: (1) In general, the current durability model predicts the durabilities of higher waste loading glasses quite well. A few of the glasses exhibited poorer durability than predicted. (2) Some of the glasses exhibited anomalous behavior with respect to durability (normalized leachate for boron (NL [B])). The quenched samples of FY09EM21-02, -07 and -21 contained no nepheline or other wasteform affecting crystals, but have unacceptable NL [B] values (> 10 g/L). The ccc sample of FY09EM21-07 has a NL [B] value that is more than one half the value of the quenched sample. These glasses also have lower concentrations of Al{sub 2}O{sub 3} and SiO{sub 2}. (3) Five of the ccc samples (EM-13, -14, -15, -29 and

  4. Idaho Chemical Processing Plant Spent Fuel and Waste Management Technology Development Program Plan

    SciTech Connect

    1993-09-01

    The Department of Energy (DOE) has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage and reprocessing since 1953. Reprocessing of SNF has resulted in an existing inventory of 1.5 million gallons of radioactive sodium-bearing liquid waste and 3800 cubic meters (m{sup 3}) of calcine, in addition to the 768 metric tons (MT) of SNF and various other fuel materials in inventory. To date, the major activity of the ICPP has been the reprocessing of SNF to recover fissile uranium; however, recent changes in world events have diminished the demand to recover and recycle this material. As a result, DOE has discontinued reprocessing SNF for uranium recovery, making the need to properly manage and dispose of these and future materials a high priority. In accordance with the Nuclear Waste Policy Act (NWPA) of 1982, as amended, disposal of SNF and high-level waste (HLW) is planned for a geological repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP Spent Fuel and Waste Management Technology Development Program (SF&WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will properly stored and prepared for final disposal. Program elements in support of acceptable interim storage and waste minimization include: developing and implementing improved radioactive waste treatment technologies; identifying and implementing enhanced decontamination and decommissioning techniques; developing radioactive scrap metal (RSM) recycle capabilities; and developing and implementing improved technologies for the interim storage of SNF.

  5. Waste Generator Instructions: Key to Successful Implementation of the US DOE's 435.1 for Transuranic Waste Packaging Instructions (LA-UR-12-24155) - 13218

    SciTech Connect

    French, David M.; Hayes, Timothy A.; Pope, Howard L.; Enriquez, Alejandro E.; Carson, Peter H.

    2013-07-01

    In times of continuing fiscal constraints, a management and operation tool that is straightforward to implement, works as advertised, and virtually ensures compliant waste packaging should be carefully considered and employed wherever practicable. In the near future, the Department of Energy (DOE) will issue the first major update to DOE Order 435.1, Radioactive Waste Management. This update will contain a requirement for sites that do not have a Waste Isolation Pilot Plant (WIPP) waste certification program to use two newly developed technical standards: Contact-Handled Defense Transuranic Waste Packaging Instructions and Remote-Handled Defense Transuranic Waste Packaging Instructions. The technical standards are being developed from the DOE O 435.1 Notice, Contact-Handled and Remote-Handled Transuranic Waste Packaging, approved August 2011. The packaging instructions will provide detailed information and instruction for packaging almost every conceivable type of transuranic (TRU) waste for disposal at WIPP. While providing specificity, the packaging instructions leave to each site's own discretion the actual mechanics of how those Instructions will be functionally implemented at the floor level. While the Technical Standards are designed to provide precise information for compliant packaging, the density of the information in the packaging instructions necessitates a type of Rosetta Stone that translates the requirements into concise, clear, easy to use and operationally practical recipes that are waste stream and facility specific for use by both first line management and hands-on operations personnel. The Waste Generator Instructions provide the operator with step-by-step instructions that will integrate the sites' various operational requirements (e.g., health and safety limits, radiological limits or dose limits) and result in a WIPP certifiable waste and package that can be transported to and emplaced at WIPP. These little known but widely productive Waste

  6. GENERIC DEGRADATION SCENARIO AND CONFIGURATION ANALYSIS FOR DOE CODISPOSAL WASTE PACKAGE

    SciTech Connect

    S.F. A. Deng

    1999-11-10

    The purpose of this analysis is to develop a generic set of degradation scenarios and associated configurations for various Department of Energy (DOE) spent nuclear fuel (SNF) types when codisposed with the high-level waste (HLW) glass inside a waste package (WP). The degradation takes place inside the WP. These scenarios and configurations are developed as refinements of the standard degradation scenarios and potentially critical configuration classes given in Section 3.1 of the ''Disposal Criticality Analysis Methodology Topical Report'' (Ref. 1). Certain degradation scenarios and configurations will change when EDA II design is baselined. In accordance with AP-3.10Q, Revision 0, ICN 0, a work direction was developed, issued, and used in the preparation of this document (Ref. 13, p. 7).

  7. Using Department of Energy (DOE) Order 435.1 To Find a Cost Effective Waste Management Option. - 12241

    SciTech Connect

    LaBarge, Matt; Frost, Matt

    2012-07-01

    The Depleted Uranium Hexafluoride Project in Portsmouth Ohio was faced with an interesting dilemma. During hot functional testing in August 2010, an upset condition caused gaseous depleted uranium hexafluoride (UF{sub 6}) to come in contact with the hydrofluoric acid (HF) vapor stream. Although the resulting uranium contamination found in the condensed aqueous hydrofluoric acid was very low, it exceeded the Department of Energy (DOE) authorized release limit. After evaluating several commercial options for treatment and disposal using the guidelines found in DOE Order 435.1, Waste Control Specialists LLC was selected for the treatment of the waste, with EnergySolutions' Clive facility selected for disposal of the treated residues. The waste was safely transported from Piketon, Ohio to Andrews, Texas, where it was treated to meet the land Disposal Restrictions (LDR), and was disposed in EnergySolutions operational mixed waste cell. The entire effort was interesting for several reasons. The waste was generated during the last year of the first Depleted Uranium Hexafluoride contractor. The waste became additional scope for the new contractor, adding time delays and introducing new personnel into the project. The effort was also unique because it demonstrated the process mandated by DOE Order 435.1 to evaluate all options, including commercial options, could reveal solutions to waste management problems that are currently available and more cost effective, but not well know within the DOE complex. (authors)

  8. X-ray reprocessing in binaries

    NASA Astrophysics Data System (ADS)

    Paul, Biswajit

    2016-07-01

    We will discuss several aspects of X-ray reprocessing into X-rays or longer wavelength radiation in different kinds of binary systems. In high mass X-ray binaries, reprocessing of hard X-rays into emission lines or lower temperature black body emission is a useful tool to investigate the reprocessing media like the stellar wind, clumpy structures in the wind, accretion disk or accretion stream. In low mass X-ray binaries, reprocessing from the surface of the companion star, the accretion disk, warps and other structures in the accretion disk produce signatures in longer wavelength radiation. X-ray sources with temporal structures like the X-ray pulsars and thermonuclear burst sources are key in such studies. We will discuss results from several new investigations of X-ray reprocessing phenomena in X-ray binaries.

  9. Improvements to the DOE low-level waste regulatory structure and process under recommendation 94-2 - progress to date

    SciTech Connect

    Regnier, E.

    1995-12-31

    Among the concerns expressed by the Defense Nuclear Facility Safety Board (DNFSB) in its Recommendation 94-2 was the lack of a clearly defined and effective internal Department of Energy (DOE) regulatory oversight and enforcement process for ensuring that low-level radioactive waste management health, safety, and environmental requirements are met. Therefore, part of the response to the DNFSB concern is a task to clarify and strengthen the low-level waste management regulatory structure. This task is being conducted in two steps. First, consistent with the requirements of the current DOE waste management order and within the framework of the current organizational structure, interim clarification of a review process and the associated organizational responsibilities has been issued. Second, in coordination with the revision of the waste management order and consistent with the organizational responsibilities resulting from the strategic alignment of DOE, a rigorous, more independent regulatory oversight structure will be developed.

  10. A DOE contractor`s perspective of environmental monitoring requirements at a low-level waste facility

    SciTech Connect

    Ferns, T.W.

    1989-11-01

    Environmental monitoring at a low-level waste disposal facility (LLWDF) should, (1) demonstrate compliance with environmental laws; (2) detect any spatial or temporal environmental changes; and (3) provide information on the potential or actual exposure of humans and/or the environment to disposed waste and/or waste by-products. Under the DOE Order system the LLWDF site manager has more freedom of implementation for a monitoring program than either the semi-prescriptive NRC, or the prescriptive EPA hazardous waste programs. This paper will attempt to compare and contrast environmental monitoring under the different systems (DOE, NRC, and EPA), and determine if the DOE might benefit from a more prescriptive system.

  11. Comparison of alternative treatment systems for DOE mixed low-level waste

    SciTech Connect

    Schwinkendorf, W.E.

    1997-03-01

    From 1993 to 1996, the Department of Energy, Environmental Management, Office of Science and Technology (OST), has sponsored a series of systems analyses to guide its future research and development (R&D) programs for the treatment of mixed low-level waste (MLLW) stored in the DOE complex. The two original studies were of 20 mature and innovative thermal systems. As a result of a technical review of these thermal system studies, a similar study of five innovative nonthermal systems was conducted in which unit operations are limited to temperatures less than 350{degrees}C to minimize volatilization of heavy metals and radionuclides, and de novo production of dioxins and furans in the offgas. Public involvement in the INTS study was established through a working group of 20 tribal and stakeholder representatives to provide input to the INTS studies and identify principles against which the systems should be designed and evaluated. Pre-conceptual designs were developed for all systems to treat the same waste input (2927 lbs/hr) in a single centralized facility operating 4032 hours per year for 20 years. This inventory consisted of a wide range of combustible and non-combustible materials such as paper, plastics, metals, concrete, soils, sludges, liquids, etc., contaminated with trace quantities of radioactive materials and RCRA regulated wastes. From this inventory, an average waste profile was developed for simulated treatment using ASPEN PLUS{copyright} for mass balance calculations. Seven representative thermal systems were selected for comparison with the five nonthermal systems. This report presents the comparisons against the TSWG principles, of total life cycle cost (TLCC), and of other system performance indicators such as energy requirements, reagent requirements, land use, final waste volume, aqueous and gaseous effluents, etc.

  12. Report of the DOD-DOE Workshop on Converting Waste to Energy Using Fuel Cells

    DTIC Science & Technology

    2011-10-01

    Summary on Converting Waste to Energy Using Fuel Cells million metric tons of carbon dioxide equivalent.7 DOD’s high energy dependence and reliance on...as te Animal W as te L andfill Methane (Vol%) 80–100 ~50–60 ~50–70 45–60 40–55 Carbon Dioxide (Vol%) < 3 30–40 25–45 35–50 35–50 Nitrogen (Vol...transitioning to a low- carbon economy. A key focus area of the MOU is DOD-DOE collaboration on a broad range of innovative, technology-driven solutions

  13. Spent nuclear fuel reprocessing modeling

    SciTech Connect

    Tretyakova, S.; Shmidt, O.; Podymova, T.; Shadrin, A.; Tkachenko, V.; Makeyeva, I.; Tkachenko, V.; Verbitskaya, O.; Schultz, O.; Peshkichev, I.

    2013-07-01

    The long-term wide development of nuclear power requires new approaches towards the realization of nuclear fuel cycle, namely, closed nuclear fuel cycle (CNFC) with respect to fission materials. Plant nuclear fuel cycle (PNFC), which is in fact the reprocessing of spent nuclear fuel unloaded from the reactor and the production of new nuclear fuel (NF) at the same place together with reactor plant, can be one variant of CNFC. Developing and projecting of PNFC is a complicated high-technology innovative process that requires modern information support. One of the components of this information support is developed by the authors. This component is the programme conducting calculations for various variants of process flow sheets for reprocessing SNF and production of NF. Central in this programme is the blocks library, where the blocks contain mathematical description of separate processes and operations. The calculating programme itself has such a structure that one can configure the complex of blocks and correlations between blocks, appropriate for any given flow sheet. For the ready sequence of operations balance calculations are made of all flows, i.e. expenses, element and substance makeup, heat emission and radiation rate are determined. The programme is open and the block library can be updated. This means that more complicated and detailed models of technological processes will be added to the library basing on the results of testing processes using real equipment, in test operating mode. The development of the model for the realization of technical-economic analysis of various variants of technologic PNFC schemes and the organization of 'operator's advisor' is expected. (authors)

  14. EVALUATION OF THOR MINERALIZED WASTE FORMS FOR THE DOE ADVANCED REMEDIATION TECHNOLOGIES PHASE 2 PROJECT

    SciTech Connect

    Crawford, C.; Jantzen, C.

    2012-02-02

    The U.S. Department of Energy's (DOE) Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP's LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). In addition, the WTP LAW Vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as Cs-137, I-129, Tc-99, Cl, F, and SO{sub 4} that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap. The current waste disposal path for the WTP-SW is to recycle it to the supplemental LAW treatment to avoid a large steady state accumulation in the pretreatment-vitrification loop. Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750 C) continuous method by which LAW and/or WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates

  15. Radioactive waste management

    SciTech Connect

    Flax, S.J.

    1981-01-01

    This article examines the technical and legal considerations of nuclear waste management. The first three sections describe the technical aspects of spent-fuel-rod production, reprocessing, and temporary storage. The next two sections discuss permanent disposal of high-level wastes and spent-fuel rods. Finally, legislative and judicial responses to the nuclear-waste crisis.

  16. Low-temperature setting phosphate ceramics for stabilization of DOE problem low level mixed-waste: I. Material and waste form development

    SciTech Connect

    Singh, D.; Wagh, A.; Knox, L.; Mayberry, J.

    1994-03-01

    Chemically bonded phosphate ceramics are proposed as candidates for solidification and stabilization of some of the {open_quotes}problem{close_quotes} DOE low-level mixed wastes at low-temperatures. Development of these materials is crucial for stabilization of waste streams which have volatile species and any use of high-temperature technology leads to generation of off-gas secondary waste streams. Several phosphates of Mg, Al, and Zr have been investigated as candidate materials. Monoliths of these phosphates were synthesized using chemical routes at room or slightly elevated temperatures. Detailed physical and chemical characterizations have been conducted on some of these phosphates to establish their durability. Magnesium ammonium phosphate has shown to possess excellent mechanical and as well chemical properties. These phosphates were also used to stabilize a surrogate ash waste with a loading ranging from 25-35 wt.%. Characterization of the final waste forms show that waste immobilization is due to both chemical stabilization and physical encapsulation of the surrogate waste which is desirable for waste immobilization.

  17. Quantities and characteristics of the contact-handled low-level mixed waste streams for the DOE complex

    SciTech Connect

    Huebner, T.L.; Wilson, J.M.; Ruhter, A.H.; Bonney, S.J.

    1994-08-01

    This report supports the Integrated Thermal Treatment System (ITTS) Study initiated by the Department of Energy (DOE) Office of Technology Development (EM-50), which is a system engineering assessment of a variety of mixed waste treatment process. The DOE generates and stores large quantities of mixed wastes that are contaminated with both chemically hazardous and radioactive species. The treatment of these mixed wastes requires meeting the standards established by the Environmental Protection Agency for the specific hazardous contaminants regulated under the Resource Conservation and Recovery Act while also providing adequate control of the radionuclides. The thrust of the study is to develop preconceptual designs and life-cycle cost estimates for integrated thermal treatment systems ranging from conventional incinerators, such as rotary kiln and controlled air systems, to more innovative but not yet established technologies, such as molten salt and molten metal waste destruction systems. Prior to this engineering activity, the physical and chemical characteristics of the DOE low-level mixed waste streams to be treated must be defined or estimated. This report describes efforts to estimate the DOE waste stream characteristics.

  18. DOE underground storage tank waste remediation chemical processing hazards. Part I: Technology dictionary

    SciTech Connect

    DeMuth, S.F.

    1996-10-01

    This document has been prepared to aid in the development of Regulating guidelines for the Privatization of Hanford underground storage tank waste remediation. The document has been prepared it two parts to facilitate their preparation. Part II is the primary focus of this effort in that it describes the technical basis for established and potential chemical processing hazards associated with Underground Storage Tank (UST) nuclear waste remediation across the DOE complex. The established hazards involve those at Sites for which Safety Analysis Reviews (SARs) have already been prepared. Potential hazards are those involving technologies currently being developed for future applications. Part I of this document outlines the scope of Part II by briefly describing the established and potential technologies. In addition to providing the scope, Part I can be used as a technical introduction and bibliography for Regulatory personnel new to the UST waste remediation, and in particular Privatization effort. Part II of this document is not intended to provide examples of a SAR Hazards Analysis, but rather provide an intelligence gathering source for Regulatory personnel who must eventually evaluate the Privatization SAR Hazards Analysis.

  19. DOE final report, phase one startup, Waste Receiving and Processing Facility (WRAP)

    SciTech Connect

    Jasen, W.G.

    1998-01-07

    This document is to validate that the WRAP facility is physically ready to start up phase 1, and that the managers and operators are prepared to safely manage and operate the facility when all pre-start findings have been satisfactorily corrected. The DOE Readiness Assessment (RA) team spent a week on-site at Waste Receiving and Processing Module 1 (WRAP-1) to validate the readiness for phase 1 start up of facility. The Contractor and DOE staff were exceptionally cooperative and contributed significantly to the overall success of the RA. The procedures and Conduct of Operations areas had significant discrepancies, many of which should have been found by the contractor review team. In addition the findings of the contractor review team should have led the WRAP-1 management team to correcting the root causes of the findings prior to the DOE RA team review. The findings and observations include many issues that the team believes should have been found by the contractor review and corrective actions taken. A significantly improved Operational Readiness Review (ORR) process and corrective actions of root causes must be fully implemented by the contractor prior to the performance of the contractor ORR for phase 2 operations. The pre-start findings as a result of this independent DOE Readiness Assessment are presented.

  20. The Fourth (A)ATSR Data Reprocessing

    NASA Astrophysics Data System (ADS)

    Goryl, Philippe; Cocevar, Pauline; Done, Fay; Aatsr Quality Working Group

    2016-08-01

    This paper aims to inform users of the upcoming Fourth Reprocessing of ATSR-1, ATSR-2 and AATSR data. The main objective of the Fourth Reprocessing is to generate (A)ATSR Level 1B data products in a similar format to SLSTR products from Sentinel-3. In this way, users can easily access the 20-year dataset from the ERS and ENVISAT (A)ATSR missions and carry the analysis forward into the Sentinel era. In addition to the product format change, the dataset will build on the improvements implemented in the Third Reprocessing, and will contain further improvements and enhancements, as described below.

  1. Radioactive Iodine and Krypton Control for Nuclear Fuel Reprocessing Facilities

    SciTech Connect

    Soelberg, Nick R.; Garn, Troy G.; Greenhalgh, Mitchell R.; Law, Jack D.; Jubin, Robert; Strachan, Denis M.; Thallapally, Praveen K.

    2013-01-01

    The removal of volatile radionuclides generated during used nuclear fuel reprocessing in the US is almost certain to be necessary for the licensing of a reprocessing facility in the US. Various control technologies have been developed, tested, or used over the past 50 years for control of volatile radionuclide emissions from used fuel reprocessing plants. The US DOE has sponsored, since 2009, an Off-gas Sigma Team to perform research and development focused on the most pressing volatile radionuclide control and immobilization problems. In this paper, we focus on the control requirements and methodologies for85Kr and129I. Numerous candidate technologies have been studied and developed at laboratory and pilot-plant scales in an effort to meet the need for high iodine control efficiency and to advance alternatives to cryogenic separations for krypton control. Several of these show promising results. Iodine decontamination factors as high as 105, iodine loading capacities, and other adsorption parameters including adsorption rates have been demonstrated under some conditions for both silver zeolite (AgZ) and Ag-functionalized aerogel. Sorbents, including an engineered form of AgZ and selected metal organic framework materials (MOFs), have been successfully demonstrated to capture Kr and Xe without the need for separations at cryogenic temperatures.

  2. Evaluation of the Quality of Reprocessing of Gastrointestinal Endoscopes.

    PubMed

    Saviuc, Philippe; Picot-Guéraud, Romain; Shum Cheong Sing, Jacqueline; Batailler, Pierre; Pelloux, Isabelle; Brenier-Pinchart, Marie-Pierre; Dobremez, Valérie; Mallaret, Marie-Reine

    2015-09-01

    To evaluate the quality of gastrointestinal endoscope reprocessing and discuss the advantages of microbiological surveillance testing of these endoscopes. Retrospective analysis of the results of endoscope sampling performed from October 1, 2006, through December 31, 2014, in a gastrointestinal endoscopy unit of a teaching hospital equipped with 89 endoscopes and 3 automated endoscope reprocessors, with an endoscopy quality assurance program in place. The compliance rate was defined as the proportion of the results classified at target or alert levels according to the French guidelines. A multivariate analysis (logistic regression) was used to identify the parameters influencing compliance. A total of 846 samples were taken. The overall compliance rate was 86% and differed significantly depending on the sampling context (scheduled or not scheduled), the type of endoscope, and the season. No other parameter was associated with compliance. A total of 118 samples carried indicator microorganisms such as Pseudomonas aeruginosa, Stenotrophomonas maltophilia, Enterobacteriaceae, and Candida sp. The systematic use of an automated endoscope reprocessor does not provide totally satisfactory compliance. Microbiological surveillance is indispensable to monitor reprocessing, reinforce good practices (endoscopes, reprocessing units), and detect endoscopes requiring early technical maintenance.

  3. A proliferation of nuclear waste for the Southeast.

    PubMed

    Alvarez, Robert; Smith, Stephen

    2007-12-01

    The U.S. Department of Energy's (DOE) Global Nuclear Energy Partnership (GNEP) is being promoted as a program to bring about the expansion of worldwide nuclear energy. Here in the U.S. much of this proposed nuclear power expansion is slated to happen in the Southeast, including here in South Carolina. Under the GNEP plan, the United States and its nuclear partners would sell nuclear power plants to developing nations that agree not to pursue technologies that would aid nuclear weapons production, notably reprocessing and uranium enrichment. As part of the deal, the United States would take highly radioactive spent ("used") fuel rods to a reprocessing center in this country. Upon analysis of the proposal, it is clear that DOE lacks a credible plan for the safe management and disposal of radioactive wastes stemming from the GNEP program and that the high costs and possible public health and environmental impacts from the program pose significant risks, especially to this region. Given past failures to address waste problems before they were created, DOE's rush to invest major public funds for deployment of reprocessing should be suspended.

  4. Review of Analytes of Concern and Sample Methods for Closure of DOE High Level Waste Storage Tanks

    SciTech Connect

    Thomas, T.R.

    2002-05-06

    Sampling residual waste after tank cleaning and analysis for analytes of concern to support closure and cleaning targets of large underground tanks used for storage of legacy high level radioactive waste (HLW) at Department of Energy (DOE) sites has been underway since about 1995. The DOE Tanks Focus Area (TFA) has been working with DOE tank sites to develop new sampling plans, and sampling methods for assessment of residual waste inventories. This paper discusses regulatory analytes of concern, sampling plans, and sampling methods that support closure and cleaning target activities for large storage tanks at the Hanford Site, the Savannah River Site (SRS), the Idaho National Engineering and Environmental Laboratory (INEEL), and the West Valley Demonstration Project (WVDP).

  5. Review of Analytes of Concern and Sample Methods for Closure of DOE High Level Waste Storage Tanks

    SciTech Connect

    Thomas, Thomas Russell

    2002-08-01

    Sampling residual waste after tank cleaning and analysis for analytes of concern to support closure and cleaning targets of large underground tanks used for storage of legacy high level radioactive waste (HLW) at Department of Energy (DOE) sites has been underway since about 1995. The DOE Tanks Focus Area (TFA) has been working with DOE tank sites to develop new sampling plans, and sampling methods for assessment of residual waste inventories. This paper discusses regulatory analytes of concern, sampling plans, and sampling methods that support closure and cleaning target activities for large storage tanks at the Hanford Site, the Savannah River Site (SRS), the Idaho National Engineering and Environmental Laboratory (INEEL), and the West Valley Demonstration Project (WVDP).

  6. Development of Advanced Electrochemical Emission Spectroscopy for Monitoring Corrosion in Simulated DOE Liquid Waste

    SciTech Connect

    Digby Macdonald; Brian Marx; Balaji Soundararajan; Morgan Smith

    2005-07-28

    The different tasks that have been carried out under the current program are as follows: (1) Theoretical and experimental assessment of general corrosion of iron/steel in borate buffer solutions by using electrochemical impedance spectroscopy (EIS), ellipsometry and XPS techniques; (2) Development of a damage function analysis (DFA), which would help in predicting the accumulation of damage due to pitting corrosion in an environment prototypical of DOE liquid waste systems; (3) Experimental measurement of crack growth rate, acoustic emission signals, and coupling currents for fracture in carbon and low alloy steels as functions of mechanical (stress intensity), chemical (conductivity), electrochemical (corrosion potential, ECP), and microstructural (grain size, precipitate size, etc) variables in a systematic manner, with particular attention being focused on the structure of the noise in the current and its correlation with the acoustic emissions; (4) Development of fracture mechanisms for carbon and low alloy steels that are consistent with the crack growth rate, coupling current data and acoustic emissions; (5) Inserting advanced crack growth rate models for SCC into existing deterministic codes for predicting the evolution of corrosion damage in DOE liquid waste storage tanks; (6) Computer simulation of the anodic and cathodic activity on the surface of the steel samples in order to exactly predict the corrosion mechanisms; (7) Wavelet analysis of EC noise data from steel samples undergoing corrosion in an environment similar to that of the high level waste storage containers, to extract data pertaining to general, pitting and stress corrosion processes, from the overall data. The work has yielded a number of important findings, including an unequivocal demonstration of the role of chloride ion in passivity breakdown on nickel in terms of cation vacancy generation within the passive film, the first detection and characterization of individual micro fracture

  7. LLNL radioactive waste management plan as per DOE Order 5820. 2

    SciTech Connect

    Not Available

    1984-12-10

    The following aspects of LLNL's radioactive waste management plan are discussed: program administration; description of waste generating processes; radioactive waste collection, treatment, and disposal; sanitary waste management; site 300 operations; schedules and major milestones for waste management activities; and environmental monitoring programs (sampling and analysis).

  8. Proceedings of the 23rd DOE/NRC nuclear air cleaning conference

    SciTech Connect

    First, M.W.

    1995-02-01

    The report contains the papers presented at the 23rd DOE/NRC Nuclear Air Cleaning Conference and the associated discussions. Major topics are: (1) nuclear air cleaning codes, (2) nuclear waste, (3) filters and filtration, (4) effluent stack monitoring, (5) gas processing, (6) adsorption, (7) air treatment systems, (8) source terms and accident analysis, and (9) fuel reprocessing. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

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

    SciTech Connect

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

    1989-10-01

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

  10. Optimizing High Level Waste Disposal

    SciTech Connect

    Dirk Gombert

    2005-09-01

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

  11. Sodium Recycle Economics for Waste Treatment Plant Operations

    SciTech Connect

    Sevigny, Gary J.; Poloski, Adam P.; Fountain, Matthew S.

    2008-03-01

    Sodium recycle at the Hanford Waste Treatment Plant (WTP) would reduce the number of glass canisters produced, and has the potential to save the U.S. Department of Energy (DOE) tens of millions of dollars. The sodium, added in the form of sodium hydroxide, was originally added to minimize corrosion of carbon-steel storage tanks from acidic reprocessing wastes. In the baseline Hanford treatment process, sodium hydroxide is required to leach gibbsite and boehmite from the high level waste (HLW) sludge. In turn, this reduces the amount of HLW glass produced. Currently, a significant amount of additional sodium hydroxide will be added to the process to maintain aluminate solubility at ambient temperatures during ion exchange of cesium. The vitrification of radioactive waste is limited by sodium content, and this additional sodium mass will increase low-activity waste-glass mass.

  12. Waste Estimates for a Future Recycling Plant in the US Based Upon AREVA Operating Experience - 13206

    SciTech Connect

    Foare, Genevieve; Meze, Florian; Bader, Sven; McGee, Don; Murray, Paul; Prud'homme, Pascal

    2013-07-01

    Estimates of process and secondary wastes produced by a recycling plant built in the U.S., which is composed of a used nuclear fuel (UNF) reprocessing facility and a mixed oxide (MOX) fuel fabrication facility, are performed as part of a U.S. Department of Energy (DOE) sponsored study [1]. In this study, a set of common inputs, assumptions, and constraints were identified to allow for comparison of these wastes between different industrial teams. AREVA produced a model of a reprocessing facility, an associated fuel fabrication facility, and waste treatment facilities to develop the results for this study. These facilities were divided into a number of discrete functional areas for which inlet and outlet flow streams were clearly identified to allow for an accurate determination of the radionuclide balance throughout the facility and the waste streams. AREVA relied primarily on its decades of experience and feedback from its La Hague (reprocessing) and MELOX (MOX fuel fabrication) commercial operating facilities in France to support this assessment. However, to perform these estimates for a U.S. facility with different regulatory requirements and to take advantage of some technological advancements, such as in the potential treatment of off-gases, some deviations from this experience were necessary. A summary of AREVA's approach and results for the recycling of 800 metric tonnes of initial heavy metal (MTIHM) of LWR UNF per year into MOX fuel under the assumptions and constraints identified for this DOE study are presented. (authors)

  13. Development of Advanced Electrochemical Emission Spectroscopy for Monitoring Corrosion in Simulated DOE Liquid Waste

    SciTech Connect

    Macdonald, Digby; Liu, Jun; Liu, Sue; Al-Rifaie, Mohammed; Sikora; Elzbieta

    2000-06-01

    The principal goals of this project are to develop advanced electrochemical emission spectroscopic (EES) methods for monitoring the corrosion of carbon steel in simulated DOE liquid waste and to develop a better understanding of the mechanisms of the corrosion of metals (e.g. iron, nickel, and chromium) and alloys (carbon steel, low alloy steels, stainless steels) in thes e environments. During the first two years of this project, significant advances have been made in developing a better understanding of the corrosion of iron in aqueous solutions as a function of pH, on developing a better understanding of the growth of passive films on metal surfaces, and on developing EES techniques for corrosion monitoring. This report summarizes work on beginning the third year of the 3-year project.

  14. Consolidated Fuel Reprocessing Program. National Program Plan, FY 1985

    SciTech Connect

    Not Available

    1985-03-01

    This FY 1985 National Program Plan for the Consolidated Fuel Reprocessing Program (CFRP) provides specific guidance from the Department of Energy (DOE) for FY 1985 CFRP activities and a baseline for activities in future years. This is the third issue of the Plan, which is updated annually and summarizes program objectives, plans, and schedules, budget allocations, contractor involvements, and interfaces with other research programs. The National Program Plan is a controlling document for the Technical Program Plan, which is prepared annually by the CFRP at Oak Ridge National Laboratory (ORNL) and is one of a hierarchical group of planning documents. The CFRP is a part of the DOE's program of research and development (R and D) on nuclear fission systems and is directed by the LMFBR Fuel Cycle Projects Division, Office of Spent Fuel Management and Reprocessing Systems at DOE-Headquarters through the Oak Ridge Operations Office (ORO). The strategy of the program is to maintain the capability to commit to a breeder option through a strong fuel cycle R and D program and international technical exchanges.

  15. Comparison of low-level waste disposal programs of DOE and selected international countries

    SciTech Connect

    Meagher, B.G.; Cole, L.T.

    1996-06-01

    The purpose of this report is to examine and compare the approaches and practices of selected countries for disposal of low-level radioactive waste (LLW) with those of the US Department of Energy (DOE). The report addresses the programs for disposing of wastes into engineered LLW disposal facilities and is not intended to address in-situ options and practices associated with environmental restoration activities or the management of mill tailings and mixed LLW. The countries chosen for comparison are France, Sweden, Canada, and the United Kingdom. The countries were selected as typical examples of the LLW programs which have evolved under differing technical constraints, regulatory requirements, and political/social systems. France was the first country to demonstrate use of engineered structure-type disposal facilities. The UK has been actively disposing of LLW since 1959. Sweden has been disposing of LLW since 1983 in an intermediate-depth disposal facility rather than a near-surface disposal facility. To date, Canada has been storing its LLW but will soon begin operation of Canada`s first demonstration LLW disposal facility.

  16. Treatment of DOE and commercial mixed waste by the private sector

    SciTech Connect

    Garrison, T.W.; Apel, M.L.; Owens, C.M.

    1993-03-01

    This paper presents a conceptual approach for private sector treatment of mixed low-level radioactive waste generated by the US Department of Energy and commercial industries. This approach focuses on MLLW treatment technologies and capacities available through the private sector in the near term. Wastestream characterization data for 108 MLLW streams at the Idaho National Engineering Laboratory (INEL) were collected and combined with similar data for MLLWs generated through commercial practices. These data were then provided to private treatment facilities and vendors to determine if, and to what extent, they could successfully treat these wastes. Data obtained from this project have provided an initial assessment of private sector capability and capacity to treat a variety of MLLW streams. This information will help formulate plans for future treatment of these and similar wastestreams at DOE facilities. This paper presents details of the MLLW data-gathering efforts used in this research, private sector assessment methods employed, and results of this assessment. Advantages of private sector treatment, as well as barriers to its present use, are also addressed.

  17. Direct Reduction of Waste through Refining of DOE Metal Assets - 13632

    SciTech Connect

    Hargett, Michael C.; Terekhov, Dimitri; Khozan, Kamran M.

    2013-07-01

    CVMR{sup R} presents a technology for refining nickel from the enrichment barrier materials of the DOE that is proven through 100 years of use by the metals industry. CVMR{sup R} applies modern controls, instrumentation for process and monitoring of the system, and innovative production methods to produce a wide spectrum of products that generate new technology applications and improvements to our society and economy. CVMR{sup R} will receive barrier materials as a secure operation and size reduce the metal to a shred that is fed to a carbonylation reactor where nickel is reacted with carbon monoxide and generate nickel carbonyl. The carbonyl will be filtered and decomposed with heat to form a variety of products that include high value nano powders, coated substrates, net shapes and pure nickel. The residue from the reactor will retain radionuclides from enrichment activities. The carbon monoxide will only react and extract nickel under the operating conditions to leave volumetric contamination in the unreacted residue. A demonstration plant was designed and built by CVMR{sup R} and operated by BWXT, to demonstrate the systems capabilities to DOE in 2006. A pilot plant operation precedes the detailed design of the nickel refinery and provides essential data for design, safe work practices, waste characterizations and system kinetics and confirms the project feasibility. CVMR{sup R} produces nickel products that are cleaner than the nickel in U.S. commerce and used by industry today. The CVMR{sup R} process and systems for nickel refining is well suited for DOE materials and will provide value through environmental stewardship, recovery of high value assets, and support of the DOE environmental remediation programs as the refined nickel generates additional long term benefits to local communities. (authors)

  18. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    SciTech Connect

    Mcwilliams, A. J.

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  19. FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION

    SciTech Connect

    Carter, J.

    2011-01-03

    The United States (U.S.) currently utilizes a once-through fuel cycle where used nuclear fuel (UNF) is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. Within the Department of Energy's (DOE) Office of Nuclear Energy (DOE-NE), the Fuel Cycle Research and Development Program (FCR&D) develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development of advanced fuel cycles, including modified open and closed cycles. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Yet, the routine disposal of used nuclear fuel and radioactive waste remains problematic. Advanced fuel cycles will generate different quantities and forms of waste than the current LWR fleet. This study analyzes the quantities and characteristics of potential waste forms including differing waste matrices, as a function of a variety of potential fuel cycle alternatives including: (1) Commercial UNF generated by uranium fuel light water reactors (LWR). Four once through fuel cycles analyzed in this study differ by varying the assumed expansion/contraction of nuclear power in the U.S. (2) Four alternative LWR used fuel recycling processes analyzed differ in the reprocessing method (aqueous vs. electro-chemical), complexity (Pu only or full transuranic (TRU) recovery) and waste forms generated. (3) Used Mixed Oxide (MOX) fuel derived from the recovered Pu utilizing a single reactor pass. (4) Potential waste forms generated by the reprocessing of fuels derived from recovered TRU utilizing multiple reactor passes.

  20. FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION

    SciTech Connect

    Jones, R.; Carter, J.

    2010-10-13

    The United States (U.S.) currently utilizes a once-through fuel cycle where used nuclear fuel (UNF) is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. Within the Department of Energy's (DOE) Office of Nuclear Energy (DOE-NE), the Fuel Cycle Research and Development Program (FCR&D) develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development of advanced fuel cycles, including modified open and closed cycles. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Yet, the routine disposal of used nuclear fuel and radioactive waste remains problematic. Advanced fuel cycles will generate different quantities and forms of waste than the current LWR fleet. This study analyzes the quantities and characteristics of potential waste forms including differing waste matrices, as a function of a variety of potential fuel cycle alternatives including: (1) Commercial UNF generated by uranium fuel light water reactors (LWR). Four once through fuel cycles analyzed in this study differ by varying the assumed expansion/contraction of nuclear power in the U.S; (2) Four alternative LWR used fuel recycling processes analyzed differ in the reprocessing method (aqueous vs. electro-chemical), complexity (Pu only or full transuranic (TRU) recovery) and waste forms generated; (3) Used Mixed Oxide (MOX) fuel derived from the recovered Pu utilizing a single reactor pass; and (4) Potential waste forms generated by the reprocessing of fuels derived from recovered TRU utilizing multiple reactor passes.

  1. 1994 Annual report on waste generation and waste minimization progress as required by DOE Order 5400.1, Hanford Site

    SciTech Connect

    1995-09-01

    Many Waste Minimization/Pollution Prevention successes at the Hanford Site occur every day without formal recognition. A few of the successful projects are: T-Plant helps facilities reuse equipment by offering decontamination services for items such as gas cylinders, trucks, and railcars, thus saving disposal and equipment replacement costs. Custodial Services reviewed its use of 168 hazardous cleaning products, and, through a variety of measures, replaced them with 38 safer substitutes, one for each task. Scrap steel contaminated with low level radioactivity from the interim stabilization of 107-K and 107-C was decontaminated and sold to a vendor for recycling. Site-wide programs include the following: the Pollution Prevention Opportunity Assessment (P2OA) program at the Hanford site was launched during 1994, including a training class, a guidance document, technical assistance, and goals; control over hazardous materials purchased was achieved by reviewing all purchase requisitions of a chemical nature; the Office Supply Reuse Program was established to redeploy unused or unwanted office supply items. In 1994, pollution prevention activities reduced approximately 274,000 kilograms of hazardous waste, 2,100 cubic meters of radioactive and mixed waste, 14,500,000 kilograms of sanitary waste, and 215,000 cubic meters off liquid waste and waste water. Pollution Prevention activities also saved almost $4.2 million in disposal, product, and labor costs. Overall waste generation increased in 1994 due to increased work and activity typical for a site with an environmental restoration mission. However, without any Waste Minimization/Pollution Prevention activities, solid radioactive waste generation at Hanford would have been 25% higher, solid hazardous waste generation would have been 30% higher, and solid sanitary waste generation would have been 60% higher.

  2. Environmental assessment for DOE permission for off-loading activities to support the movement of commercial low level nuclear waste across the Savannah River Site

    SciTech Connect

    1995-02-01

    This environmental assessment investigates the potential environmental and safety effects which could result from the land transport of low level radioactive wastes across the Savannah River Plant. Chem-Nuclear Systems operates a low level radioactive waste burial facility adjacent to the Savannah River Plant and is seeking permission from the DOE to transport the waste across Savannah River Plant.

  3. Development of Advanced Electrochemical Emission Spectroscopy for Monitoring Corrosion in Simulated DOE Liquid Waste

    SciTech Connect

    Digby D. Macdonald; Brian M. Marx; Sejin Ahn; Julio de Ruiz; Balaji Soundararaja; Morgan Smith; and Wendy Coulson

    2008-01-15

    Various forms of general and localized corrosion represent principal threats to the integrity of DOE liquid waste storage tanks. These tanks, which are of a single wall or double wall design, depending upon their age, are fabricated from welded carbon steel and contain a complex waste-form comprised of NaOH and NaNO{sub 3}, along with trace amounts of phosphate, sulfate, carbonate, and chloride. Because waste leakage can have a profound environmental impact, considerable interest exists in predicting the accumulation of corrosion damage, so as to more effectively schedule maintenance and repair. The different tasks that are being carried out under the current program are as follows: (1) Theoretical and experimental assessment of general corrosion of iron/steel in borate buffer solutions by using electrochemical impedance spectroscopy (EIS), ellipsometry and XPS techniques; (2) Development of a damage function analysis (DFA) which would help in predicting the accumulation of damage due to pitting corrosion in an environment prototypical of DOE liquid waste systems; (3) Experimental measurement of crack growth rate, acoustic emission signals and coupling currents for fracture in carbon and low alloy steels as functions of mechanical (stress intensity), chemical (conductivity), electrochemical (corrosion potential, ECP), and microstructural (grain size, precipitate size, etc) variables in a systematic manner, with particular attention being focused on the structure of the noise in the current and its correlation with the acoustic emissions; (4) Development of fracture mechanisms for carbon and low alloy steels that are consistent with the crack growth rate, coupling current data and acoustic emissions; (5) Inserting advanced crack growth rate models for SCC into existing deterministic codes for predicting the evolution of corrosion damage in DOE liquid waste storage tanks; (6) Computer simulation of the anodic and cathodic activity on the surface of the steel samples

  4. MicroRaman measurements for nuclear fuel reprocessing applications

    DOE PAGES

    Casella, Amanda; Lines, Amanda; Nelson, Gilbert; ...

    2016-12-01

    Treatment and reuse of used nuclear fuel is a key component in closing the nuclear fuel cycle. Solvent extraction reprocessing methods that have been developed contain various steps tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper willmore » focus on the development and application of a new MicroRaman probe for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid in solutions relevant to used nuclear fuel reprocessing. Previous research has successfully demonstrated the applicability on the macroscopic scale, using sample probes requiring larger solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 μL or less, which is a scale reduction of over 1,000 fold in sample size. Finally, this paper will summarize our current work in this area including: comparisons between the macroscopic and microscopic probes for detection limits, optimized channel focusing, and application in a flow cell with varying levels of HNO3, and UO2(NO3)2.« less

  5. MicroRaman measurements for nuclear fuel reprocessing applications

    SciTech Connect

    Casella, Amanda; Lines, Amanda; Nelson, Gilbert; Bello, Job; Bryan, Samuel

    2016-12-01

    Treatment and reuse of used nuclear fuel is a key component in closing the nuclear fuel cycle. Solvent extraction reprocessing methods that have been developed contain various steps tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper will focus on the development and application of a new MicroRaman probe for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid in solutions relevant to used nuclear fuel reprocessing. Previous research has successfully demonstrated the applicability on the macroscopic scale, using sample probes requiring larger solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 μL or less, which is a scale reduction of over 1,000 fold in sample size. Finally, this paper will summarize our current work in this area including: comparisons between the macroscopic and microscopic probes for detection limits, optimized channel focusing, and application in a flow cell with varying levels of HNO3, and UO2(NO3)2.

  6. MicroRaman Measurements for Nuclear Fuel Reprocessing Applications

    SciTech Connect

    Casella, Amanda; Lines, Amanda; Nelson, Gilbert; Bello, Job; Bryan, Samuel

    2016-01-01

    Treatment and reuse of used nuclear fuel is a key component in closing the nuclear fuel cycle. Solvent extraction reprocessing methods that have been developed contain various steps tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper will focus on the development and application of a new MicroRaman probe for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid in solutions relevant to used nuclear fuel reprocessing. Previous research has successfully demonstrated the applicability on the macroscopic scale, using sample probes requiring larger solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 μL or less, which is a scale reduction of over 1,000 fold in sample size. This paper will summarize our current work in this area including: comparisons between the macroscopic and microscopic probes for detection limits, optimized channel focusing, and application in a flow cell with varying levels of HNO3, and UO2(NO3)2.

  7. Vugraph presentations of the fourth DOE Industry/University/Lab Forum on Robotics for Environmental Restoration and Waste Management

    SciTech Connect

    Not Available

    1993-10-01

    This document is a compilation of various presentations from the Fourth DOE Industry/University/Lab Forum on Robotics for Environmental Restoration and Waste Management held in Albuquerque, New Mexico July 19--21, 1993. Separate abstracts were prepared for each presentation of this report.

  8. Ecological assessments at DOE hazardous waste sites: Current procedures and problems

    SciTech Connect

    Hlohowskyj, I.; Krummel, J.R.; Irving, J.S.; Vinikour, W.S.

    1989-01-01

    Major actions at US Department of Energy (DOE) hazardous waste sites require CERCLA compliance that meets NEPA considerations. Although NEPA compliance includes ecological considerations, neither the Council on Environmental Quality (CEQ) nor the DOE provide detailed guidance for conducting ecological assessments under NEPA. However, the identification of the form and magnitude of potential ecological impacts associated with a proposed action is directly dependent on the quality of the baseline data available for a particular site. Using the Surplus Facilities Management Program Weldon Spring site as an example, we discuss the collection of baseline ecological data for the site. This site is surrounded by approximately 17,000 acres of wildlife area. Available wildlife data consisted of qualitative, county-level species lists, and vegetation data was in the form of a regional qualitative narrative. Detailed site-specific occurrence data for listed species and high quality natural communities was provided by the Missouri Department of Conservation Heritage data base. 30 refs., 1 fig., 1 tab.

  9. 1995 Annual report on waste generation and waste mainization progress as required by DOE order 5400.1, Hanford site

    SciTech Connect

    Betsch, M.D.

    1996-09-24

    While waste generation numbers are important, the true measure of success is waste minimized. Many Waste Minimization/Pollution Prevention (WMin/P2) successes at the Hanford Site occur every day without formal recognition as pollution prevention, as they have become part of a culture of best management practices. As an example, the success of the excess and reuse program, both informal and formal, documents the Wmin/P2 culture that exists in the pollution prevention representatives and employees at the facilities.

  10. Reprocessing of anaesthetic and ventilatory equipment.

    PubMed

    Geiss, H K

    1995-06-01

    Uniform and standardized recommendations for reprocessing of anaesthetic and ventilatory equipment are still lacking. The uncertainty in this field is underscored by the various methods which are described in the literature which include pasteurization, immersion baths, formaldehyde cabinets, automated washers/disinfectors and sterilization procedures like autoclaving, ethylene oxide and gaseous formaldehyde. Based on the classification of anaesthetic and ventilatory equipment as semi-critical items, high level disinfection must be regarded as the appropriate decontamination procedure. In contrast to automated washers the other above-mentioned disinfection procedures lack an integrated and all inclusive reprocessing cycle which consists of cleaning, disinfection, rinsing and drying. In view of the increasing demands of employee safety, environmental suitability, cost-effectiveness and quality assurance in hospital hygiene, only automated washers/disinfectors--either based on hot water disinfection or chemothermic processing--fulfil the basic requirements for safe and standardized reprocessing of anaesthetic and ventilatory equipment.

  11. Review of DOE waste package program. Subtask 1.1. National waste package program, April-September 1983. Volume 5

    SciTech Connect

    Soo, P.

    1984-08-01

    The current effort is part of an ongoing task to review the national high-level waste package effort. It includes evaluations of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt, basalt, and tuff repositories. In the current Biannual Report a section on carbon steel container corrosion has been included to complement prior work on TiCode-12 and Type 304 stainless steel. The use of crushed tuff as a packing material is discussed and waste package component interaction test data are included. Licensing data requirements to estimate the degree of compliance with NRC performance objectives are specified. 41 figures, 24 tables.

  12. Review of DOE waste package program. Subtask 1.1 - National Waste Package Program, October 1983-March 1984. Volume 6

    SciTech Connect

    Soo, P.

    1985-03-01

    The present effort is part of an ongoing task to review the national high-level waste package effort. It includes evaluation of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt, basalt, tuff, and granite repositories. In the current Biannual Report a review of progress in the new crystalline repository (granite) program is described. Other foreign data for this host rock have also been outlined where relevant. The use of crushed salt, and bentonite- and zeolite-containing packing materials is discussed. The effects of temperature and gamma irradiation are shown to be important with respect to defining the localized environmental conditions around a waste package and the long-term integrity of the packing.

  13. Consolidated Fuel Reprocessing Program: Progress report for period October 1 to December 31, 1986

    SciTech Connect

    Groenier, W.S.; Meacham, S.A.; Stradley, J.G.

    1987-06-01

    All research and development (R and D) on civilian power reactor oxide fuel reprocessing in the United States is managed under the Consolidated Fuel Reprocessing Program (CFRP) centered at Oak Ridge National Laboratory (ORNL). A prime focus of present work is on technical exchanges and collaboration with other countries. In this context, the US Department of Energy (DOE) is in the process of negotiating a major collaboration with Japan. Both work associated with the foreign exchanges and collaboration and some on-going work are reported in overview fashion in this series of quarterly progress reports.

  14. International technology catalogue: Foreign technologies to support the environmental restoration and waste management needs of the DOE complex

    SciTech Connect

    Matalucci, R.V.; Jimenez, R.D.; Esparza-Baca, C.

    1995-07-01

    This document represents a summary of 27 foreign-based environmental restoration and waste management technologies that have been screened and technically evaluated for application to the cleanup problems of the Department of Energy (DOE) nuclear weapons complex. The evaluation of these technologies was initiated in 1992 and completed in 1995 under the DOE`s International Technology Coordination Program of the Office of Technology Development. A methodology was developed for conducting a country-by-country survey of several regions of the world where specific environmental technology capabilities and market potential were investigated. The countries that were selected from a rank-ordering process for the survey included: then West Germany, the Netherlands, France, Japan, Taiwan, the Czech and Slovak Republics, and the Former Soviet Union. The notably innovative foreign technologies included in this document were screened initially from a list of several hundred, and then evaluated based on criteria that examined for level of maturity, suitability to the DOE needs, and for potential cost effective application at a DOE site. Each of the selected foreign technologies that were evaluated in this effort for DOE application were subsequently matched with site-specific environmental problem units across the DOE complex using the Technology Needs Assessment CROSSWALK Report. For ease of tracking these technologies to site problem units, and to facilitate their input into the DOE EnviroTRADE Information System, they were categorized into the following three areas: (1) characterization, monitoring and sensors, (2) waste treatment and separations, and (3) waste containment. Technical data profiles regarding these technologies include title and description, performance information, development status, key regulatory considerations, intellectual property rights, institute and contact personnel, and references.

  15. Principles of qualification of the PAMELA process for the vitrification of HLLW of the Karlsruhe Reprocessing Plant (WAK)

    SciTech Connect

    Ewest, E.; Kunz, W.; Demonie, M.; Martens, B.R.; Goeyse, M. de

    1993-12-31

    After having reprocessed about 211 t of Uranium, the WAK Karlsruhe Pilot Reprocessing Plant was shut down in 1991. While all the other radioactive waste arising from reprocessing were conditioned parallel to the plant operation, some 60 m{sup 3} of High Level Liquid Waste (HLLW) having a specific {beta}, {gamma}-activity of about 2 E13 Bq/l is not yet processed. The waste is stored in two tanks, having a different activity level and chemical composition. In order to obtain a uniform product both solutions will be blended in a suitable way. It is intended to ship this waste to the PAMELA Vitrification Plant located on the Belgoprocess (BP) site in Dessel, Belgium. The vitrified product shall be returned to Germany. As from October 1986 until September 1991, the facility was operated by a mixed Belgian-German crew under the responsibility of BP for the vitrification of 800 m{sup 3} of HEWC (concentrated high-level waste from the reprocessing of high-enriched uranium fuels). Between October 1, 1985 and September 1, 1991, the total amount of 907 m{sup 3} of EUROCHEMIC HLLW has been successfully vitrified and conditioned in about 2,200 canisters. The typical composition of the different types of glass products are compared with the design data of the WAK glass product.

  16. Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

    SciTech Connect

    Carlsen, Brett; Chiguer, Mustapha; Grahn, Per; Sampson, Michele; Wolff, Dietmar; Bevilaqua, Arturo; Wasinger, Karl; Saegusa, Toshiari; Seelev, Igor

    2016-09-01

    Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years until reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is primarily on

  17. TIGA Tide Gauge Data Reprocessing at GFZ

    NASA Astrophysics Data System (ADS)

    Deng, Zhiguo; Schöne, Tilo; Gendt, Gerd

    2014-05-01

    To analyse the tide gauge measurements for the purpose of global long-term sea level change research a well-defined absolute reference frame is required by oceanographic community. To create such frame the data from a global GNSS network located at or near tide gauges are processed. For analyzing the GNSS data on a preferably continuous basis the International GNSS Service (IGS) Tide Gauge Benchmark Monitoring Working Group (TIGA-WG) is responsible. As one of the TIGA Analysis Centers the German Research Centre for Geosciences (GFZ) is contributing to the IGS TIGA Reprocessing Campaign. The solutions of the TIGA Reprocessing Campaign will also contribute to 2nd IGS Data Reprocessing Campaign with GFZ IGS reprocessing solution. After the first IGS reprocessing finished in 2010 some improvements were implemented into the latest GFZ software version EPOS.P8: reference frame IGb08 based on ITRF2008, antenna calibration igs08.atx, geopotential model (EGM2008), higher-order ionospheric effects, new a priori meteorological model (GPT2), VMF mapping function, and other minor improvements. GPS data of the globally distributed tracking network of 794 stations for the time span from 1994 until end of 2012 are used for the TIGA reprocessing. To handle such large network a new processing strategy is developed and described in detail. In the TIGA reprocessing the GPS@TIGA data are processed in precise point positioning (PPP) mode to clean data using the IGS reprocessing orbit and clock products. To validate the quality of the PPP coordinate results the rates of 80 GPS@TIGA station vertical movement are estimated from the PPP results using Maximum Likelihood Estimation (MLE) method. The rates are compared with the solution of University of LaRochelle Consortium (ULR) (named ULR5). 56 of the 80 stations have a difference of the vertical velocities below 1 mm/yr. The error bars of PPP rates are significant larger than those of ULR5, which indicates large time correlated noise in

  18. Thirteenth annual U.S. DOE low-level radioactive waste management conference: Proceedings

    SciTech Connect

    1991-12-31

    The 40 papers in this document comprise the proceedings of the Department of Energy`s Thirteenth Annual Low-Level Radioactive Waste Management Conference that was held in Atlanta, Georgia, on November 19--21, 1991. General subjects addressed during the conference included: disposal facility design; greater-than-class C low-level waste; public acceptance considerations; waste certification; site characterization; performance assessment; licensing and documentation; emerging low-level waste technologies; waste minimization; mixed waste; tracking and transportation; storage; and regulatory changes. Papers have been processed separately for inclusion on the data base.

  19. MATRIX 1 RESULTS OF THE FY07 ENHANCED DOE HIGH-LEVEL WASTE MELTER THROUGHPUT STUDIES AT SRNL

    SciTech Connect

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

    2008-09-23

    High-level waste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). It has been proposed that a team of glass formulation and processing experts at the Pacific Northwest National Laboratory (PNNL), Savannah River National Laboratory (SRNL), and Vitreous State Laboratory (VSL) at Catholic University of America develop a systematic approach to increase HLW throughput (by increasing WL with minimal or positive impacts on melt rate). Programmatically, this task is aimed at proof-of-principle testing and the development of tools to improve waste loading and melt rate, which will lead to higher waste throughput. The following four specific tasks have been proposed to meet this programmatic objective: (1) Integration and Oversight, (2) Crystal Accumulation Modeling (led by PNNL)/Higher Waste Loading Glasses (led by SRNL), (3) Melt Rate Evaluation and Modeling, and (4) Melter Scale Demonstrations. The details of these tasks can be found in the associated task plan WSRC-STI-2007-00483. The current study is focused on Task 2 (crystal accumulation modeling and higher waste loading glasses) and involves glass formulation and physical property testing by both PNNL and SRNL (as defined in the PNNL and SRNL test plans). The intent of this report is to document the chemical composition and Product Consistency Test (PCT) results and statistical analysis of PNNL's Test Matrix 1 glasses. Note that this document is only a compilation of the data collected by SRNL for PNNL's glasses in support of this task and no conclusions will be drawn.

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

    SciTech Connect

    MANN, F.M.

    2000-08-01

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

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

    SciTech Connect

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

    1993-06-01

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

  2. Updated Site Response Analyses for the Waste Treatment Plant, DOE Hanford, Site, Washington.

    SciTech Connect

    Youngs, Robert R.

    2007-06-29

    This document describes the calculations performed to develop updated relative amplification functions for the Waste Treatment and Immobilization Plant (WTP) facility at the DOE Hanford Site, Washington State. The original 2,000-year return period design spectra for the WTP were based on the results of a probabilistic seismic hazard analysis (PSHA) performed for the DOE Hanford Site by Geomatrix (1996). Geomatrix (1996) performed the PSHA using empirical soil-site ground motion models based primarily on recordings from California. As part of that study, site response analyses were performed to evaluate ground motions at the Hanford sites and California deep soil sites. As described in Appendix A of Geomatrix (1996), characteristic site profiles and dynamic soil properties representative of conditions at various Hanford sites and California deep soil strong motion recording stations were defined. Relative site responses of the Hanford profiles and California profiles were then compared. Based on the results of those site response analyses, it was concluded that ground motions at the Hanford sites underlain by deep soil deposits are similar in character to those on California deep soil sites and it was judged appropriate to use empirical deep soil site attenuation relationships based primarily on California ground motion data to develop design spectra for the Hanford sites. In a subsequent analysis, Geomatrix (2003) updated the site response analyses of Geomatrix (1996, Appendix A) to incorporate randomization of the California and Hanford profiles. The results of that analysis also led to the conclusion that the response of the Hanford profiles was similar to the response of deep soil sites in California.

  3. Updated Site Response Analyses for the Waste Treatment Plant, DOE Hanford Site, Washington

    SciTech Connect

    Youngs RR

    2007-06-01

    This document describes the calculations performed to develop updated relative amplification functions for the Waste Treatment and Immobilization Plant (WTP) facility at the DOE Hanford Site, Washington State. The original 2,000-year return period design spectra for the WTP were based on the results of a probabilistic seismic hazard analysis (PSHA) performed for the DOE Hanford Site by Geomatrix (1996). Geomatrix (1996) performed the PSHA using empirical soil-site ground motion models based primarily on recordings from California. As part of that study, site response analyses were performed to evaluate ground motions at the Hanford sites and California deep soil sites. As described in Appendix A of Geomatrix (1996), characteristic site profiles and dynamic soil properties representative of conditions at various Hanford sites and California deep soil strong motion recording stations were defined. Relative site responses of the Hanford profiles and California profiles were then compared. Based on the results of those site response analyses, it was concluded that ground motions at the Hanford sites underlain by deep soil deposits are similar in character to those on California deep soil sites and it was judged appropriate to use empirical deep soil site attenuation relationships based primarily on California ground motion data to develop design spectra for the Hanford sites. In a subsequent analysis, Geomatrix (2003) updated the site response analyses of Geomatrix (1996, Appendix A) to incorporate randomization of the California and Hanford profiles. The results of that analysis also led to the conclusion that the response of the Hanford profiles was similar to the response of deep soil sites in California.

  4. 1997 annual report on waste generation and waste minimization progress as required by DOE Order 5400.1, Hanford Site

    SciTech Connect

    Segall, P.

    1998-04-13

    Hanford`s missions are to safely clean up and manage the site`s legacy wastes, and to develop and deploy science and technology. Through these missions Hanford will contribute to economic diversification of the region. Hanford`s environmental management or cleanup mission is to protect the health and safety of the public, workers, and the environment; control hazardous materials; and utilize the assets (people, infra structure, site) for other missions. Hanford`s science and technology mission is to develop and deploy science and technology in the service of the nation including stewardship of the Hanford Site. Pollution Prevention is a key to the success of these missions by reducing the amount of waste to be managed and identifying/implementing cost effective waste reduction projects. Hanford`s original mission, the production of nuclear materials for the nation`s defense programs, lasted more than 40 years, and like most manufacturing operations, Hanford`s operations generated large quantities of waste and pollution. However, the by-products from Hanford operations pose unique problems like radiation hazards, vast volumes of contaminated water and soil, and many contaminated structures including reactors, chemical plants and evaporation ponds. The cleanup activity is an immense and challenging undertaking, which includes characterization and decommissioning of 149 single shell storage tanks, treating 28 double shell tanks, safely disposing of over 2,100 metric tons of spent nuclear fuel stored on site, removing numerous structures, and dealing with significant solid waste, ground water, and land restoration issues.

  5. A decision methodology for the evaluation of mixed low-level radioactive waste management options for DOE sites

    SciTech Connect

    Bassi, J.; Abashian, M.S.; Chakraborti, S.; Devarakonda, M.; Djordjevic, S.M.

    1993-03-01

    Currently, many DOE sites are developing site-specific solutions to manage their mixed low-level wastes. These site-specific MLLW programs often result in duplication of efforts between the different sites, and consequently, inefficient use of DOE system resources. A nationally integrated program for MLLW eliminates unnecessary duplication of effort, but requires a comprehensive analysis of waste management options to ensure that all site issues are addressed. A methodology for comprehensive analysis of the complete DOE MLLW system is being developed by DOE-HQ to establish an integrated and standardized solution for managing MLLW. To be effective, the comprehensive systems analysis must consider all aspects of MLLW management from cradle-to-grave (i.e. from MLLW generation to disposal). The results of the analysis will include recommendations for alternative management options for the complete DOE MLLW system based on various components such as effectiveness, cost, health and safety risks, and the probability of regulatory acceptance for an option. Because of the diverse nature of these various components and the associated difficulties in comparing between them, a decision methodology is being developed that will integrate the above components into a single evaluation scheme for performing relative comparisons between different MLLW management options. The remainder of this paper provides an overview of the roles and responsibilities of the various participants of the DOE MLLW Program, and discusses in detail the components involved in the development of the decision methodology for a comprehensive systems analysis.

  6. Summary of nuclear fuel reprocessing activities around the world

    SciTech Connect

    Mellinger, P.J.; Harmon, K.M.; Lakey, L.T.

    1984-11-01

    This review of international practices for nuclear fuel reprocessing was prepared to provide a nontechnical summary of the current status of nuclear fuel reprocessing activities around the world. The sources of information are widely varied.

  7. STATUS OF EPA/DOE MOU TECHNICAL WORKGROUP ACTIVITIES: HG WASTE TREATMENT

    EPA Science Inventory

    EPA's Land Disposal Restrictions program currently has technology-specific treatment standards for hazardous wastes containing greater than or equal to 260ppm total mercury (Hg) (i.e., high Hg subcategory wastes). The treatment standards specify RMERC for high Hg subcategory wast...

  8. STATUS OF EPA/DOE MOU TECHNICAL WORKGROUP ACTIVITIES: HG WASTE TREATMENT

    EPA Science Inventory

    EPA's Land Disposal Restrictions program currently has technology-specific treatment standards for hazardous wastes containing greater than or equal to 260ppm total mercury (Hg) (i.e., high Hg subcategory wastes). The treatment standards specify RMERC for high Hg subcategory wast...

  9. Development of nuclear analysis capabilities for DOE waste management activities. 1998 annual progress report

    SciTech Connect

    Parks, C.V.; DeHart, M.D.; Broadhead, B.L.; Hopper, C.M.

    1998-06-01

    'The objective of this project is to develop and demonstrate prototypic analysis capabilities that can be used by the nuclear safety analysis practitioners to: (1) demonstrate a more thorough understanding of the underlying physics phenomena that can lead to improved reliability and defensibility of safety evaluations; and (2) optimize operations related to the handling, storage, transportation, and disposal of fissile material and DOE spent fuel. To address these problems, the project will investigate the implementation of sensitivity and uncertainty methods within existing Monte Carlo codes used for criticality safety analyses, as well as within a new deterministic code that allows specification of arbitrary grids to accurately model the geometry details required in a criticality safety analysis. This capability can facilitate improved estimations of the required subcritical margin and potentially enable the use of a broader range of experiments in the validation process. The new arbitrary-grid radiation transport code will also enable detailed geometric modeling valuable for improved accuracy in application to a myriad of other problems related to waste characterization. Application to these problems will also be explored. This report summarizes the progress achieved after only seven months of work on a three-year project.'

  10. EMSP project summary (Project ID: 60077): Development of nuclear analysis capabilities for DOE waste management activities

    SciTech Connect

    Parks, C.V.; Rearden, B.T.; DeHart, M.D.; Broadhead, B.L.; Hopper, C.M.; Petrie, L.M.

    2000-02-01

    The objective of this project is to develop and demonstrate prototypical analysis capabilities that can be used by nuclear safety analysis practitioners to: (1) demonstrate a more thorough understanding of the underlying physics phenomena that can lead to improved reliability and defensibility of safety evaluations; and (2) optimize operations related to the handling, storage, transportation, and disposal of fissile material and DOE spent fuel. To address these problems, this project has been investigating the implementation of sensitivity and uncertainty methods within existing Monte Carlo codes used for criticality safety analyses. It is also investigating the use of a new deterministic code that allows for specification of arbitrary grids to accurately model geometric details required in a criticality safety analysis. This capability can facilitate improved estimations of the required subcritical margin and potentially enable the use of a broader range of experiments in the validation process. The new arbitrary-grid radiation transport code will also enable detailed geometric modeling valuable for improved accuracy in application to a myriad of other problems related to waste characterization. Application to these problems will also be explored.

  11. Design of experiment (DOE) based screening of factors affecting municipal solid waste (MSW) composting.

    PubMed

    Kazemi, Khoshrooz; Zhang, Baiyu; Lye, Leonard M; Cai, Qinghong; Cao, Tong

    2016-12-01

    A design of experiment (DOE) based methodology was adopted in this study to investigate the effects of multiple factors and their interactions on the performance of a municipal solid waste (MSW) composting process. The impact of four factors, carbon/nitrogen ratio (C/N), moisture content (MC), type of bulking agent (BA) and aeration rate (AR) on the maturity, stability and toxicity of compost product was investigated. The statistically significant factors were identified using final C/N, germination index (GI) and especially the enzyme activities as responses. Experimental results validated the use of enzyme activities as proper indices during the course of composting. Maximum enzyme activities occurred during the active phase of decomposition. MC has a significant effect on dehydrogenase activity (DGH), β-glucosidase activity (BGH), phosphodiesterase activity (PDE) and the final moisture content of the compost. C/N is statistically significant for final C/N, DGH, BGH, and GI. The results provided guidance to optimize a MSW composting system that will lead to increased decomposition rate and the production of more stable and mature compost.

  12. Assessment of thermal analysis software for the DOE Office of Civilian Radioactive Waste Management

    SciTech Connect

    Williams, P.T.; Graham, R.F.; Lagerberg, G.N.; Chung, T.C.

    1989-07-01

    This assessment uses several recent assessments and the more general code compilations that have been completed to produce a list of 116 codes that can be used for thermal analysis. This list is then compared with criteria prepared especially for the Department of Energy Office of Civilian Radioactive Waste Management (DOE/OCRWM). Based on these criteria, fifteen codes are narrowed to three primary codes and four secondary codes for use by the OCRWM thermal analyst. The analyst is cautioned that since no single code is sufficient for all applications, a code must be selected based upon the predominate heat transfer mode of the problem to be solved, but the codes suggested in this report have been used successfully for a range of OCRWM applications. The report concludes with a series of recommendations for additional work of which the major points include the following: The codes suggested by this report must be benchmarked with the existing US and international problems and validated when possible; An interactive code selection tool could be developed or, perhaps even more useful, a users group could be supported to ensure the proper selection of thermal codes and dissemination of information on the latest version; The status of the 116 codes identified by this report should be verified, and methods for maintaining the still active codes must be established; and special capabilities of each code in phase change, convection and radiation should be improved to better enable the thermal analyst to model OCRWM applications. 37 refs., 3 figs., 12 tabs.

  13. CHARACTERIZING DOE HANFORD SITE WASTE ENCAPSULATION STORAGE FACILITY CELLS USING RADBALL

    SciTech Connect

    Farfan, E.; Coleman, R.

    2011-03-31

    RadBall{trademark} is a novel technology that can locate and quantify unknown radioactive hazards within contaminated areas, hot cells, and gloveboxes. The device consists of a colander-like outer tungsten collimator that houses a radiation-sensitive polymer semi-sphere. The collimator has a number of small holes with tungsten inserts; as a result, specific areas of the polymer are exposed to radiation becoming increasingly more opaque in proportion to the absorbed dose. The polymer semi-sphere is imaged in an optical computed tomography scanner that produces a high resolution 3D map of optical attenuation coefficients. A subsequent analysis of the optical attenuation data using a reverse ray tracing or backprojection technique provides information on the spatial distribution of gamma-ray sources in a given area forming a 3D characterization of the area of interest. RadBall{trademark} was originally designed for dry deployments and several tests, completed at Savannah River National Laboratory and Oak Ridge National Laboratory, substantiate its modeled capabilities. This study involves the investigation of the RadBall{trademark} technology during four submerged deployments in two water filled cells at the DOE Hanford Site's Waste Encapsulation Storage Facility.

  14. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 21 Food and Drugs 4 2012-04-01 2012-04-01 false Reprocessing. 211.115 Section 211.115 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) DRUGS: GENERAL CURRENT GOOD MANUFACTURING PRACTICE FOR FINISHED PHARMACEUTICALS Production and Process Controls § 211.115...

  15. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 21 Food and Drugs 4 2013-04-01 2013-04-01 false Reprocessing. 211.115 Section 211.115 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) DRUGS: GENERAL CURRENT GOOD MANUFACTURING PRACTICE FOR FINISHED PHARMACEUTICALS Production and Process Controls § 211.115...

  16. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 21 Food and Drugs 4 2014-04-01 2014-04-01 false Reprocessing. 211.115 Section 211.115 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) DRUGS: GENERAL CURRENT GOOD MANUFACTURING PRACTICE FOR FINISHED PHARMACEUTICALS Production and Process Controls § 211.115...

  17. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 21 Food and Drugs 4 2011-04-01 2011-04-01 false Reprocessing. 211.115 Section 211.115 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) DRUGS: GENERAL CURRENT GOOD MANUFACTURING PRACTICE FOR FINISHED PHARMACEUTICALS Production and Process Controls § 211.115...

  18. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 21 Food and Drugs 4 2010-04-01 2010-04-01 false Reprocessing. 211.115 Section 211.115 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) DRUGS: GENERAL CURRENT GOOD MANUFACTURING PRACTICE FOR FINISHED PHARMACEUTICALS Production and Process Controls § 211.115...

  19. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    SciTech Connect

    Jubin, R. T.; Strachan, D. M.; Ilas, G.; Spencer, B. B.; Soelberg, N. R.

    2014-09-01

    In nuclear fuel reprocessing, various radioactive elements enter the gas phase from the unit operations found in the reprocessing facility. In previous reports, the pathways and required removal were discussed for four radionuclides known to be volatile, 14C, 3H, 129I, and 85Kr. Other, less volatile isotopes can also report to the off-gas streams in a reprocessing facility. These were reported to be isotopes of Cs, Cd, Ru, Sb, Tc, and Te. In this report, an effort is made to determine which, if any, of 24 semivolatile radionuclides could be released from a reprocessing plant and, if so, what would be the likely quantities released. As part of this study of semivolatile elements, the amount of each generated during fission is included as part of the assessment for the need to control their emission. Also included in this study is the assessment of the cooling time (time out of reactor) before the fuel is processed. This aspect is important for the short-lived isotopes shown in the list, especially for cooling times approaching 10 y. The approach taken in this study was to determine if semivolatile radionuclides need to be included in a list of gas-phase radionuclides that might need to be removed to meet Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. A list of possible elements was developed through a literature search and through knowledge and literature on the chemical processes in typical aqueous processing of nuclear fuels. A long list of possible radionuclides present in irradiated fuel was generated and then trimmed by considering isotope half-life and calculating the dose from each to a maximum exposed individual with the US EPA airborne radiological dispersion and risk assessment code CAP88 (Rosnick 1992) to yield a short list of elements that actually need to be considered for control because they require high decontamination factors to meet a reasonable fraction of the regulated release. Each of these elements is

  20. Effects of reprocessing on nanoalumina polymer composites

    NASA Astrophysics Data System (ADS)

    Huang, Chunchia

    The life cycle of reprocessed polymer nanocomposites is a critical factor associated with their growing use, but the limited work on reprocessing of nanocomposites has focused solely on the effects of organoclays. This research investigated of the structure and property changes during reprocessing of polypropylene (PP) and polycarbonate (PC) nanocomposites with 3 wt. % nanoalumina. Neat PP and PC were used as controls. Reprocessing of the neat polymers and nanocomposites produced no indication of oxidation (in FTIR), no changes in the glass transition temperature of PC and the melting temperatures of PP, and no changes in thermal stability (as measured using thermogravimetric analysis). Significant decreases, however, occurred in the melt viscosity of the materials. The introduction of nanoalumina during twin screw extrusion also produced a significant decrease in the viscosity and a 10°C decrease in the glass transition temperature of the PC nanocomposite. Color changes did not correspond to the chain scission in PP and PC; neat PP and PP composite yellowed, neat PC turned brown, and the PC nanocomposite did not change color. Dispersion of the nanoalumina in both PP and PC improved with repeated reprocessing, the crystallinity in the PP/nanoalumina composites remained constant. The Young's moduli of both the PP and PP/A12O3 nanocomposite were similar, whereas the Young's modulus values of the PC/A1 2O3 nanocomposite was slighted lower than that of the neat PC. In contrast, the elongations at break of the PP/A12O3 and PC/A12O3 nanocomposites were, respectively, 50% and 16% of the values measured for the neat resins. All modulus and elongation at break values, however, remained constant over five reprocessing cycles. This behavior suggests that the major degradation mechanism during reprocessing of neat PP and PP nanocomposites was thermal-mechanical polymer chain scission and that the nanoalumina enhanced this degradation in the PC/nanoalumina. This additional

  1. EOS Data Products Latency and Reprocessing Evaluation

    NASA Astrophysics Data System (ADS)

    Ramapriyan, H. K.; Wanchoo, L.

    2012-12-01

    NASA's Earth Observing System (EOS) Data and Information System (EOSDIS) program has been processing, archiving, and distributing EOS data since the launch of Terra platform in 1999. The EOSDIS Distributed Active Archive Centers (DAACs) and Science-Investigator-led Processing Systems (SIPSs) are generating over 5000 unique products with a daily average volume of 1.7 Petabytes. Initially EOSDIS had requirements to make process data products within 24 hours of receiving all inputs needed for generating them. Thus, generally, the latency would be slightly over 24 and 48 hours after satellite data acquisition, respectively, for Level 1 and Level 2 products. Due to budgetary constraints these requirements were relaxed, with the requirement being to avoid a growing backlog of unprocessed data. However, the data providers have been generating these products in as timely a manner as possible. The reduction in costs of computing hardware has helped considerably. It is of interest to analyze the actual latencies achieved over the past several years in processing and inserting the data products into the EOSDIS archives for the users to support various scientific studies such as land processes, oceanography, hydrology, atmospheric science, cryospheric science, etc. The instrument science teams have continuously evaluated the data products since the launches of EOS satellites and improved the science algorithms to provide high quality products. Data providers have periodically reprocessed the previously acquired data with these improved algorithms. The reprocessing campaigns run for an extended time period in parallel with forward processing, since all data starting from the beginning of the mission need to be reprocessed. Each reprocessing activity involves more data than the previous reprocessing. The historical record of the reprocessing times would be of interest to future missions, especially those involving large volumes of data and/or computational loads due to

  2. Nuclear Fuel Reprocessing: U.S. Policy Development

    DTIC Science & Technology

    2006-11-29

    Nuclear Fuel Reprocessing: U.S. Policy Development Anthony Andrews Specialist in Industrial Engineering and Infrastructure Policy Resources, Science...separate and recover fissionable plutonium from irradiated nuclear fuel. In the early stage of commercial nuclear power, reprocessing was thought essential...to supplying nuclear fuel. Federally sponsored breeder reactor development included research into advanced reprocessing technology. Several

  3. Alternative Waste Forms for Electro-Chemical Salt Waste

    SciTech Connect

    Crum, Jarrod V.; Sundaram, S. K.; Riley, Brian J.; Matyas, Josef; Arreguin, Shelly A.; Vienna, John D.

    2009-10-28

    This study was undertaken to examine alternate crystalline (ceramic/mineral) and glass waste forms for immobilizing spent salt from the Advanced Fuel Cycle Initiative (AFCI) electrochemical separations process. The AFCI is a program sponsored by U.S. Department of Energy (DOE) to develop and demonstrate a process for recycling spent nuclear fuel (SNF). The electrochemical process is a molten salt process for the reprocessing of spent nuclear fuel in an electrorefiner and generates spent salt that is contaminated with alkali, alkaline earths, and lanthanide fission products (FP) that must either be cleaned of fission products or eventually replaced with new salt to maintain separations efficiency. Currently, these spent salts are mixed with zeolite to form sodalite in a glass-bonded waste form. The focus of this study was to investigate alternate waste forms to immobilize spent salt. On a mole basis, the spent salt is dominated by alkali and Cl with minor amounts of alkaline earth and lanthanides. In the study reported here, we made an effort to explore glass systems that are more compatible with Cl and have not been previously considered for use as waste forms. In addition, alternate methods were explored with the hope of finding a way to produce a sodalite that is more accepting of as many FP present in the spent salt as possible. This study was done to investigate two different options: (1) alternate glass families that incorporate increased concentrations of Cl; and (2) alternate methods to produce a mineral waste form.

  4. Waste disposal options report. Volume 1

    SciTech Connect

    Russell, N.E.; McDonald, T.G.; Banaee, J.; Barnes, C.M.; Fish, L.W.; Losinski, S.J.; Peterson, H.K.; Sterbentz, J.W.; Wenzel, D.R.

    1998-02-01

    This report summarizes the potential options for the processing and disposal of mixed waste generated by reprocessing spent nuclear fuel at the Idaho Chemical Processing Plant. It compares the proposed waste-immobilization processes, quantifies and characterizes the resulting waste forms, identifies potential disposal sites and their primary acceptance criteria, and addresses disposal issues for hazardous waste.

  5. Canister arrangement for storing radioactive waste

    DOEpatents

    Lorenzo, D.K.; Van Cleve, J.E. Jr.

    1980-04-23

    The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

  6. Canister arrangement for storing radioactive waste

    DOEpatents

    Lorenzo, Donald K.; Van Cleve, Jr., John E.

    1982-01-01

    The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

  7. Twenty-third DOE/NRC Nuclear Air Cleaning and Treatment Conference

    SciTech Connect

    Bellamy, R.R.; Hayes, J.J.; First, M.W.

    1995-03-24

    This paper presents the details of the Nuclear Air Cleaning and Treatment Conference held in Buffalo, New York during July 1994. Topics discussed include: nuclear air cleaning codes and standards; waste disposal; particulate filter developments; sampling and monitoring of process and effluent streams; off-gasses from fuel reprocessing; adsorbents and adsorption; accident control and analysis; revised source terms for power plant accidents; and the highlight of the conference concerned operations at the West Valley DOE facility where construction is underway to solidify radioactive wastes.

  8. Hanford Waste Physical and Rheological Properties: Data and Gaps

    SciTech Connect

    Wells, Beric E.; Kurath, Dean E.; Mahoney, Lenna A.; Onishi, Yasuo; Huckaby, James L.; Cooley, Scott K.; Burns, Carolyn A.; Buck, Edgar C.; Tingey, Joel M.; Daniel, Richard C.; Anderson, K. K.

    2011-08-01

    The Hanford Site in Washington State manages 177 underground storage tanks containing approximately 250,000 m3 of waste generated during past defense reprocessing and waste management operations. These tanks contain a mixture of sludge, saltcake and supernatant liquids. The insoluble sludge fraction of the waste consists of metal oxides and hydroxides and contains the bulk of many radionuclides such as the transuranic components and 90Sr. The saltcake, generated by extensive evaporation of aqueous solutions, consists primarily of dried sodium salts. The supernates consist of concentrated (5-15 M) aqueous solutions of sodium and potassium salts. The 177 storage tanks include 149 single-shell tanks (SSTs) and 28 double -hell tanks (DSTs). Ultimately the wastes need to be retrieved from the tanks for treatment and disposal. The SSTs contain minimal amounts of liquid wastes, and the Tank Operations Contractor is continuing a program of moving solid wastes from SSTs to interim storage in the DSTs. The Hanford DST system provides the staging location for waste feed delivery to the Department of Energy (DOE) Office of River Protection’s (ORP) Hanford Tank Waste Treatment and Immobilization Plant (WTP). The WTP is being designed and constructed to pretreat and then vitrify a large portion of the wastes in Hanford’s 177 underground waste storage tanks.

  9. ANL technical support program for DOE Environmental Restoration and Waste Management. Annual report, October 1991--September 1992

    SciTech Connect

    Bates, J.K.; Bradley, C.R.; Buck, E.C.; Cunnane, J.C.; Dietz, N.L.; Ebert, W.L.; Emery, J.W.; Feng, X.; Gerding, T.J.; Gong, M.; Hoh, J.C.; Mazer, J.J.; Wronkiewicz, D.J.; Bourcier, W.L.; Morgan, L.E.; Newton, L.; Nielsen, J.K.; Phillips, B.L.; Ewing, R.C.; Wang, L.; Li, H.; Tomozawa, M.

    1993-05-01

    A program was established for DOE Environmental Restoration and Waste Management (EM) to evaluate factors that are anticipated to affect waste glass reaction during repository disposal, especially in an unsaturated environment typical of what may be expected for the proposed Yucca Mountain repository site. This report covers progress in FY 1992 on the following tasks: 1. A compendium of the characteristics of high-level nuclear waste borosilicate glass has been written. 2. A critical review of important parameters that affect the reactivity of glass in an unsaturated environment is being prepared. 3. A series of tests has been started to evaluate the reactivity of fully radioactive glasses in a high-level waste repository environment and compare it to the reactivity of synthetic, nonradioactive glasses of similar composition. 4. The effect of radiation upon the durability of waste glasses at a high glass surface area-to-liquid volume (SA/V) ratio and a high gas-to-liquid volume ratio will be assessed. These tests address both vapor and high SA/V liquid conditions. 5. A series of tests is being performed to compare the extent of reaction of nuclear waste glasses at various SAN ratios. Such differences in the SAN ratio may significantly affect glass durability. 6. A series of natural analogue tests is being analyzed to demonstrate a meaningful relationship between experimental and natural alteration conditions. 7. Analytical electron microscopy (AEM), infrared spectroscopys and nuclear resonant profiling are being used to assess the glass/water reaction pathway by identifying intermediate phases that appear on the reacting glass. Additionally, colloids from the leach solutions are being studied using AEM. 8. A technical review of AEM results is being provided. 9. A study of water diffusion involving nuclear waste glasses is being performed. 10. A mechanistically based model is being developed to predict the performance of glass over repository-relevant time periods.

  10. Effect of reprocessing cycles on the degradation of PP/PBAT-thermoplastic starch blends.

    PubMed

    Oliveira, Thainá A; Oliveira, Rosimery R; Barbosa, Renata; Azevedo, Joyce B; Alves, Tatianny S

    2017-07-15

    The solid waste management problems caused by the accumulation of plastics require measures to mitigate the environmental damage, and mechanical recycling of plastics is among the possible solutions. In this context, the present study aimed to evaluate the effects of mechanical recycling on the properties of a polypropylene/poly(butylene adipate co-terephthalate)-thermoplastic starch blend (PP/PBAT-Thermoplastic starch blend) when it was subjected to seven reprocessing cycles by a single-screw extruder. The observations by infrared spectroscopy indicated that the chemical structures of the blend and the polypropylene matrix did not present significant changes with the reprocessing cycles. The X-ray diffraction analyses showed that the PP crystals were most affected when reprocessed in their pure form. The observations by thermogravimetry and differential scanning calorimetry indicated that the thermal stability of the blend was higher than that of polypropylene during the extrusion cycles. The scanning electron microscopy images indicated a weak interfacial interaction between the components of the blend, and the mechanical properties showed that the reprocessing improved the elasticity modulus and yield stress, with a consequential decrease of the impact strength. Copyright © 2017 Elsevier Ltd. All rights reserved.

  11. SUMMARY OF 2010 DOE EM INTERNATIONAL PROGRAM STUDIES OF WASTE GLASS MELT RATE ENHANCEMENT

    SciTech Connect

    Fox, K.; Marra, J.

    2011-01-19

    be determined. Overall, the results show an excellent potential for these additives to significantly improve waste throughput for DOE vitrification facilities. A complete report from KRI is included as an appendix to this document.

  12. An Overview of Food Loss and Waste: why does it Matter?

    NASA Astrophysics Data System (ADS)

    Ghosh, Purabi R.; Sharma, Shashi B.; Haigh, Yvonne T.; Evers, A. L. Barbara; Ho, Goen

    2015-10-01

    This paper provides an overview of food waste in the context of food security, resources management and environment health. It compares approaches taken by various governments, community groups, civil societies and private sector organisations to reduce food waste in the developed and developing countries. What constitutes ‘food waste’ is not as simple as it may appear due to diverse food waste measurement protocols and different data documentation methods used worldwide. There is a need to improve food waste data collection methods and implementation of effective strategies, policies and actions to reduce food waste. Global initiatives are urgently needed to: enhance awareness of the value of food; encourage countries to develop policies that motivate community and businesses to reduce food waste; encourage and provide assistance to needy countries for improving markets, transport and storage infrastructure to minimise food waste across the value chain; and, develop incentives that encourage businesses to donate food. In some countries, particularly in Europe, initiatives on food waste management have started to gain momentum. Food waste is a global problem and it needs urgent attention and integrated actions of stakeholders across the food value chain to develop global solutions for the present and future generations.

  13. Enclosure (from letter from EPA to DOE sent 3/26/04) - Technical and Regulatory Support Document: EPA's Remote Handled Waste Characterization Determination

    EPA Pesticide Factsheets

    Based on the requirements presented in 40 CFR 194.24(c )(2) to (4) and 194.22(a)(1) and using experience gained as part of the CH waste characterization program, EPA examined the DOE's RH Waste Characterization Proposal as presented in the WCPIP.

  14. International safeguards for reprocessing plants. Final report

    SciTech Connect

    Kratzer, M.; Scheinman, L.; Sievering, N.; Wonder, E.; Lipman, D.; Immerman, W.; Elliott, J.M.; Crane, F.

    1981-04-01

    Proliferation risks inherent in reprocessing show the need to employ technically effective safeguards which can detect, with a high degree of assurance and on a timely basis, the diversion of significant quantities of fissionable material. A balance must be struck between what is technically feasible and effective and what is institutionally acceptable. Purpose of this report is to examine the several technical approaches to safeguards in light of their prospective acceptability. This study defines the economic, political and institutional nature of the safeguards problem; surveys generically alternative technical approaches to international safeguards including their effectiveness and relative development; characterizes the institutional implications and uncertainties associated with the acceptance and implementation of each technical alternative; and integrates these assessments into a set of overall judgments on feasible directions for reprocessing plant safeguards systems.

  15. Asbestos hazard in the reprocessed textile industry

    SciTech Connect

    Quinn, M.M.; Kriebel, D.; Buiatti, E.; Paci, E.; Sini, S.; Vannucchi, G.; Zappa, M.

    1987-01-01

    Epidemiologic studies have identified an excess risk of lung cancer and mesothelioma among workers in the reprocessed textile industry in Prato, Italy. These studies suggested that there may have been asbestos hazard in this industry although exposure was not known to exist. An industrial hygiene investigation was conducted to determine whether there was previous or current asbestos exposure in the industry. Walk-through surveys, environmental sampling, process documentation, and management and worker interviews were conducted in 13 textile reprocessing establishments. Polypropylene bags that once contained asbestos were found in 2 of the 13. Asbestos bags were cut open and used to cover bales of rags which were then distributed throughout the world. Workers were exposed to asbestos while handling the bags which were contaminated with chrysotile, amosite, and crocidolite. Additional sources of asbestos exposure that may have existed in the past in the industry are also discussed.

  16. Reprocessing technology development for irradiated beryllium

    SciTech Connect

    Kawamura, H.; Sakamoto, N.; Tatenuma, K.

    1995-09-01

    At present, beryllium is under consideration as a main candidate material for neutron multiplier and plasma facing material in a fusion reactor. Therefore, it is necessary to develop the beryllium reprocessing technology for effective resource use. And, we have proposed reprocessing technology development on irradiated beryllium used in a fusion reactor. The preliminary reprocessing tests were performed using un-irradiated and irradiated beryllium. At first, we performed beryllium separation tests using un-irradiated beryllium specimens. Un-irradiated beryllium with beryllium oxide which is a main impurity and some other impurities were heat-treated under chlorine gas flow diluted with Ar gas. As the results high purity beryllium chloride was obtained in high yield. And it appeared that beryllium oxide and some other impurities were removed as the unreactive matter, and the other chloride impurities were separated by the difference of sublimation temperature on beryllium chloride. Next, we performed some kinds of beryllium purification tests from beryllium chloride. And, metallic beryllium could be recovered from beryllium chloride by the reduction with dry process. In addition, as the results of separation and purification tests using irradiated beryllium specimens, it appeared that separation efficiency of Co-60 from beryllium was above 96%. It is considered that about 4% Co-60 was carried from irradiated beryllium specimen in the form of cobalt chloride. And removal efficiency of tritium from irradiated beryllium was above 95%.

  17. Noble gas atmospheric monitoring at reprocessing facilities

    SciTech Connect

    Nakhleh, C.W.; Perry, R.T. Jr.; Poths, J.; Stanbro, W.D.; Wilson, W.B.; Fearey, B.L.

    1997-05-01

    The discovery in Iraq after the Gulf War of the existence of a large clandestine nuclear-weapon program has led to an across-the-board international effort, dubbed Programme 93+2, to improve the effectiveness and efficiency of International Atomic Energy Agency (IAEA) safeguards. One particularly significant potential change is the introduction of environmental monitoring (EM) techniques as an adjunct to traditional safeguards methods. Monitoring of stable noble gas (Kr, Xe) isotopic abundances at reprocessing plant stacks appears to be able to yield information on the burnup and type of the fuel being processed. To estimate the size of these signals, model calculations of the production of stable Kr, Xe nuclides in reactor fuel and the subsequent dilution of these nuclides in the plant stack are carried out for two case studies: reprocessing of PWR fuel with a burnup of 35 GWd/tU, and reprocessing of CAND fuel with a burnup of 1 GWd/tU. For each case, a maximum-likelihood analysis is used to determine the fuel burnup and type from the isotopic data.

  18. Effect of reprocessing cycles on the degradation of polypropylene copolymer filled with talc or montmorillonite during injection molding process

    SciTech Connect

    Demori, R.; Mauler, R. S.; Ashton, E.; Weschenfelder, V. F.; Cândido, L. H. A.; Kindlein, W.

    2015-05-22

    Mechanical recycling of polymeric materials is a favorable technique resulting in economic and environmental benefits, especially in the case of polymers with a high production volume as the polypropylene copolymer (PP). However, recycling by reprocessing techniques can lead to thermal, mechanical or thermo-oxidative degradation that can affect the structure of the polymer and subsequently the material properties. PP filled with montmorillonite (MMT) or talc are widely produced and studied, however, its degradation reactions by reprocessing cycles are poorly studied so far. In this study, the effects of reprocessing cycles in the structure and in the properties of the PP/MMT and PP/Talc were evaluated. The samples were mixed with 5% talc or MMT Cloisite C15A in a twin-screw extrusion. After extrusion, this filled material was submitted to five reprocessing cycles through an injection molding process. In order to evaluate the changes induced by reprocessing techniques, the samples were characterized by DSC, FT-IR, Izod impact and tensile strength tests. The study showed that Young modulus, elongation at brake and Izod impact were not affected by reprocessing cycles, except when using talc. In this case, the elongation at brake reduced until the fourth cycle, showing rigidity increase. The DSC results showed that melting and crystallization temperature were not affected. A comparison of FT-IR spectra of the reprocessed indicated that in both samples, between the first and the fifth cycle, no noticeable change has occurred. Thus, there is no evidence of thermo oxidative degradation. In general, these results suggest that PP reprocessing cycles using MMT or talc does not change the material properties until the fifth cycle.

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

    SciTech Connect

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

    2009-11-18

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

  20. An exposure assessment of radionuclide emissions associated with potential mixed-low level waste disposal facilities at fifteen DOE sites

    SciTech Connect

    Lombardi, D.A.; Socolof, M.L.

    1996-05-01

    A screening method was developed to compare the doses received via the atmospheric pathway at 15 potential DOE MLLW (mixed low-level waste) sites. Permissible waste concentrations were back calculated using the radioactivity NESHAP (National Emissions Standards for Hazardous Air Pollutants) in 40 FR 61 (DOE Order 5820.2A performance objective). Site-specific soil and meteorological data were used to determine permissible waste concentrations (PORK). For a particular radionuclide, perks for each site do not vary by more than one order of magnitude. perks of {sup 14}C are about six orders of magnitude more restrictive than perks of {sup 3}H because of differences in liquid/vapor partitioning, decay, and exposure dose. When comparing results from the atmospheric pathway to the water and intruder pathways, {sup 14}C disposal concentrations were limited by the atmospheric pathway for most arid sites; for {sup 3}H, the atmospheric pathway was not limiting at any of the sites. Results of this performance evaluation process are to be used for planning for siting of disposal facilities.

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

    SciTech Connect

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

    2008-01-23

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

  2. Design and Construction of Deinococcus Radiodurans for Biodegradation of Organic Toxins at Radioactive DOE Waste Sites

    SciTech Connect

    Michael J. Daly; Lawrence P. Wackett; James K. Fredrickson

    2001-04-22

    Seventy million cubic meters of ground and three trillion liters of groundwater have been contaminated by leaking radioactive waste generated in the United States during the Cold War. A cleanup technology is being developed based on the extremely radiation resistant bacterium Deinococcus radiodurans that is being engineered to express bioremediating functions. Research aimed at developing D. radiodurans for organic toxin degradation in highly radioactive waste sites containing radionuclides, heavy metals, and toxic organic compounds was started by this group.Work funded by the existing grant has already contributed to eleven papers on the fundamental biology of D. radiodurans and its design for bioremediation of highly radioactive waste environments

  3. Head-end process for the reprocessing of HTGR spent fuel

    SciTech Connect

    Chen, J.; Wen, M.

    2013-07-01

    The reprocessing of HTGR spent fuels is in favor of the sustainable development of nuclear energy to realize the maximal use of nuclear resource and the minimum disposal of nuclear waste. The head-end of HTGR spent fuels reprocessing is different from that of the LWR spent fuels reprocessing because of the difference of spent fuel structure. The dismantling of the graphite spent fuel element and the highly effective dissolution of fuel kernel is the most difficult process in the head end of the reprocessing. Recently, some work on the head-end has been done in China. First, the electrochemical method with nitrate salt as electrolyte was studied to disintegrate the graphite matrix from HTGR fuel elements and release the coated fuel particles, to provide an option for the head-end technology of reprocessing. The results show that the graphite matrix can be effectively separated from the coated particle without any damage to the SiC layer. Secondly, the microwave-assisted heating was applied to dissolve the UO{sub 2} kernel from the crashed coated fuel particles. The ceramic UO{sub 2} as the solute has a good ability to absorb the microwave energy. The results of UO{sub 2} kernel dissolution from crushed coated particles by microwave heating show that the total dissolution percentage of UO{sub 2} is more than 99.99% after 3 times cross-flow dissolution with the following parameters: 8 mol/L HNO{sub 3}, temperature 100 Celsius degrees, initial ratio of solid to liquid 1.2 g/ml. (authors)

  4. Influence of uncertainties of isotopic composition of the reprocessed uranium on effectiveness of its enrichment in gas centrifuge cascades

    NASA Astrophysics Data System (ADS)

    Smirnov, A. Yu; Mustafin, A. R.; Nevinitsa, V. A.; Sulaberidze, G. A.; Dudnikov, A. A.; Gusev, V. E.

    2017-01-01

    The effect of the uncertainties of the isotopic composition of the reprocessed uranium on its enrichment process in gas centrifuge cascades while diluting it by adding low-enriched uranium (LEU) and waste uranium. It is shown that changing the content of 232U and 236U isotopes in the initial reprocessed uranium within 15% (rel.) can significantly change natural uranium consumption and separative work (up to 2-3%). However, even in case of increase of these parameters is possible to find the ratio of diluents, where the cascade with three feed flows (depleted uranium, LEU and reprocessed uranium) will be more effective than ordinary separation cascade with one feed point for producing LEU from natural uranium.

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

    SciTech Connect

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

    1981-12-01

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

  6. ANL Technical Support Program for DOE Environmental Restoration and Waste Management; Annual report, October 1992--September 1993

    SciTech Connect

    Bates, J.K.; Bourcier, W.L.; Bradley, C.R.

    1994-06-01

    This report is an overview of the progress during FY 1993 for the Technical Support Program that is part of the ANL Technology Support Activity for DOE Environmental Restoration and Waste Management (EM). The purpose is to evaluate, before hot start-up of the Defense Waste Processing Facility (DWPF) and the West Valley Demonstration Project (WVDP), factors that are anticipated to affect glass reaction in an unsaturated environment typical of what may be expected for the candidate Yucca Mountain repository site. Specific goals for the testing program include the following: reviewing and evaluating available data on parameters that will be important in establishing the long-term performance of glass in a repository environment; performing tests to further quantify the effects of important variables where there are deficiencies in the available data; and initiating long-term tests to determine glass performance under a range of conditions applicable to repository disposal.

  7. ANL Technical Support Program for DOE Environmental Restoration and Waste Management. Annual report, October 1990--September 1991

    SciTech Connect

    Bates, J.K.; Bradley, C.R.; Buck, E.C.; Cunnane, J.C.; Dietz, N.L.; Ebert, W.L.; Emery, J.W.; Feng, X.; Gerding, T.J.; Gong, M.; Hoh, J.C.; Mazer, J.J.; Wronkiewicz, D.J.; Bourcier, W.L.; Morgan, L.E.; Nielsen, J.K.; Steward, S.A.; Ewing, R.C.; Wang, L.M.; Han, W.T.; Tomozawa, M.

    1992-03-01

    This report provides an overview of progress during FY 1991 for the Technical Support Program that is part of the ANL Technology Support Activity for DOE, Environmental Restoration and Waste Management (EM). The purpose is to evaluate, before hot start-up of the Defenses Waste Processing Facility (DWPF) and the West Valley Demonstration Project (WVDP), factors that are likely to affect glass reaction in an unsaturated environment typical of what may be expected for the candidate Yucca Mountain repository site. Specific goals for the testing program include the following: (1) to review and evaluate available information on parameters that will be important in establishing the long-term performance of glass in a repository environment; (2) to perform testing to further quantify the effects of important variables where there are deficiencies in the available data; and (3) to initiate long-term testing that will bound glass performance under a range of conditions applicable to repository disposal.

  8. Evaluation and compilation of DOE waste package test data; Biannual report, February 1989--July 1989: Volume 7

    SciTech Connect

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

    1991-12-01

    This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six-month period, February through July 1989. This includes reviews of related materials research and plans, information on the Yucca Mountain, Nevada disposal site activities, and other information regarding supporting research and special assistance. Outlines for planned interpretative reports on the topics of aqueous corrosion of copper, mechanisms of stress corrosion cracking and internal failure modes of Zircaloy cladding are included. For the publications reviewed during this reporting period, short discussions are given to supplement the completed reviews and evaluations. Included in this report is an overall review of a 1984 report on glass leaching mechanisms, as well as reviews for each of the seven chapters of this report.

  9. Summary of expenditures of rebates from the DOE low-level radioactive waste surcharge escrow account for calendar year 1986

    SciTech Connect

    Not Available

    1987-06-01

    The Low-Level Radioactive Waste Policy Amendments Act of 1985, Public Law 99-240, requires the Department of Energy (DOE) to manage an escrow account creatd by collection of 25% of the non-penalty surcharge fees paid by the generators in non-sited regions and nonmember states to sited states for disposal of low-level radioactive waste. For the milestone period ending June 30, 1986, a total of $921,807.84, representing surcharge fees collected and interest earned, was in escrow during 1986 for rebate to the nonmember states, non-sited compact regions, and sited states. As of December 31, 1986, $802,194.54 had been rebated from the Escrow Account with an additional $118,517.62 scheduled for rebate in early 1987. The remaining rebate to be disbursed under this milestone is $1,095.68 for the state of Delaware. At the request of the state of Delaware, this rebate amount is being held in the Escrow Account until the state provides specific instructions for its disbursement. Individual rebate expenditure reports were submitted to DOE by all the non-sited compact regions and nonmember states that received rebates in 1986. Only $14.00 of these rebates were expended in 1986. DOE reviewed all of these reports and concluded that the single expenditure complies with the expenditure limitations stated in the Act.

  10. Understanding radioactive waste

    SciTech Connect

    Murray, R.L.

    1981-12-01

    This document contains information on all aspects of radioactive wastes. Facts are presented about radioactive wastes simply, clearly and in an unbiased manner which makes the information readily accessible to the interested public. The contents are as follows: questions and concerns about wastes; atoms and chemistry; radioactivity; kinds of radiation; biological effects of radiation; radiation standards and protection; fission and fission products; the Manhattan Project; defense and development; uses of isotopes and radiation; classification of wastes; spent fuels from nuclear reactors; storage of spent fuel; reprocessing, recycling, and resources; uranium mill tailings; low-level wastes; transportation; methods of handling high-level nuclear wastes; project salt vault; multiple barrier approach; research on waste isolation; legal requiremnts; the national waste management program; societal aspects of radioactive wastes; perspectives; glossary; appendix A (scientific American articles); appendix B (reference material on wastes). (ATT)

  11. Impact of Rheological Modifiers on Various Slurries Supporting DOE Waste Processing

    SciTech Connect

    Chun, Jaehun; Bredt, Paul R.; Hansen, Erich; Bhosale, Prasad S.; Berg, John C.

    2010-03-11

    Controlling the stability and subsequent rheological properties of slurries has been an important but challenging issue in nuclear waste treatment, one that previous research has yet to sufficiently address. At the Hanford and Savannah River sites, operation of the waste treatment facilities at increased solids loading reduces the evaporative load on the melter systems and thereby increases waste processing rates. However, at these higher solids loadings, increased slurry rheology becomes a significant processing issue. The current study evaluates the use of several rheological modifiers to alleviate increased slurry rheology at high waste solids concentrations. Rheological modifiers change particle interactions in slurry. For colloidal slurries, modifiers mainly alter the electrostatic and steric interactions between particles, leading to a change in slurry rheology. Weak organic acid type rheological modifiers strengthen electrostatic repulsion whereas nonionic/polymer surfactant type rheological modifiers introduce a steric repulsion. We investigated various rheological modifiers using high level waste (HLW) nuclear waste simulants characterized typically by high ionic strength and a wide range of pH from 4 to 13. Using rheological analysis, it was found that citric acid and polyacrylic acid would be good rheological modifiers for the HLW simulants tested, effectively reducing slurry rheology by 40% or more. Physical insights into the mechanisms driving stabilization by these rheological modifiers will be discussed.

  12. Does performance evaluation help public managers? A Balanced Scorecard approach in urban waste services.

    PubMed

    Guimarães, Bernardo; Simões, Pedro; Marques, Rui Cunha

    2010-12-01

    The urban waste market has evolved significantly in the past decades, which among other changes, has led to the creation of new utilities and new business models. However, very few things have changed for the users. Urban waste collection remains mainly under the responsibility of local authorities and the charges paid by the users in most countries are very low compared to the provision costs. This situation forces the injection of public money into the system, encouraging the 'quiet-life' within the utilities and, therefore, inefficiency. The present study intends to analyze the potential for the application of the Balanced Scorecard (BSc) methodology into the waste utilities. After a comprehensive revision of the urban waste sector in Portugal, the methodology of BSc and its application in local public services is described and discussed. Focusing on implementation rather than on strategy, a set of performance indicators is proposed to be utilized in the different management models of waste utilities in Portugal: the municipalities, semi-autonomous utilities, municipal companies and mixed companies. This implementation is then exemplified through four case studies, one for each type of utility. This paper provides a flexible framework proposal to be applied to waste utilities operating both in Portugal and abroad. Copyright © 2010 Elsevier Ltd. All rights reserved.

  13. Effect of Reprocessing and Accelerated Weathering on Impact-Modified Recycled Blend

    NASA Astrophysics Data System (ADS)

    Ramesh, V.; Mohanty, Smita; Biswal, Manoranjan; Nayak, Sanjay K.

    2015-12-01

    Recovery of recycled polycarbonate, acrylonitrile butadiene styrene, high-impact polystyrene, and its blends from waste electrical and electronic equipment plastics products properties were enhanced by the addition of virgin polycarbonate and impact modifier. The optimized blend formulation was processed through five cycles, at processing temperature, 220-240 °C and accelerated weathering up to 700 h. Moreover, the effect of reprocessing and accelerated weathering in the physical properties of the modified blends was investigated by mechanical, thermal, rheological, and morphological studies. The results show that in each reprocessing cycle, the tensile strength and impact strength decreased significantly and the similar behavior has been observed from accelerated weathering. Subsequently, the viscosity decreases and this decrease becomes the effect of thermal and photo-oxidative degradation. This can be correlated with FTIR analysis.

  14. The consequences of disposal of low-level radioactive waste from the Fernald Environmental Management Project: Report of the DOE/Nevada Independent Panel

    SciTech Connect

    Crowe, B.; Hansen, W.; Waters, R.; Sully, M.; Levitt, D.

    1998-04-01

    The Department of Energy (DOE) convened a panel of independent scientists to assess the performance impact of shallow burial of low-level radioactive waste from the Fernald Environmental Management Project, in light of a transportation incident in December 1997 involving this waste stream. The Fernald waste has been transported to the Nevada Test Site and disposed in the Area 5 Radioactive Waste Management Site (RWMS) since 1993. A separate DOE investigation of the incident established that the waste has been buried in stress-fractured metal boxes, and some of the waste contained excess moisture (high-volumetric water contents). The Independent Panel was charged with determining whether disposition of this waste in the Area 5 RWMS has impacted the conclusions of a previously completed performance assessment in which the site was judged to meet required performance objectives. To assess the performance impact on Area 5, the panel members developed a series of questions. The three areas addressed in these questions were (1) reduced container integrity, (2) the impact of reduced container integrity on subsidence of waste in the disposal pits and (3) excess moisture in the waste. The panel has concluded that there is no performance impact from reduced container integrity--no performance is allocated to the container in the conservative assumptions used in performance assessment. Similarly, the process controlling post-closure subsidence results primarily from void space within and between containers, and the container is assumed to degrade and collapse within 100 years.

  15. Guidance document for revision of DOE Order 5820.2A, Radioactive Waste Technical Support Program. Revision 1

    SciTech Connect

    Kudera, D.E.; McMurtrey, C.D.; Meagher, B.G.

    1993-04-01

    This document provides guidance for the revision of DOE Order 5820.2A, ``Radioactive Waste Management.`` Technical Working Groups have been established and are responsible for writing the revised order. The Technical Working Groups will use this document as a reference for polices and procedures that have been established for the revision process. The overall intent of this guidance is to outline how the order will be revised and how the revision process will be managed. In addition, this document outlines technical issues considered for inclusion by a Department of Energy Steering Committee.

  16. Development of advanced electrochemical emission spectroscopy for monitoring corrosion in simulated DOE liquid waste. 1998 annual progress report

    SciTech Connect

    MacDonald, D.D.

    1998-06-01

    'Objective of this project is to develop and use Electrochemical Emission Spectroscopy (EES) and other electrochemical techniques as in situ tools for exploring corrosion mechanisms of iron and carbon steel in highly alkaline solutions and for continuously monitoring corrosion on structural materials in DOE liquid waste storage system. In particular, the author will explore the fundamental aspects of the passive behavior of pure iron since breakdown of passivity leads to localized corrosion. This report summarizes work after 1 year of a 3 year project.'

  17. HWMA closure plan for the Waste Calcining Facility at the Idaho National Engineering Laboratory

    SciTech Connect

    1996-05-01

    The Waste Calcining Facility (WCF) calcined and evaporated aqueous wastes generated from the reprocessing of spent nuclear fuel. The calciner operated from 1963 to 1981, primarily processing high level waste from the first cycle of spent fuel extraction. Following the calciner shutdown the evaporator system concentrated high activity aqueous waste from 1983 until 1987. In 1988, US Department of Energy Idaho Operations Office (DOE-ID) requested interim status for the evaporator system, in anticipation of future use of the evaporator system. The evaporator system has not been operated since it received interim status. At the present time, DOE-ID is completing construction on a new evaporator at the New Waste Calcining Facility (NWCF) and the evaporator at the WCF is not needed. The decision to not use the WCF evaporator requires Lockheed Idaho Technologies Company (LITCO) and DOE-ID to close these units. After a detailed evaluation of closure options, LITCO and DOE-ID have determined the safest option is to fill the voids (grout the vessels, cells and waste pile) and close the WCF to meet the requirements applicable to landfills. The WCF will be covered with a concrete cap that will meet the closure standards. In addition, it was decided to apply these closure standards to the calcining system since it is contained within the WCF building. The paper describes the site, waste inventory, closure activities, and post-closure care plans.

  18. Structural damage and chemical contaminants on reprocessed arthroscopic shaver blades.

    PubMed

    Kobayashi, Masahiko; Nakagawa, Yasuaki; Okamoto, Yukihiro; Nakamura, Shinichiro; Nakamura, Takashi

    2009-02-01

    In response to socioeconomic pressure to cut budgets in medicine, single-use surgical instruments are often reprocessed despite potential biological hazard. To evaluate the quality and contaminants of reprocessed shaver blades. Reprocessed shaver blades have mechanical damage and chemical contamination. Controlled laboratory study. Seven blades and 3 abraders were reprocessed 1 time or 3 times and then were assessed. In the first part of the study, structural damage on the blades after 3 reprocessings was compared to that after 1 reprocessing using optical microscopy. In the second part, surface damage was observed using optical microscopy and scanning electron microscopy; elemental and chemical analyses of contaminants found by the microscopy were performed using scanning electron microscopy/energy dispersive x-ray spectroscopy, scanning Auger microscopy, and Fourier transform infrared spectroscopy. Optical microscopic examination revealed abrasion on the surface of the inner blade and cracks on the inner tube after 1 reprocessing. These changes were more evident after 3 reprocessings. Scanning electron microscopy/energy dispersive x-ray spectroscopy of the inner cutter of the blade reprocessed once showed contaminants containing calcium, carbon, oxygen, and silicon, and Fourier transform infrared spectroscopy demonstrated biological protein consisting mainly of collagen, some type of salts, and polycarbonate used in plastic molding. Scanning electron microscopy/energy dispersive x-ray spectroscopy of the inner cutter of the reprocessed abrader revealed contaminants containing carbon, calcium, phosphorous, and oxygen, and Fourier transform infrared spectroscopy showed H2O, hydroxyapatite, and hydroxyl proteins. Scanning Auger microscopy showed that the tin-nickel plating on the moving blade and abrader was missing in some locations. This is the first study to evaluate both mechanical damage and chemical contaminants containing collagen, hydroxyapatite, and salts

  19. Korean Society of Gastrointestinal Endoscopy Guidelines for Endoscope Reprocessing

    PubMed Central

    Son, Byoung Kwan; Kim, Byung-Wook; Kim, Won Hee; Myung, Dae-Sung; Cho, Young-Seok; Jang, Byung Ik

    2017-01-01

    The Korean Society of Gastrointestinal Endoscopy (KSGE) issued guidelines for endoscope reprocessing for the first time in 1995, and the version of the guidelines was updated in August 2009, August 2012, and March 2015. Guidelines for endoscope reprocessing should be revised continuously, because new disinfectants and devices are developed and introduced. The current official version of the KSGE guidelines for endoscope reprocessing is explained herein to assist the reader in understanding the KSGE requirements for cleaning and disinfecting endoscopes. PMID:28301923

  20. Final disposal of radioactive waste

    NASA Astrophysics Data System (ADS)

    Freiesleben, H.

    2013-06-01

    In this paper the origin and properties of radioactive waste as well as its classification scheme (low-level waste - LLW, intermediate-level waste - ILW, high-level waste - HLW) are presented. The various options for conditioning of waste of different levels of radioactivity are reviewed. The composition, radiotoxicity and reprocessing of spent fuel and their effect on storage and options for final disposal are discussed. The current situation of final waste disposal in a selected number of countries is mentioned. Also, the role of the International Atomic Energy Agency with regard to the development and monitoring of international safety standards for both spent nuclear fuel and radioactive waste management is described.

  1. Analysis of nuclear proliferation resistance reprocessing and recycling technologies

    SciTech Connect

    Patricia Paviet-Hartmann; Gary Cerefice; Marcela Stacey; Steven Bakhtiar

    2011-05-01

    The PUREX process has been progressively and continuously improved during the past three decades, and these improvements account for successful commercialization of reprocessing in a few countries. The renewed interest in nuclear energy and the international growth of nuclear electricity generation do not equate – and should not be equated -with increasing proliferation risks. Indeed, the nuclear renaissance presents a unique opportunity to enhance the culture of non-proliferation. With the recent revival of interest in nuclear technology, technical methods for prevention of nuclear proliferation are being revisited. Robust strategies to develop new advanced separation technologies are emerging worldwide for sustainability and advancement of nuclear energy with enhanced proliferation resistance. On the other hand, at this moment, there are no proliferation resistance advanced technologies. . Until now proliferation resistance as it applies to reprocessing has been focused on not separating a pure stream of weapons-usable plutonium. France, as an example, has proposed a variant of the PUREX process, the COEX TM process, which does not result on a pure plutonium product stream. A further step is to implement a process based on group extraction of actinides and fission products associated with a homogeneous recycling strategy (UNEX process in the US, GANEX process in France). Such scheme will most likely not be deployable on an industrial scale before 2030 or so because it requires intensive R&D and robust flowsheets. Finally, future generation recycling schemes will handle the used nuclear fuel in fast neutron reactors. This means that the plutonium throughput of the recycling process may increase. The need is obvious for advanced aqueous recycling technologies that are intrinsically more proliferation resistant than the commercial PUREX process. In this paper, we review the actual PUREX process along with the advanced recycling technologies that will enhance

  2. Research on jet mixing of settled sludges in nuclear waste tanks at Hanford and other DOE sites: A historical perspective

    SciTech Connect

    Powell, M.R.; Onishi, Y.; Shekarriz, R.

    1997-09-01

    Jet mixer pumps will be used in the Hanford Site double-shell tanks to mobilize and mix the settled solids layer (sludge) with the tank supernatant liquid. Predicting the performance of the jet mixer pumps has been the subject of analysis and testing at Hanford and other U.S. Department of Energy (DOE) waste sites. One important aspect of mixer pump performance is sludge mobilization. The research that correlates mixer pump design and operation with the extent of sludge mobilization is the subject of this report. Sludge mobilization tests have been conducted in tanks ranging from 1/25-scale (3 ft-diameter) to full scale have been conducted at Hanford and other DOE sites over the past 20 years. These tests are described in Sections 3.0 and 4.0 of this report. The computational modeling of sludge mobilization and mixing that has been performed at Hanford is discussed in Section 5.0.

  3. Closure of hazardous and mixed radioactive waste management units at DOE facilities. [Contains glossary

    SciTech Connect

    Not Available

    1990-06-01

    This is document addresses the Federal regulations governing the closure of hazardous and mixed waste units subject to Resource Conservation and Recovery Act (RCRA) requirements. It provides a brief overview of the RCRA permitting program and the extensive RCRA facility design and operating standards. It provides detailed guidance on the procedural requirements for closure and post-closure care of hazardous and mixed waste management units, including guidance on the preparation of closure and post-closure plans that must be submitted with facility permit applications. This document also provides guidance on technical activities that must be conducted both during and after closure of each of the following hazardous waste management units regulated under RCRA.

  4. Removal of actinides from nuclear fuel reprocessing wastes: engineering studies

    SciTech Connect

    Maxey, H. R.; McIsaac, L. D.; Chamberlain, D. B.; McManus, G. J.

    1980-01-01

    The major pilot plant effort has been directed towards understanding the behavior of the extraction column. This emphasis was placed on extraction because the actinide-free raffinate from this column is the process product. Although actual radioactive feeds have been used only in a very limited number of tests, the combination of simulated column tests, minimixer-settler tests, and pilot plant experiments have shown that an actinide-free raffinate can be produced. A pulsed sieve plate column can achieve this separation in a reasonable height. In the design of an extraction process, there are several tradeoffs that have to be examined. One such tradeoff is that between column capacity and separation efficiency; an increase in column diameter (or capacity) means a faster processing rate, but, unfortunately, it also means a decrease in column efficiency. Therefore, the greater the diameter, the larger the HTU and the higher the column has to be to achieve the desired separation. In nuclear applications this is very important since column height is limited by the size of the process containment building. In increasing the diameter from 3 to 8 inches, an increase in HTU's up to 50% was observed by Sege. For plant operations, the pulse columns could probably be pulsed at 40 CPM, 2.5 cm per pulse. At these conditions, the HTU should be no greater than 0.75 m, even considering changes resulting from scaleup.

  5. DOE Waste Package Project. Quarterly progress report, October 1, 1993--December 31, 1993

    SciTech Connect

    Ladkany, S.G.

    1993-12-31

    The Waste Package Project research activities continued in all research areas. The areas include: Container structural and stress analysis; Nuclear fission criticality studies; Investigation of canister design concepts and corrosion studies; Heat transfer studies; Fluid flow in porous media and radionuclide transport in near field rock; Studies of stresses and stability of the rock formations resulting from the thermal loading of the fuel elements and the multi tunnel concept being analyzed; Characterization of a Faulted Rock Tunnel Model Using Photoelastic and Finite Element Studies; Experiment studies of the dynamic response of a flexible three-link robot using strain gages and Lagrange polynomials; and Robotic manipulation of the nuclear waste container.

  6. Research Program to Determine Redox Reactions and Their Effects on Speciation and Mobility of Plutonium in DOE Wastes

    SciTech Connect

    Choppin, G.R.; Rai, D.

    2000-10-01

    Plutonium in geologic matrices undergoes a variety of complex reactions which complicate its environmental behavior. These complexities in plutonium chemistry whereby a large variety of precipitation, dissolution, adsorption/desorption, and redox reactions control plutonium speciation and concentrations, result in the need for a rather large amount of reliable, fundamental data to predict Pu behavior in geologic media. These data are also needed for evaluation of remediation strategies that involve removing most of the contaminants by selective methods, followed by in situ immobilization of residual contaminants. Two areas were studied during this project: (1) thermodynamic data for Th(IV) and Pu(IV) complexes of EDTA and for Pu(V) interactions with chloride; (2) kinetic data for redox reactions of Pu in the presence of common redox agents (e.g., H{sub 2}O{sub 2}, MnO{sub 2}, and NaOCl) encountered under waste disposal conditions. These studies are relevant to understanding Pu behavior in wastes disposed of in diverse geologic conditions (e.g., at the WIPP and YUCCA Mountain repositories and in contaminated sediments at many different DOE sites) and also for developing effective remediation strategies (e.g., processing of high level waste tanks). These studies have yielded data to address redox reactions of plutonium in the presence of environmentally important agents (e.g. organic and inorganic oxidants/reductants).

  7. Waste generation forecast for DOE-ORO`s Environmental Restoration OR-1 Project: FY 1995-FY 2002, September 1994 revision

    SciTech Connect

    Not Available

    1994-12-01

    A comprehensive waste-forecasting task was initiated in FY 1991 to provide a consistent, documented estimate of the volumes of waste expected to be generated as a result of U.S. Department of Energy-Oak Ridge Operations (DOE-ORO) Environmental Restoration (ER) OR-1 Project activities. Continual changes in the scope and schedules for remedial action (RA) and decontamination and decommissioning (D&D) activities have required that an integrated data base system be developed that can be easily revised to keep pace with changes and provide appropriate tabular and graphical output. The output can then be analyzed and used to drive planning assumptions for treatment, storage, and disposal (TSD) facilities. The results of this forecasting effort and a description of the data base developed to support it are provided herein. The initial waste-generation forecast results were compiled in November 1991. Since the initial forecast report, the forecast data have been revised annually. This report reflects revisions as of September 1994.

  8. Data validation and security for reprocessing.

    SciTech Connect

    Tolk, Keith Michael; Merkle, Peter Benedict; DurÔan, Felicia Angelica; Cipiti, Benjamin B.

    2008-10-01

    Next generation nuclear fuel cycle facilities will face strict requirements on security and safeguards of nuclear material. These requirements can result in expensive facilities. The purpose of this project was to investigate how to incorporate safeguards and security into one plant monitoring system early in the design process to take better advantage of all plant process data, to improve confidence in the operation of the plant, and to optimize costs. An existing reprocessing plant materials accountancy model was examined for use in evaluating integration of safeguards (both domestic and international) and security. International safeguards require independent, secure, and authenticated measurements for materials accountability--it may be best to design stand-alone systems in addition to domestic safeguards instrumentation to minimize impact on operations. In some cases, joint-use equipment may be appropriate. Existing domestic materials accountancy instrumentation can be used in conjunction with other monitoring equipment for plant security as well as through the use of material assurance indicators, a new metric for material control that is under development. Future efforts will take the results of this work to demonstrate integration on the reprocessing plant model.

  9. Report: EPA Does Not Effectively Control or Monitor Imports of Hazardous Waste

    EPA Pesticide Factsheets

    Report #15-P-0172, July 6, 2015. The EPA lacks explicit authority to block imported shipments of hazardous waste that lack prior EPA consent. This could lead to improper handling and disposal, resulting in unknown human and environmental exposure to toxic

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

    SciTech Connect

    Not Available

    1983-12-01

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

  11. Design and construction of deinococcus radiodurans for biodegradation of organic toxins at radioactive DOE waste sites. 1998 annual progress report

    SciTech Connect

    Daly, M.J.; Wackett, L.P.; Minton, K.W.

    1998-06-01

    'A 1992 survey of DOE waste sites indicates that about 32% of soils and 45% of groundwaters at these sites contain radionuclides and metals plus an organic toxin class. The most commonly reported combinations of these hazardous compounds being radionuclides and metals (e.g., U, Pu, Cs, Pb, Cr, As) plus chlorinated hydrocarbons (e.g., trichloroethylene), fuel hydrocarbons (e.g., toluene), or polychlorinated biphenyls (e.g., Arochlor 1248). These wastes are some of the most hazardous pollutants and pose an increasing risk to human health as they leach into the environment. The objective of this research is to develop novel organisms, that are highly resistant to radiation and the toxic effects of metals and radionuclides, for in-situ bioremediation of organic toxins. Few organisms exist that are able to remediate such environmental organic pollutants, and among those that can, the bacteria belonging to the genus Pseudomonas are the most characterized. Unfortunately, these bacteria are very radiation sensitive. For example, Pseudomonas spp. is even more sensitive than Escherichia coli and, thus, is not suitable as a bioremediation host in environments subjected to radiation. By contrast, D. radiodurans, a natural soil bacterium, is the most radiation resistant organism yet discovered; it is several thousand times more resistant to ionizing radiation than Pseudomonas. The sophisticated gene transfer and expression systems the authors have developed for D. radiodurans over the last eight years make this organism an ideal candidate for high-level expression of genes that degrade organic toxins, in radioactive environments. The authors ultimate aim is to develop organisms and approaches that will be useful for remediating the large variety of toxic organic compounds found in DOE waste sites that are too radioactive to support other bioremediation organisms. This report summarizes work after the first 6 months of a 3-year project.'

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

    SciTech Connect

    Wilhite, E

    2008-03-31

    This Performance Assessment for the Savannah River Site E-Area Low-Level Waste Facility was prepared to meet requirements of Chapter IV of the Department of Energy Order 435.1-1. The Order specifies that a Performance Assessment should provide reasonable assurance that a low-level waste disposal facility will comply with the performance objectives of the Order. The Order also requires assessments of impacts to water resources and to hypothetical inadvertent intruders for purposes of establishing limits on radionuclides that may be disposed near-surface. According to the Order, calculations of potential doses and releases from the facility should address a 1,000-year period after facility closure. The point of compliance for the performance measures relevant to the all pathways and air pathway performance objective, as well as to the impact on water resources assessment requirement, must correspond to the point of highest projected dose or concentration beyond a 100-m buffer zone surrounding the disposed waste following the assumed end of active institutional controls 100 years after facility closure. During the operational and institutional control periods, the point of compliance for the all pathways and air pathway performance measures is the SRS boundary. However, for the water resources impact assessment, the point of compliance remains the point of highest projected dose or concentration beyond a 100-m buffer zone surrounding the disposed waste during the operational and institutional control periods. For performance measures relevant to radon and inadvertent intruders, the points of compliance are the disposal facility surface for all time periods and the disposal facility after the assumed loss of active institutional controls 100 years after facility closure, respectively. The E-Area Low-Level Waste Facility is located in the central region of the SRS known as the General Separations Area. It is an elbow-shaped, cleared area, which curves to the northwest

  13. Status of radioiodine control for nuclear fuel reprocessing plants

    SciTech Connect

    Burger, L.L.; Scheele, R.D.

    1983-07-01

    This report summarizes the status of radioiodine control in a nuclear fuel reprocessing plant with respect to capture, fixation, and disposal. Where possible, we refer the reader to a number of survey documents which have been published in the last four years. We provide updates where necessary. Also discussed are factors which must be considered in developing criteria for iodine control. For capture from gas streams, silver mordenite and a silver nitrate impregnated silica (AC-6120) are considered state-of-the-art and are recommended. Three aqueous scrubbing processes have been demonstrated: Caustic scrubbing is simple but probably will not give an adequate iodine retention by itself. Mercurex (mercuric nitrate-nitric acid scrubbing) has a number of disadvantages including the use of toxic mercury. Iodox (hyperazeotropic nitric acid scrubbing) is effective but employs a very corrosive and hazardous material. Other technologies have been tested but require extensive development. The waste forms recommended for long-term storage or disposal are silver iodide, the iodates of barium, strontium, or calcium, and silver loaded sorbents, all fixed in cement. Copper iodide in bitumen (asphalt) is a possibility but requires testing. The selection of a specific form will be influenced by the capture process used.

  14. 9 CFR 114.18 - Reprocessing of biological products.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 9 Animals and Animal Products 1 2011-01-01 2011-01-01 false Reprocessing of biological products. 114.18 Section 114.18 Animals and Animal Products ANIMAL AND PLANT HEALTH INSPECTION SERVICE... REQUIREMENTS FOR BIOLOGICAL PRODUCTS § 114.18 Reprocessing of biological products. The Administrator may...

  15. 9 CFR 114.18 - Reprocessing of biological products.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Reprocessing of biological products. 114.18 Section 114.18 Animals and Animal Products ANIMAL AND PLANT HEALTH INSPECTION SERVICE... REQUIREMENTS FOR BIOLOGICAL PRODUCTS § 114.18 Reprocessing of biological products. The Administrator may...

  16. 76 FR 24495 - Reprocessing of Reusable Medical Devices; Public Workshop

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-05-02

    ... HUMAN SERVICES Food and Drug Administration Reprocessing of Reusable Medical Devices; Public Workshop... Medical Devices Workshop.'' The purpose of the workshop is to discuss factors affecting the reprocessing of reusable medical devices and FDA's plans to address the identified issues. This workshop is part...

  17. Method and article for primary containment of cesium wastes. [DOE patent application

    DOEpatents

    Angelini, P.; Lackey, W.J.; Stinton, D.P.; Blanco, R.E.; Bond, W.D.; Arnold, W.D. Jr.

    1981-09-03

    A method for producing a cesium-retentive waste form, characterized by a high degree of compositional stability and mechanical integrity, is provided by subjecting a cesium-loaded zeolite to heat under conditions suitable for stabilizing the zeolite and immobilizing the cesium, and coating said zeolite for sufficient duration within a suitable environment with at least one dense layer of pyrolytic carbon to seal therein said cesium to produce a final, cesium-bearing waste form. Typically, the zolite is stabilized and the cesium immobilized in less than four hours by confinement within an air environment maintained at about 600/sup 0/C. Coatings are thereafter applied by confining the calcined zeolite within a coating environment comprising inert fluidizing and carbon donor gases maintained at 1000/sup 0/C for a suitable duration.

  18. TECHNOLOGY NEEDS AND STATUS ON CLOSURE OF DOE RADIOACTIVE WASTE TANK ANCILLARY SYSTEMS

    SciTech Connect

    Burns, H; Sharon Marra, S; Christine Langton, C

    2009-01-21

    This paper summarizes the current state of art of sampling, characterizing, retrieving, transferring and treating the incidental waste and stabilizing the void space in tank ancillary systems and the needs involved with closure of these systems. The overall effort for closing tank and ancillary systems is very large and is in the initial stages of being addressed in a systematic manner. It was recognized in doing this effort, that gaps in both technology and material application for characterization and removal of residual waste and closure of ancillary systems would be identified. Great efficiencies are to be gained by defining the technology need areas early in the closure process and providing recommendations for technical programs to improve the closure strategies. Therefore, this paper will not only summarize the state of closure of ancillary systems but also provide recommendations to address the technology gaps identified in this assessment.

  19. Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

    DOE PAGES

    Youinou, Gilles J.

    2017-05-04

    This article presents the results of a neutronics analysis related to the homogeneous recycling of transuranics (TRU) in PWRs with a MOX fuel using enriched uranium instead of depleted uranium. It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely the use of reprocessed uranium. From a neutronics point of view, it is possible to multi-recycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one third and two thirds of the fleet. Recycling neptunium and americium with plutonium significantlymore » decreases the decay heat of the waste stream between 100 to 1,000 years compared to those of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235uranium and recycling it makes a major difference on the natural uranium needs. For example, a PWR fleet recycling its plutonium, neptunium and americium in MOXEU needs 28 percent more natural uranium than a reference UO2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19 percent less if the reprocessed uranium is recycled back in the reactors, i.e. a 47 percent difference.« less

  20. A West Valley Demonstration Project Milestone - Achieving Certification to Ship Waste to the Nevada Test Site

    SciTech Connect

    Jackson, J. P.; Pastor, R. S.

    2002-02-28

    The West Valley Demonstration Project (WVDP) has successfully pretreated and vitrified nearly all of the 600,000 gallons of liquid high-level radioactive waste that was generated at the site of the only commercial nuclear fuel reprocessing plant to have operated in the United States. Low-level waste (LLW) generated during the course of the cleanup effort now requires disposal. Currently the WVDP only ships Class A LLW for off-site disposal. It has been shipping Class A wastes to Envirocare of Utah, Inc. since 1997. However, the WVDP may also have a future need to ship Class B and Class C waste, which Envirocare is not currently authorized to accept. The Nevada Test Site (NTS), a U.S. Department of Energy (DOE) facility, can accept all three waste classifications. The WVDP set a goal to receive certification to begin shipping Class A wastes to NTS by 2001. Formal certification/approval was granted by the DOE Nevada Operations Office on July 12, 2001. This paper discusses how the WVDP contractor, West Valley Nuclear Services Company (WVNSCO), completed the activities required to achieve NTS certification in 2001 to ship waste to its facility. The information and lessons learned provided are significant because the WVDP is the only new generator receiving certification based on an NTS audit in January 2001 that resulted in no findings and only two observations--a rating that is unparalleled in the DOE Complex.

  1. Waste Generation Forecast for DOE-ORO`s Environmental Restoration OR-1 Project: FY 1994--FY 2001. Environmental Restoration Program, September 1993 Revision

    SciTech Connect

    Not Available

    1993-12-01

    This Waste Generation Forecast for DOE-ORO`s Environmental Restoration OR-1 Project. FY 1994--FY 2001 is the third in a series of documents that report current estimates of the waste volumes expected to be generated as a result of Environmental Restoration activities at Department of Energy, Oak Ridge Operations Office (DOE-ORO), sites. Considered in the scope of this document are volumes of waste expected to be generated as a result of remedial action and decontamination and decommissioning activities taking place at these sites. Sites contributing to the total estimates make up the DOE-ORO Environmental Restoration OR-1 Project: the Oak Ridge K-25 Site, the Oak Ridge National Laboratory, the Y-12 Plant, the Paducah Gaseous Diffusion Plant, the Portsmouth Gaseous Diffusion Plant, and the off-site contaminated areas adjacent to the Oak Ridge facilities (collectively referred to as the Oak Ridge Reservation Off-Site area). Estimates are available for the entire fife of all waste generating activities. This document summarizes waste estimates forecasted for the 8-year period of FY 1994-FY 2001. Updates with varying degrees of change are expected throughout the refinement of restoration strategies currently in progress at each of the sites. Waste forecast data are relatively fluid, and this document represents remediation plans only as reported through September 1993.

  2. Proceedings of the eighth annual DOE low-level waste management forum: Executive summary, opening plenary session, closing plenary session, attendees

    SciTech Connect

    Not Available

    1987-02-01

    The Eighth Annual DOE (Department of Energy) Low-Level Waste Management Forum was held in September 1986, in Denver, Colorado, to provide a forum for exchange of information on low-level radioactive waste (LLW) management activities, requirements, and plans. The one hundred ninety attendees included representatives from the DOE Nuclear Energy and Defense Low-Level Waste Management Programs, DOE Operations Offices and their contractors; representatives from the US Nuclear Regulatory Commission (NRC), US Environmental Protection Agency (EPA), US Geological Survey, and their contractors; representatives of states and regions responsible for development of new commercial low-level waste disposal facilities; representatives of utilities, private contractors, disposal facility operators, and other parties concerned with low-level waste management issues. Plenary sessions were held at the beginning and conclusion of the meeting, while eight concurrent topical sessions were held during the intervening two days. The meeting was organized by topical areas to allow for information exchange and discussion on current and future low-level radioactive waste management challenges. Session chairmen presented summaries of the discussions and conclusions resulting from their respective sessions. Selected papers in this volume have been processed for inclusion in the Energy Data Base.

  3. Accelerator-driven Transmutation of Waste

    NASA Astrophysics Data System (ADS)

    Venneri, Francesco

    1998-04-01

    Nuclear waste from commercial power plants contains large quantities of plutonium, other fissionable actinides, and long-lived fission products that are potential proliferation concerns and create challenges for the long-term storage. Different strategies for dealing with nuclear waste are being followed by various countries because of their geologic situations and their views on nuclear energy, reprocessing and non-proliferation. The current United States policy is to store unprocessed spent reactor fuel in a geologic repository. Other countries are opting for treatment of nuclear waste, including partial utilization of the fissile material contained in the spent fuel, prior to geologic storage. Long-term uncertainties are hampering the acceptability and eventual licensing of a geologic repository for nuclear spent fuel in the US, and driving up its cost. The greatest concerns are with the potential for radiation release and exposure from the spent fuel for tens of thousands of years and the possible diversion and use of the actinides contained in the waste for weapons construction. Taking advantage of the recent breakthroughs in accelerator technology and of the natural flexibility of subcritical systems, the Accelerator-driven Transmutation of Waste (ATW) concept offers the United States and other countries the possibility to greatly reduce plutonium, higher actinides and environmentally hazardous fission products from the waste stream destined for permanent storage. ATW does not eliminate the need for, but instead enhances the viability of permanent waste repositories. Far from being limited to waste destruction, the ATW concept also brings to the table new technologies that could be relevant for next-generation power producing reactors. In the ATW concept, spent fuel would be shipped to the ATW site where the plutonium, transuranics and selected long-lived fission products would be destroyed by fission or transmutation in their first and only pass through the

  4. DOE ORDER 435.1, IMPLEMENTATION AND COMPLIANCE DECLARATION AT THE SAVANNAH RIVER SITE AND ACROSS THE DOE COMPLEX IN CONTRAST TO CURRENT PUSHBACK EFFORTS FROM THE ''TOP-TO-BOTTOM'' REVIEW

    SciTech Connect

    GOLDSTON, WELFORD T.; SMITH, WINCHESTER IV

    2003-02-27

    DOE issued Order 435.1, ''Radioactive Waste Management,'' on July 9, 1999 for immediate implementation. The requirements for Low Level Mixed, Transuranic, and High Level Waste have been completely rewritten. The entire DOE complex has been struggling with how to implement these new requirements within the one year required timeframe. This paper will chronicle the implementation strategy and actual results of the work to carry out that strategy at the Savannah River Site. DOE-SR and the site contractors worked closely together to implement each of the new requirements across the SRS, crossing many barriers and providing innovative solutions to the many problems that surfaced throughout the year. The results are that SRS declared compliance with all of the requirements of the Order within the prescribed timeframe. The challenge included all waste types in SRS facilities and programs that handle LLW, MLLW, TRU, and HLW. This paper will describe the implementation details for development of Radioactive Waste Management Basis for each facility, Identification of Wastes with No Path to Disposal, Waste Incidental to Reprocessing Determinations, Low Level Waste 90-Day Staging and One Year Limits for Storage Programs, to name a few of the requirements that were addressed by the SRS 435.1 Implementation Team. This paper will trace the implementation, problems (both technical and administrative), and the current pushback efforts associated with the DOE ''Top-to-Bottom'' review.

  5. Effect of repeated tracheostomy tube reprocessing on biofilm formation.

    PubMed

    Rodney, Jennifer; Ojano-Dirain, Carolyn P; Antonelli, Patrick J; Silva, Rodrigo C

    2016-04-01

    To determine the effect of repeated reprocessing of pediatric tracheostomy tubes (TTs) on biofilm formation. In vitro microbiological study. Pediatric, uncuffed, polyvinyl chloride (PVC) TTs from two different manufacturers (Tracoe Mini and Shiley) were reprocessed mechanically with household detergent and soaked in sodium hypochlorite (bleach). Two TTs of each brand were reprocessed 0 (control), 10, or 20 times. Twenty 2-mm coupons were then obtained from each TT, immersed in human mucus, and cultured with either Staphylococcus aureus or Pseudomonas aeruginosa. Biofilm formation was evaluated with bacterial counts. Bacterial counts of S. aureus for both brands were significantly higher on the TTs that were reprocessed 20 times compared to those that were not reprocessed (Tracoe: P = .040, Shiley: P  <  .0001) or those that were reprocessed 10 times (Tracoe: P = .022, Shiley: P = .0002). There was no difference between controls and TTs reprocessed 10 times (Tracoe: P = .76, Shiley: P = .24). P. aeuruginosa counts were not significantly different among the varying numbers of reprocessing cycles for either Tracoe or Shiley TTs (P = .08 and P = .97, respectively). Repeated reprocessing of PVC TTs with detergent and bleach paradoxically promotes S. aureus biofilm development, possibly due to degradation of the tube surface that facilitates bacterial attachment. Further investigation is needed to determine the optimal technique and limits of reprocessing TTs in clinical practice. NA. Laryngoscope published by Wiley on behalf of the American Laryngological, Rhinological and Otological Society, Inc, “The Triological Society” and American Laryngological Association (the “Owner”).

  6. DOE Waste Package Project. Quarterly progress report, April 1, 1993--June 30, 1993 and end of year summary report

    SciTech Connect

    Ladkany, S.G.

    1993-08-01

    Contents of this report are as follows: Overview and progress of waste package project and container design; waste container alternate design considerations; structural analysis and design of nuclear waste package canister; manipulation of the nuclear waste container; design requirements of various rock tunnel shapes for long term storage of high level waste; and transport phenomena in the near field.

  7. Residuals on medical devices following reprocessing.

    PubMed

    Martiny, H; Floss, H

    2001-08-01

    Micro-organisms may be transmitted by medical devices. A large variety of infectious agents may be involved in infections transmitted by endoscopic procedures. We review a series of examples that demonstrate to what extent micro-organisms can be detected on medical devices and how transmission on to subsequently examined persons due to inadequate reprocessing can occur. Hardly any data are available regarding residuals of process chemicals, although numerous published cases of glutaraldehyde-related colitis demonstrate that this issue requires urgent clarification. A risk of endoscope contamination exists, interalia, if washer-disinfectors are technically defective or are incorrectly operated. In particular, a final rinse water of poor microbiological quality can lead to recontamination of endoscopes.

  8. Methods of reprocessing complex medical equipment.

    PubMed

    Babb, J R

    1988-02-01

    The choice as to which of the two gaseous processes is best suited to individual hospital needs is a difficult one. Very few items are unable to tolerate 73 degrees C (LTSF) and these few can withstand 37 degrees C or 55 degrees C (EO). Unfortunately, LTSF is a 'moist' process and sterilizers have a poor history of providing sterilization without modification, and consequently few are used. Ethylene oxide is more reliable, but environmental hazards are greater and running costs high. Both processes are time-consuming and the use of sporicidal disinfectants such as glutaraldehyde is often the only practical alternative. Before purchasing any gaseous sterilizer it is essential to consider throughput and the availability of alternative processes. It may prove sensible to share facilities or at least offer a regional facility. It is certainly not worthwhile purchasing expensive gas sterilizers for reprocessing inexpensive single-use items or for those that require disinfection only. Low temperature steam is safe, inexpensive and no special environmental provisions are necessary. It is, however, not a sterilization process. Disinfectants, hot water and steam will continue to be the only suitable methods for reprocessing items outside the hospital sterile supply department or disinfection unit. Concern over the decontamination of blood-stained instruments following use on patients with hepatitis B or HIV has led to an upsurge of interest in boilers and inexpensive bench top ovens and autoclaves. Such processes are likely to prove more effective than disinfectants but should heat treatment prove impractical then 2% glutaraldehyde or 70% alcohol may be used.(ABSTRACT TRUNCATED AT 250 WORDS)

  9. SUMMARY OF 2010 DOE EM INTERNATIONAL PROGRAM STUDIES OF WASTE GLASS STRUCTURE AND PROPERTIES

    SciTech Connect

    Fox, K.; Choi, A.; Marra, J.; Billings, A.

    2011-02-07

    Collaborative work between the Savannah River National Laboratory (SRNL) and SIA Radon in Russia was divided among three tasks for calendar year 2010. The first task focused on the study of simplified high level waste glass compositions with the objective of identifying the compositional drivers that lead to crystallization and poor chemical durability. The second task focused on detailed characterization of more complex waste glass compositions with unexpectedly poor chemical durabilities. The third task focused on determining the structure of select high level waste glasses made with varying frit compositions in order to improve models under development for predicting the melt rate of the Defense Waste Processing Facility (DWPF) glasses. The majority of these tasks were carried out at SIA Radon. Selection and fabrication of the glass compositions, along with chemical composition measurements and evaluations of durability were carried out at SRNL and are described in this report. SIA Radon provided three summary reports based on the outcome of the three tasks. These reports are included as appendices to this document. Briefly, the result of characterization of the Task 1 glasses may indicate that glass compositions where iron is predominantly tetrahedrally coordinated have more of a tendency to crystallize nepheline or nepheline-like phases. For the Task 2 glasses, the results suggested that the relatively low fraction of tetrahedrally coordinated boron and the relatively low concentrations of Al{sub 2}O{sub 3} available to form [BO{sub 4/2}]{sup -}Me{sup +} and [AlO{sub 4/2}]{sup -}Me{sup +} tetrahedral units are not sufficient to consume all of the alkali ions, and thus these alkali ions are easily leached from the glasses. All of the twelve Task 3 glass compositions were determined to be mainly amorphous, with some minor spinel phases. Several key structural units such as metasilicate chains and rings were identified, which confirms the current modeling

  10. MERCURY REMOVAL FROM DOE SOLID MIXED WASTE USING THE GEMEP(sm) TECHNOLOGY

    SciTech Connect

    1999-03-01

    Under the sponsorship of the Federal Energy Technology Center (FETC), Metcalf and Eddy (M and E), in association with General Electric Corporate Research and Development Center (GE-CRD), Colorado Minerals Research Institute (CMRI), and Oak Ridge National Laboratory (ORNL), conducted laboratory-scale and bench-scale tests of the General Electric Mercury Extraction Process technology on two mercury-contaminated mixed solid wastes from U. S. Department of Energy sites: sediment from the East Fork of Poplar Creek, Oak Ridge (samples supplied by Oak Ridge National Laboratory), and drummed soils from Idaho National Environmental and Engineering Laboratory (INEEL). Fluorescent lamps provided by GE-CRD were also studied. The GEMEP technology, invented and patented by the General Electric Company, uses an extraction solution composed of aqueous potassium iodide plus iodine to remove mercury from soils and other wastes. The extraction solution is regenerated by chemical oxidation and reused, after the solubilized mercury is removed from solution by reducing it to the metallic state. The results of the laboratory- and bench-scale testing conducted for this project included: (1) GEMEP extraction tests to optimize extraction conditions and determine the extent of co-extraction of radionuclides; (2) pre-screening (pre-segregation) tests to determine if initial separation steps could be used effectively to reduce the volume of material needing GEMEP extraction; and (3) demonstration of the complete extraction, mercury recovery, and iodine recovery and regeneration process (known as locked-cycle testing).

  11. Combination gas-producing and waste-water disposal well. [DOE patent application

    DOEpatents

    Malinchak, R.M.

    1981-09-03

    The present invention is directed to a waste-water disposal system for use in a gas recovery well penetrating a subterranean water-containing and methane gas-bearing coal formation. A cased bore hole penetrates the coal formation and extends downwardly therefrom into a further earth formation which has sufficient permeability to absorb the waste water entering the borehole from the coal formation. Pump means are disposed in the casing below the coal formation for pumping the water through a main conduit towards the water-absorbing earth formation. A barrier or water plug is disposed about the main conduit to prevent water flow through the casing except for through the main conduit. Bypass conduits disposed above the barrier communicate with the main conduit to provide an unpumped flow of water to the water-absorbing earth formation. One-way valves are in the main conduit and in the bypass conduits to provide flow of water therethrough only in the direction towards the water-absorbing earth formation.

  12. Materials and Security Consolidation Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

    SciTech Connect

    Not Listed

    2011-09-01

    Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Materials and Security Consolidation Center facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool for developing the radioactive waste management basis.

  13. Research and Education Campus Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

    SciTech Connect

    L. Harvego; Brion Bennett

    2011-11-01

    U.S. Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory Research and Education Campus facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool to develop the radioactive waste management basis.

  14. Materials and Fuels Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

    SciTech Connect

    Lisa Harvego; Brion Bennett

    2011-09-01

    Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Materials and Fuels Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool for developing the radioactive waste management basis.

  15. Central Facilities Area Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

    SciTech Connect

    Lisa Harvego; Brion Bennett

    2011-11-01

    Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Central Facilities Area facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facilityspecific documents. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool for developing the radioactive waste management basis.

  16. Geology Report: Area 3 Radioactive Waste Management Site DOE/Nevada Test Site, Nye County, Nevada

    SciTech Connect

    NSTec Environmental Management

    2006-07-01

    Surficial geologic studies near the Area 3 Radioactive Waste Management Site (RWMS) were conducted as part of a site characterization program. Studies included evaluation of the potential for future volcanism and Area 3 fault activity that could impact waste disposal operations at the Area 3 RWMS. Future volcanic activity could lead to disruption of the Area 3 RWMS. Local and regional studies of volcanic risk indicate that major changes in regional volcanic activity within the next 1,000 years are not likely. Mapped basalts of Paiute Ridge, Nye Canyon, and nearby Scarp Canyon are Miocene in age. There is a lack of evidence for post-Miocene volcanism in the subsurface of Yucca Flat, and the hazard of basaltic volcanism at the Area 3 RWMS, within the 1,000-year regulatory period, is very low and not a forseeable future event. Studies included a literature review and data analysis to evaluate unclassified published and unpublished information regarding the Area 3 and East Branch Area 3 faults mapped in Area 3 and southern Area 7. Two trenches were excavated along the Area 3 fault to search for evidence of near-surface movement prior to nuclear testing. Allostratigraphic units and fractures were mapped in Trenches ST02 and ST03. The Area 3 fault is a plane of weakness that has undergone strain resulting from stress imposed by natural events and underground nuclear testing. No major vertical displacement on the Area 3 fault since the Early Holocene, and probably since the Middle Pleistocene, can be demonstrated. The lack of major displacement within this time frame and minimal vertical extent of minor fractures suggest that waste disposal operations at the Area 3 RWMS will not be impacted substantially by the Area 3 fault, within the regulatory compliance period. A geomorphic surface map of Yucca Flat utilizes the recent geomorphology and soil characterization work done in adjacent northern Frenchman Flat. The approach taken was to adopt the map unit boundaries (line

  17. Process monitoring in international safeguards for reprocessing plants: A demonstration

    SciTech Connect

    Ehinger, M.H.

    1989-01-01

    In the period 1985--1987, the Oak Ridge National Laboratory investigated the possible role of process monitoring for international safeguards applications in fuel reprocessing plants. This activity was conducted under Task C.59, ''Review of Process Monitoring Safeguards Technology for Reprocessing Facilities'' of the US program of Technical Assistance to the International Atomic Energy Agency (IAEA) Safeguards program. The final phase was a demonstration of process monitoring applied in a prototypical reprocessing plant test facility at ORNL. This report documents the demonstration and test results. 35 figs.

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

    SciTech Connect

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

    1997-01-01

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

  19. Proceedings of the fourth annual participants' information meeting, DOE Low-Level Waste Management Program

    SciTech Connect

    Large, D.E.: Mezga, L.J.; Stratton, L.E.; Rose, R.R.

    1982-10-01

    The Fourth Annual Participants' Information Meeting of the Department of Energy Low-Level Waste Management Program was held in Denver, Colorado, August 31 to September 2, 1982. The purpose of the meeting was to report and evaluate technology development funded by the program and to examine mechanisms for technology transfer. The meeting consisted of an introductory plenary session, followed by two concurrent overview sessions and then six concurrent technical sessions. There were two group meetings to review the findings of the technical sessions. The meeting concluded with a plenary summary session in which the major findings of the meeting were addressed. All papers have been abstracted and indexed for the Energy Data Base.

  20. Evaluation and compilation of DOE waste package test data: Biannual report, February 1987--July 1987

    SciTech Connect

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

    1988-05-01

    The waste package is a proposed engineering 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 steels, stainless steels, and copper. The current level of understanding of several canister materials is questioned for the candidate repository in tuff. Three issues are addressed, the possibility of the stress-induced failure of Zircaloy, the possible corrosion of copper and copper alloys, and the lack of site-specific characterization data. Discussions are given on problems concerning localized corrosion and environmentally assisted cracking of AISI 1020 steel at elevated temperatures (150{degree}C). For the proposed salt site, the importance of the duration of corrosion tests and some of the conditions that may preclude prompt initiation of needed long-term testing are two issues that are discussed. 31 refs., 5 figs.

  1. CONTAINMENT OF LOW-LEVEL RADIOACTIVE WASTE AT THE DOE SALTSTONE DISPOSAL FACILITY

    SciTech Connect

    Jordan, J.; Flach, G.

    2012-03-29

    As facilities look for permanent storage of toxic materials, they are forced to address the long-term impacts to the environment as well as any individuals living in affected area. As these materials are stored underground, modeling of the contaminant transport through the ground is an essential part of the evaluation. The contaminant transport model must address the long-term degradation of the containment system as well as any movement of the contaminant through the soil and into the groundwater. In order for disposal facilities to meet their performance objectives, engineered and natural barriers are relied upon. Engineered barriers include things like the design of the disposal unit, while natural barriers include things like the depth of soil between the disposal unit and the water table. The Saltstone Disposal Facility (SDF) at the Savannah River Site (SRS) in South Carolina is an example of a waste disposal unit that must be evaluated over a timeframe of thousands of years. The engineered and natural barriers for the SDF allow it to meet its performance objective over the long time frame. Some waste disposal facilities are required to meet certain standards to ensure public safety. These type of facilities require an engineered containment system to ensure that these requirements are met. The Saltstone Disposal Facility (SDF) at the Savannah River Site (SRS) is an example of this type of facility. The facility is evaluated based on a groundwater pathway analysis which considers long-term changes to material properties due to physical and chemical degradation processes. The facility is able to meet these performance objectives due to the multiple engineered and natural barriers to contaminant migration.

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

    SciTech Connect

    1995-11-01

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

  3. Krypton-85 health risk assessment for a nuclear fuel reprocessing plant

    SciTech Connect

    Mellinger, P.J.; Brackenbush, L.W.; Tanner, J.E.; Gilbert, E.S.

    1984-08-01

    The risks involved in the routine release of /sup 85/Kr from nuclear fuel reprocessing operations to the environment were compared to those resulting from the capture and storage of /sup 85/Kr. Instead of releasing the /sup 85/Kr to the environment when fuel is reprocessed, it can be captured, immobilized and stored. Two alternative methods of capturing /sup 85/Kr (cryogenic distillation and fluorocarbon absorption) and one method of immobilizing the captured gas (ion implantation/sputtering) were theoretically incorporated into a representative fuel reprocessing plant, the Barnwell Nuclear Fuel Plant, even though there are no known plans to start up this facility. Given the uncertainties in the models used to generate lifetime risk numbers (0.02 to 0.027 radiation induced fatal cancers expected in the occupational workforce and 0.017 fatal cancers in the general population), the differences in total risks for the three situations, (i.e., no-capture and two-capture alternatives) cannot be considered meaningful. It is possible that no risks would occur from any of the three situations. There is certainly no reason to conclude that risks from /sup 85/Kr routinely released to the environment are greater than those that would result from the other two situations considered. Present regulations mandate recovery and disposal of /sup 85/Kr from the off gases of a facility reprocessing spent fuel from commercial sources. Because of the lack of a clear-cut indication that recovery woud be beneficial, it does not seem prudent to burden the facilities with a requirement for /sup 85/Kr recovery, at least until operating experience demonstrates the incentive. The probable high aging of the early fuel to be processed and the higher dose resulting from the release of the unregulated /sup 3/H and /sup 14/C also encourage delaying implementation of the /sup 85/Kr recovery in the early plants.

  4. DOE`s Phytoremediation Program

    SciTech Connect

    Levine, R.S.

    1996-12-31

    This presentation contains an outline of the US DOE`s phytoremediation program. A brief overview of the goals, infrastructure, and results of the program is presented. Environmental contaminants addressed include chlorinated hydrocarbons, metals, radionuclides, inorganic wastes, and mixed hazardous and radioactive wastes. Studies of soil remediation using phytoextraction and water remediation using rhizofiltration are briefly described.

  5. Spent Fuel Reprocessing: More Value for Money Spent in a Geological Repository?

    SciTech Connect

    Kaplan, P.; Vinoche, R.; Devezeaux, J-G.; Bailly, F.

    2003-02-25

    Today, each utility or country operating nuclear power plants can select between two long-term spent fuel management policies: either, spent fuel is considered as waste to dispose of through direct disposal or, spent fuel is considered a resource of valuable material through reprocessing-recycling. Reading and listening to what is said in the nuclear community, we understand that most people consider that the choice of policy is, actually, a choice among two technical paths to handle spent fuel: direct disposal versus reprocessing. This very simple situation has been recently challenged by analysis coming from countries where both policies are on survey. For example, ONDRAF of Belgium published an interesting study showing that, economically speaking for final disposal, it is worth treating spent fuel rather than dispose of it as a whole, even if there is no possibility to recycle the valuable part of it. So, the question is raised: is there such a one-to-one link between long term spent fuel management political option and industrial option? The purpose of the presentation is to discuss the potential advantages and drawbacks of spent fuel treatment as an implementation of the policy that considers spent fuel as waste to dispose of. Based on technical considerations and industrial experience, we will study qualitatively, and quantitatively when possible, the different answers proposed by treatment to the main concerns of spent-fuel-as-a-whole geological disposal.

  6. Transuranic contaminated waste functional definition and implementation

    SciTech Connect

    Kniazewycz, B.G.

    1980-03-01

    The purpose of this report is to examine the problem(s) of TRU waste classification and to document the development of an easy-to-apply standard(s) to determine whether or not this waste package should be emplaced in a geologic repository for final disposition. Transuranic wastes are especially significant because they have long half-lives and some are rather radiotoxic. Transuranic radionuclides are primarily produced by single or multiple neutron capture by U-238 in fuel elements during the operation of a nuclear reactor. Reprocessing of spent fuel elements attempts to remove plutonium, but since the separation is not complete, the resulting high-activity liquids still contain some plutonium as well as other transuranics. Likewise, transuranic contamination of low-activity wastes also occurs when the transuranic materials are handled or processed, which is primarily at federal facilities involved in R and D and nuclear weapons production. Transuranics are persistent in the environment and, as a general rule, are strongly retained by soils. They are not easily transported through most food chains, although some reconcentration does take place in the aquatic food chain. They pose no special biological hazard to humans upon ingestion because they are weakly absorbed from the gastrointestional tract. A greater hazard results from inhalation since they behave like normal dust and fractionate accordingly.

  7. High quality reprocessed GPS Zenith Total Delay dataset over Europe

    NASA Astrophysics Data System (ADS)

    Pacione, Rosa; Pace, Brigida; Bianco, Giuseppe

    2015-04-01

    The present availability of 18 years of GPS data belonging to the European Permanent Network (EPN, http://www.epncb.oma.be/) is a valuable database for the development of a climate data record of GPS tropospheric products. We homogeneously reprocessed the whole EPN network for the period 1996-2013 in a consistent way using GIPSY-OASIS II software and applying the state-of-the-art models. This ongoing reprocessing effort, part of the EPN Repro2 initiative, will provide a GPS climate data record over Europe with high potential for monitoring trend and variability in atmospheric water vapour thus improving the knowledge of climatic trends of atmospheric water vapour, being useful for global and regional NWP reanalyses and climate model simulations. These reprocessed ZTD time series will be evaluated against radiosonde data as well as independently reprocessed GPS ZTD time-series.

  8. Advanced Safeguards Approaches for New Reprocessing Facilities

    SciTech Connect

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Richard; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-06-24

    U.S. efforts to promote the international expansion of nuclear energy through the Global Nuclear Energy Partnership (GNEP) will result in a dramatic expansion of nuclear fuel cycle facilities in the United States. New demonstration facilities, such as the Advanced Fuel Cycle Facility (AFCF), the Advanced Burner Reactor (ABR), and the Consolidated Fuel Treatment Center (CFTC) will use advanced nuclear and chemical process technologies that must incorporate increased proliferation resistance to enhance nuclear safeguards. The ASA-100 Project, “Advanced Safeguards Approaches for New Nuclear Fuel Cycle Facilities,” commissioned by the NA-243 Office of NNSA, has been tasked with reviewing and developing advanced safeguards approaches for these demonstration facilities. Because one goal of GNEP is developing and sharing proliferation-resistant nuclear technology and services with partner nations, the safeguards approaches considered are consistent with international safeguards as currently implemented by the International Atomic Energy Agency (IAEA). This first report reviews possible safeguards approaches for the new fuel reprocessing processes to be deployed at the AFCF and CFTC facilities. Similar analyses addressing the ABR and transuranic (TRU) fuel fabrication lines at AFCF and CFTC will be presented in subsequent reports.

  9. Environmental Impacts on Nuclear Reprocessing Solvents

    NASA Astrophysics Data System (ADS)

    Gillens, A. R.; Fessenden, J. E.

    2009-12-01

    Nuclear tests have been employed ever since the first nuclear explosion in Alamogordo, NM during the mid-1940s. Nuclear weapons pose a threat to civil society and result in extensive biological (medical) damages. For this reason, treaties banning nuclear tests and weapons have been employed since the 1960s to cease proliferation of weapons. However, as nuclear tests continue in secrecy and actinides, such as plutonium and uranium, are eligible for theft, nuclear forensics is needed to prevent weapons proliferation. In this study, solvents [tributyl phosphate (TBP), dodecane, decanol] used in reprocessing spent nuclear fuel are analyzed using an isotope ratio mass spectrometer, which provides indisputable evidence in identifying the operation in which solvents were used. Solvent samples are observed under variable conditions in the laboratory for different time periods. It is assumed that their carbon isotope values (δ13C) will become more positive (shift heavy) with time. It is found that the solvents are hygroscopic. TBP leaves the most robust signature compared to the other solvents studied and the isotope values for all solvents under all conditions become more positive with time. This study serves as primary research in understanding how solvents behave under variable conditions in the laboratory and how this could be translated to the environment in fate and transport studies.

  10. Syndrome of inappropriate antidiuresis and cerebral salt wasting syndrome: are they different and does it matter?

    PubMed

    Moritz, Michael L

    2012-05-01

    The syndrome of inappropriate antidiudresis (SIAD) and cerebral salt wasting (CSW) are similar conditions with the main difference being the absence or presence of volume depletion. The two conditions may be indistinguishable at presentation, as volume status is difficult to assess, which can lead to under-diagnosis of CSW in patients with central nervous system (CNS) disease. Carefully conducted studies in patients with CNS disease have indicated that CSW may be more common than SIAD. CSW may be differentiated from SIAD based on the persistence of hypouricemia and increased fractional excretion of urate following the correction of hyponatremia. Hyponatremia should be prevented if possible and treated promptly when discovered in patients with CNS disease as even mild hyponatremia could lead to neurological deterioration. Fluid restriction should not be used for the prevention or treatment of hyponatremia in hospitalized patients with CNS disease as it could lead to volume depletion especially if CSW is present. 0.9% sodium chloride may not be sufficiently hypertonic for the prevention of hyponatremia in hospitalized patients with CNS disease and a more hypertonic fluid may be required. The preferred therapy for the treatment of hyponatremia in patients with CNS disease is 3% sodium chloride.

  11. Immobilization of fission products arising from pyrometallurgical reprocessing in chloride media

    NASA Astrophysics Data System (ADS)

    Leturcq, G.; Grandjean, A.; Rigaud, D.; Perouty, P.; Charlot, M.

    2005-12-01

    Spent nuclear fuel reprocessing to recover energy-producing elements such as uranium or plutonium can be performed by a pyrochemical process. In such method, the actinides and fission products are extracted by electrodeposition in a molten chloride medium. These processes generate chlorinated alkali salt flows contaminated by fission products, mainly Cs, Ba, Sr and rare earth elements constituting high-level waste. Two possible alternatives are investigated for managing this wasteform; a protocol is described for dechlorinating the fission products to allow vitrification, and mineral phases capable of immobilizing chlorides are listed to allow specification of a dedicated ceramic matrix suitable for containment of these chlorinated waste streams. The results of tests to synthesize chlorosilicate phases are also discussed.

  12. On-Line Monitoring and Control of Radiochemical Streams at Spent Fuel Reprocessing Plant

    SciTech Connect

    Levitskaia, Tatiana G.; Bryan, Samuel A.

    2008-05-23

    Techniques are needed to provide on-line monitoring and control of the radiochemical processes that are being developed and demonstrated under the Global Nuclear Energy Partnership (GNEP) initiative. The instrumentation used to monitor these processes must be robust and must be able to withstand harsh radiation and chemical environments. A new on-line monitoring system satisfying these requirements featuring Raman spectroscopy combined with a Coriolis and conductivity probes, has been recently developed by our research team. It provides immediate chemical data and flow parameters of high-level radioactive waste streams with high brine/high alkalinity generated during retrieval from Hanford nuclear waste storage tanks. We are currently applying similar methodology for monitoring the radiochemical streams generated at the spent fuel reprocessing plant. The nature of these strems calls for additional spectroscopic information, which can be gained by the utilization of UV-vis-NIR capabilities.

  13. The search for advanced remote technology in fast reactor reprocessing

    SciTech Connect

    Burch, W.D.; Herndon, J.N.; Stradley, J.G.

    1990-01-01

    Research and development in fast reactor reprocessing has been under way about 20 years in several countries throughout the world. During the past decade in France and the United Kingdom, active development programs have been carried out in breeder reprocessing. Actual fuels from their demonstration reactors have been reprocessed in small-scale facilities. Early US work in breeder reprocessing was carried out at the EBR-II facilities with the early metal fuels, and interest has renewed recently in metal fuels. A major, comprehensive program, focused on oxide fuels, has been carried out in the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) since 1974. Germany and Japan have also carried out development programs in breeder reprocessing, and Japan appears committed to major demonstration of breeder reactors and their fuel cycles. While much of the effort in all of these programs addressed process chemistry and process hardware, a significant element of many of these programs, particularly the CFRP, has been on advancements in facility concepts and remote maintenance features. This paper will focus principally on the search for improved facility concepts and better maintenance systems in the CFRP and, in turn, on how developments at ORNL have influenced the technology elsewhere.

  14. The search for advanced remote technology in fast reactor reprocessing

    SciTech Connect

    Burch, W.D.; Herndon, J.N.; Stradley, J.G. )

    1990-01-01

    Research and development in fast reactor reprocessing has been under way [approximately] 20 yr in several countries. During the past decade, France and the United Kingdom have developed active programs in breeder reprocessing. Actual fuels from their demonstration reactors have been reprocessed in small-scale facilities. Early US work in breeder reprocessing was carried out at the Experimental Breeder Reactor II (EBR-II) facilities with the early metal fuels, and interest has renewed recently in metal fuels. A major, comprehensive program, focused on oxide fuels, has been carried out in the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) since 1974. The Federal Republic of Germany (FRG) and Japan have also carried out development programs in breeder reprocessing, and Japan appears committed to major demonstration of breeder reactors and their fuel cycles. While much of the effort in these programs addressed process chemistry and process hardware, a significant element of many of these programs, particularly the CFRP, has been on advancements in facility concepts and remote maintenance features. This paper focuses on the search for improved facility concepts and better maintenance systems in the CFRP, and, in turn, on how developments at ORNL have influenced the technology elsewhere.

  15. A proposed classification system for high-level and other radioactive wastes

    SciTech Connect

    Kocher, D. C.; Croff, A. G.

    1987-06-01

    This report presents a proposal for quantitative and generally applicable risk-based definitions of high-level and other radioactive wastes. On the basis of historical descriptions and definitions of high-level waste (HLW), in which HLW has been defined in terms of its source as waste from reprocessing of spent nuclear fuel, we propose a more general definition based on the concept that HLW has two distinct attributes: HLW is (1) highly radioactive and (2) requires permanent isolation. This concept leads to a two-dimensional waste classification system in which one axis, related to ''requires permanent isolation,'' is associated with long-term risks from waste disposal and the other axis, related to ''highly radioactive,'' is associated with shorter-term risks due to high levels of decay heat and external radiation. We define wastes that require permanent isolation as wastes with concentrations of radionuclides exceeding the Class-C limits that are generally acceptable for near-surface land disposal, as specified in the US Nuclear Regulatory Commission's rulemaking 10 CFR Part 61 and its supporting documentation. HLW then is waste requiring permanent isolation that also is highly radioactive, and we define ''highly radioactive'' as a decay heat (power density) in the waste greater than 50 W/m/sup 3/ or an external radiation dose rate at a distance of 1 m from the waste greater than 100 rem/h (1 Sv/h), whichever is the more restrictive. This proposal also results in a definition of Transuranic (TRU) Waste and Equivalent as waste that requires permanent isolation but is not highly radioactive and a definition of low-level waste (LLW) as waste that does not require permanent isolation without regard to whether or not it is highly radioactive.

  16. MATRIX 2 RESULTS OF THE FY07 ENHANCED DOE HIGH-LEVEL WASTE MELTER THROUGHPUT STUDIES AT SRNL

    SciTech Connect

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

    2008-10-23

    High-level waste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). The objective of this study was to generate supplemental validation data that could be used to determine the applicability of the current liquidus temperature (TL) model to expanded DWPF glass composition regions of interest based on higher WLs. Two specific flowsheets were used in this study to provide such insight: (1) Higher WL glasses (45 and 50%) based on future sludge batches that have (and have not) undergone the Al-dissolution process. (2) Coupled operations supported by the Salt Waste Processing Facility (SWPF), which increase the TiO{sub 2} concentration in glass to greater than 2 wt%. Glasses were also selected to address technical issues associated with Al{sub 2}O{sub 3} solubility, nepheline formation, and homogeneity issues for coupled operations. A test matrix of 28 glass compositions was developed to provide insight into these issues. The glasses were fabricated and characterized using chemical composition analysis, X-ray Diffraction (XRD), TL measurement and the Product Consistency Test (PCT). The results of this study are summarized below: (1) TiO{sub 2} concentrations up to {approx} 3.5 wt% were retained in DWPF type glasses, where retention is defined as the absence of crystalline TiO{sub 2} (i.e., unreacted or undissolved) in the as-fabricated glasses. Although this TiO{sub 2} content does not bound the projected SWPF high output flowsheet (up to 6 wt% TiO{sub 2} may be required in glass), these data demonstrate the potential for increasing the TiO{sub 2} limit in glass above the current limit of 2 wt

  17. Methods for environmental monitoring of DOE waste disposal and storage sites: Proposal for optimizing a biological treatment system for denitrification of Y-12 waste streams. Semiannual progress report, November 1, 1987--March 31, 1988

    SciTech Connect

    Hicks, G.M.; Revis, N.

    1988-12-31

    The denitrification process at Y-12 involves the use of sludge to denitrify aqueous plating waste containing relatively high levels of NO{sub 3}. The process from time to time does not denitrify. The factors associated with the failure of the process remains to be resolved. The authors propose to resolve those factors by taking the following research approaches: (1) isolation and identification of microorganisms originating from sewage sludge which are associated with denitrification; (2) define physiological factors required for denitrification in this process system; and (3) define toxic factors associated with the aqueous waste that may affect the process of denitrification.

  18. Vitrification of HLW Produced by Uranium/Molybdenum Fuel Reprocessing in COGEMA's Cold Crucible Melter

    SciTech Connect

    Do Quang, R.; Petitjean, V.; Hollebecque, F.; Pinet, O.; Flament, T.; Prod'homme, A.

    2003-02-25

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R&D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed

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

    SciTech Connect

    Rechard, R.P.

    1993-12-01

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

  20. Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410

    SciTech Connect

    Phillips, Chris; Willis, William; Carter, Robert; Baker, Stephen

    2013-07-01

    Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for use as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions

  1. SRNL CRP progress report [Development of Melt Processed Ceramics for Nuclear Waste Immobilization

    SciTech Connect

    Amoroso, J.; Marra, J.

    2014-10-02

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multiphase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing.

  2. STEAM REFORMING TECHNOLOGY DEMONSTRATION FOR THE DESTRUCTION OF ORGANICS ON ACTUAL DOE SAVANNAH RIVER SITE TANK 48H WASTE 9138

    SciTech Connect

    Burket, P

    2009-02-24

    This paper describes the design of the Bench-scale Steam Reformer (BSR); a processing unit for demonstrating steam reforming technology on actual radioactive waste [1]. It describes the operating conditions of the unit used for processing a sample of Savannah River Site (SRS) Tank 48H waste. Finally, it compares the results from processing the actual waste in the BSR to processing simulant waste in the BSR to processing simulant waste in a large pilot scale unit, the Fluidized Bed Steam Reformer (FBSR), operated at Hazen Research Inc. in Golden, CO. The purpose of this work was to prove that the actual waste reacted in the same manner as the simulant waste in order to validate the work performed in the pilot scale unit which could only use simulant waste.

  3. Advanced Test Reactor Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

    SciTech Connect

    Lisa Harvego; Brion Bennett

    2011-11-01

    U.S. Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Advanced Test Reactor Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. U.S. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool to develop the radioactive waste management basis.

  4. Natural and Enhanced Attenuation of Soil and Groundwater at the Monument Valley, Arizona, DOE Legacy Waste Site—10281

    SciTech Connect

    Waugh, W.J.; Miller, D.E.; Morris, S.A.; Sheader, L.R.; Glenn, E.P.; Moore, D.; Carroll, K.C.; Benally, L.; Roanhorse, M.; Bush, R.P.; none,

    2010-03-07

    The U.S. Department of Energy (DOE), the Navajo Nation, and the University of Arizona are exploring natural and enhanced attenuation remedies for groundwater contamination at a former uranium-ore processing site near Monument Valley, Arizona. DOE removed radioactive tailings from the Monument Valley site in 1994. Nitrate and ammonium, waste products of the milling process, remain in an alluvial groundwater plume spreading from the soil source where tailings were removed. Planting and irrigating two native shrubs, fourwing saltbush and black greasewood, markedly reduced both nitrate and ammonium in the source area over an 8-year period. Total nitrogen dropped from 350 mg/kg in 2000 to less than 200 mg/kg in 2008. Most of the reduction is attributable to irrigation-enhanced microbial denitrification rather than plant uptake. However, soil moisture and percolation flux monitoring show that the plantings control the soil water balance in the source area, preventing additional leaching of nitrogen compounds. Enhanced denitrification and phytoremediation also look promising for plume remediation. Microcosm experiments, nitrogen isotopic fractionation analysis, and solute transport modeling results suggest that (1) up to 70 percent of nitrate in the plume has been lost through natural denitrification since the mill was closed in 1968, and (2) injection of ethanol may accelerate microbial denitrification in plume hot spots. A field-scale ethanol injection pilot study is underway. Landscape-scale remote sensing methods developed for the project suggest that transpiration from restored native phreatophyte populations rooted in the aquifer could limit further expansion of the plume. An evaluation of landfarm phytoremediation, the irrigation of native shrub plantings with high nitrate water pumped from the alluvial aquifer, is also underway.

  5. 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

    Lawless, W.F.

    2013-07-01

    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)

  6. Nuclear waste

    SciTech Connect

    Not Available

    1991-09-01

    Radioactive waste is mounting at U.S. nuclear power plants at a rate of more than 2,000 metric tons a year. Pursuant to statute and anticipating that a geologic repository would be available in 1998, the Department of Energy (DOE) entered into disposal contracts with nuclear utilities. Now, however, DOE does not expect the repository to be ready before 2010. For this reason, DOE does not want to develop a facility for monitored retrievable storage (MRS) by 1998. This book is concerned about how best to store the waste until a repository is available, congressional requesters asked GAO to review the alternatives of continued storage at utilities' reactor sites or transferring waste to an MRS facility, GAO assessed the likelihood of an MRSA facility operating by 1998, legal implications if DOE is not able to take delivery of wastes in 1998, propriety of using the Nuclear Waste Fund-from which DOE's waste program costs are paid-to pay utilities for on-site storage capacity added after 1998, ability of utilities to store their waste on-site until a repository is operating, and relative costs and safety of the two storage alternatives.

  7. DOE-owned spent nuclear fuel program plan

    SciTech Connect

    1995-11-01

    The Department of Energy (DOE) has produced spent nuclear fuel (SNF) for many years as part of its various missions and programs. The historical process for managing this SNF was to reprocess it whereby valuable material such as uranium or plutonium was chemically separated from the wastes. These fuels were not intended for long-term storage. As the need for uranium and plutonium decreased, it became necessary to store the SNF for extended lengths of time. This necessity resulted from a 1992 DOE decision to discontinue reprocessing SNF to recover strategic materials (although limited processing of SNF to meet repository acceptance criteria remains under consideration, no plutonium or uranium extraction for other uses is planned). Both the facilities used for storage, and the fuel itself, began experiencing aging from this extended storage. New efforts are now necessary to assure suitable fuel and facility management until long-term decisions for spent fuel disposition are made and implemented. The Program Plan consists of 14 sections as follows: Sections 2--6 describe objectives, management, the work plan, the work breakdown structure, and the responsibility assignment matrix. Sections 7--9 describe the program summary schedules, site logic diagram, SNF Program resource and support requirements. Sections 10--14 present various supplemental management requirements and quality assurance guidelines.

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

    SciTech Connect

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

    2002-02-26

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

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

    SciTech Connect

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

    2002-01-01

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

  10. Integration of the informal sector into municipal solid waste management in the Philippines--what does it need?

    PubMed

    Paul, Johannes G; Arce-Jaque, Joan; Ravena, Neil; Villamor, Salome P

    2012-11-01

    The integration of the informal sector into municipal solid waste management is a challenge many developing countries face. In Iloilo City, Philippines around 220 tons of municipal solid waste are collected every day and disposed at a 10 ha large dumpsite. In order to improve the local waste management system the Local Government decided to develop a new Waste Management Center with integrated landfill. However, the proposed area is adjacent to the presently used dumpsite where more than 300 waste pickers dwell and depend on waste picking as their source of livelihood. The Local Government recognized the hidden threat imposed by the waste picker's presence for this development project and proposed various measures to integrate the informal sector into the municipal solid waste management (MSWM) program. As a key intervention a Waste Workers Association, called USWAG Calahunan Livelihood Association Inc. (UCLA) was initiated and registered as a formal business enterprise in May 2009. Up to date, UCLA counts 240 members who commit to follow certain rules and to work within a team that jointly recovers wasted materials. As a cooperative they are empowered to explore new livelihood options such as the recovery of Alternative Fuels for commercial (cement industry) and household use, production of compost and making of handicrafts out of used packages. These activities do not only provide alternative livelihood for them but also lessen the generation of leachate and Greenhouse Gases (GHG) emissions from waste disposal, whereby the life time of the proposed new sanitary landfill can be extended likewise.

  11. Integration of the informal sector into municipal solid waste management in the Philippines - What does it need?

    SciTech Connect

    Paul, Johannes G.

    2012-11-15

    The integration of the informal sector into municipal solid waste management is a challenge many developing countries face. In Iloilo City, Philippines around 220 tons of municipal solid waste are collected every day and disposed at a 10 ha large dumpsite. In order to improve the local waste management system the Local Government decided to develop a new Waste Management Center with integrated landfill. However, the proposed area is adjacent to the presently used dumpsite where more than 300 waste pickers dwell and depend on waste picking as their source of livelihood. The Local Government recognized the hidden threat imposed by the waste picker's presence for this development project and proposed various measures to integrate the informal sector into the municipal solid waste management (MSWM) program. As a key intervention a Waste Workers Association, called USWAG Calahunan Livelihood Association Inc. (UCLA) was initiated and registered as a formal business enterprise in May 2009. Up to date, UCLA counts 240 members who commit to follow certain rules and to work within a team that jointly recovers wasted materials. As a cooperative they are empowered to explore new livelihood options such as the recovery of Alternative Fuels for commercial (cement industry) and household use, production of compost and making of handicrafts out of used packages. These activities do not only provide alternative livelihood for them but also lessen the generation of leachate and Greenhouse Gases (GHG) emissions from waste disposal, whereby the life time of the proposed new sanitary landfill can be extended likewise.

  12. Integrated process for reprocessing spent nuclear fuel

    SciTech Connect

    Forsberg, C.W.

    1991-03-06

    This invention is comprised of a process for recovering nuclear fuel from spent fuel assemblies that employs a single canister process container. The cladding and fuel are oxidized in the container, the fuel is dissolved and removed from the container for separation from the aqueous phase, the aqueous phase containing radioactive waste is returned to the container. This container is also the disposal vessel. Add solidification agents and compress container for long term storage.

  13. DKPRO: A radionuclide decay and reprocessing code

    SciTech Connect

    Wootan, D.; Schmittroth, F.A.

    1997-07-14

    The DKPRO code solves the general problem of modeling complex nuclear wastes streams using ORIGEN2 radionuclide production files. There is a continuing need for estimates of Hanford radionuclides. Physical measurements are one basis; calculational estimates, the approach represented here, are another. Given a known nuclear fuel history, it is relatively straightforward to calculate radionuclide inventories with codes such as the widely-used Oak Ridge National Laboratory code ORIGEN2.

  14. Adverse staff health outcomes associated with endoscope reprocessing.

    PubMed

    Gutterman, Elane; Jorgensen, Lindsay; Mitchell, Amber; Fua, Sherry

    2013-01-01

    There are occupational challenges associated with cleaning, disinfecting, storing, and transporting flexible endoscopes. Although the Occupational Safety and Health Administration (OSHA) has set standards to protect the safety of health workers in the United States, the standards are not specific to endoscope reprocessing, and the general standards that are in place are not fully implemented. Furthermore, adverse staff outcomes may not be fully preventable. To assess the evidence for adverse outcomes in staff associated with endoscope reprocessing, a literature review was performed in the PubMed database for articles on this topic published between Jan. 1, 2007 and March 7, 2012. Eight studies were identified, mainly European, which reported numerous adverse outcomes to healthcare personnel associated with endoscope reprocessing including respiratory ailments and physical discomfort. More scientifically rigorous studies are required to comprehensively describe adverse health outcomes in personnel engaged in reprocessing, particularly in the United States, and examine whether increased automation of the reprocessing process leads to decreased adverse health outcomes for staff.

  15. Disposable Bronchoscope Model for Simulating Endoscopic Reprocessing and Surveillance Cultures.

    PubMed

    Yassin, Mohamed H; Hariri, Rahman; Hamad, Yasir; Ferrelli, Juliet; McKibben, Leeanna; Doi, Yohei

    2017-02-01

    BACKGROUND Endoscope-associated infections are reported despite following proper reprocessing methods. Microbiological testing can confirm the adequacy of endoscope reprocessing. Multiple controversies related to the method and interpretation of microbiological testing cultures have arisen that make their routine performance a complex target. OBJECTIVE We conducted a pilot study using disposable bronchoscopes (DBs) to simulate different reprocessing times and soaking times and to compare high-level disinfection versus ethylene oxide sterilization. We also reviewed the time to reprocessing and duration of the procedures. METHODS Bronchoscopes were chosen because an alternative disposable scope is commercially available and because bronchoscopes are more prone to delays in processing. Disposable bronchoscopes were contaminated using a liquid bacterial suspension and were then incubated for 1-4 hours. Standard processing and high-level disinfection were performed on 36 endoscopes. Ethylene oxide sterilization was performed on 21 endoscopes. Endoscope cultures were performed using the standard "brush, flush, brush" technique. RESULTS After brushing was performed, a final water-flush culture procedure was the most effective method of detecting bacterial persistence on the disposable scopes. Klebsiella pneumoniae was the most commonly recovered organism after reprocessing. Ethylene oxide sterilization did not result in total elimination of viable bacteria. CONCLUSION Routine endoscopy cultures may be required to assess the adequacy of endoscopic processing. Infect Control Hosp Epidemiol 2017;38:136-142.

  16. Radiological health review of the Final Environmental Impact Statement Waste Isolation Pilot Plant. Volumes 1 and 2. DOE/EIS-0026

    SciTech Connect

    Not Available

    1981-01-01

    Purpose of the Environmental Evaluation Group (EEG) is to conduct an independent technical evaluation of the potential radiation exposure to people from the proposed Federal radioactive Waste Isolation Pilot Plant (WIPP) near Carlsbad, in order to protect the public health and safety and ensure that there is minimal environmental degradation. Analyses are conducted of reports issued by the US DOE and its contractors, other Federal agencies and other organizations, as they relate to the potential health, safety and environmental impacts from WIPP.

  17. Flood Assessment at the Area 5 Radioactive Waste Management Site and the Proposed Hazardous Waste Storage Unit, DOE/Nevada Test Site, Nye County, Nevada

    SciTech Connect

    Schmeltzer, J. S.; Millier, J. J.; Gustafson, D. L.

    1993-01-01

    A flood assessment at the Radioactive Waste Management Site (RWMS) and the proposed Hazardous Waste Storage Unit (HWSU) in Area 5 of the Nevada Test Site (NTS) was performed to determine the 100-year flood hazard at these facilities. The study was conducted to determine whether the RWMS and HWSU are located within a 100-year flood hazard as defined by the Federal Emergency Management Agency, and to provide discharges for the design of flood protection.

  18. Closed Fuel Cycle Waste Treatment Strategy

    SciTech Connect

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

    2015-02-01

    This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form

  19. Spent fuel and residue measurement instrumentation at the Sellafield nuclear fuel reprocessing facility

    SciTech Connect

    Chesterman, A.S.; Clark, P.A.

    1995-12-31

    The Sellafield reprocessing plant receives and reprocesses several thousand tonnes of spent light water reactor (LWR), advanced gas cooled reactor (AGR) and natural uranium magnesium alloy clad (Magnox) fuels each year. The safety and cost effectiveness of these operations has been supported by the development and installation, at key points in the process, of a range of special purpose radiometric instrumentation. Systems in routine operational use verify the cooling time, burn-tip and initial and final U-235 equivalent enrichment of fuel assemblies in the storage and handling ponds. Other systems determine the radionuclide inventories of fuel residues in intermediate level waste arising from plant operations. The measurement techniques employed include high resolution gamma spectrometry, passive neutron counting and neutron interrogation by the use of a Cf-252 source and deuterium-tritium (D-T) pulsed neutron generators. Details of the instruments including mechanical installation arrangements and measurement data are presented in the paper along with a discussion of possible future uses of similar instruments for burn-up credit associated with fuel and residue storage, transportation and disposal.

  20. Remote maintenance lessons learned'' on prototypical reprocessing equipment

    SciTech Connect

    Kring, C.T.; Schrock, S.L.

    1990-01-01

    Hardware representative of essentially every major equipment item necessary for reprocessing breeder reactor nuclear fuel has been installed and tested for remote maintainability. This testing took place in a cold mock-up of a remotely maintained hot cell operated by the Consolidated Fuel Reprocessing Program (CFRP) within the Fuel Recycle Division at Oak Ridge National Laboratory (ORNL). The reprocessing equipment tested included a Disassembly System, a Shear System, a Dissolver System, an Automated Sampler System, removable Equipment Racks on which various chemical process equipment items were mounted, and an advanced servomanipulator (ASM). These equipment items were disassembled and reassembled remotely by using the remote handling systems that are available within the cold mock-up area. This paper summarizes the lessons learned'' as a result of the numerous maintenance activities associated with each of these equipment items. 4 refs., 3 figs., 1 tab.

  1. Equipment specifications for an electrochemical fuel reprocessing plant

    SciTech Connect

    Hemphill, Kevin P

    2010-01-01

    Electrochemical reprocessing is a technique used to chemically separate and dissolve the components of spent nuclear fuel, in order to produce new metal fuel. There are several different variations to electrochemical reprocessing. These variations are accounted for by both the production of different types of spent nuclear fuel, as well as different states and organizations doing research in the field. For this electrochemical reprocessing plant, the spent fuel will be in the metallurgical form, a product of fast breeder reactors, which are used in many nuclear power plants. The equipment line for this process is divided into two main categories, the fuel refining equipment and the fuel fabrication equipment. The fuel refining equipment is responsible for separating out the plutonium and uranium together, while getting rid of the minor transuranic elements and fission products. The fuel fabrication equipment will then convert this plutonium and uranium mixture into readily usable metal fuel.

  2. A Review of Current Disinfectants for Gastrointestinal Endoscopic Reprocessing

    PubMed Central

    Park, Sanghoon; Koo, Ja Seol; Park, Jeong Bae; Lim, Yun Jeong; Hong, Su Jin; Kim, Sang-Woo; Chun, Hoon Jai

    2013-01-01

    Gastrointestinal endoscopy is gaining popularity for diagnostic and therapeutic purposes. However, concerns over endoscope-related nosocomial infections are increasing, together with interest by the general public in safe and efficient endoscopy. For this reason, reprocessing the gastrointestinal endoscope is an important step for effective performance of endoscopy. Disinfectants are essential to the endoscope reprocessing procedure. Before selecting an appropriate disinfectant, their characteristics, limitations and means of use must be fully understood. Herein, we review the characteristics of several currently available disinfectants, including their uses, potency, advantages, and disadvantages. Most disinfectants can be used to reprocess gastrointestinal endoscopes if the manufacturer's guidelines are followed. The selection and use of a suitable disinfectant depends on the individual circumstances of each endoscopy suite. PMID:23964330

  3. DEVELOPMENT OF GLASS COMPOSITIONS TO IMMOBILIZE ALKALI, ALKALINE EARTH, LANTHANIDE AND TRANSITION METAL FISSION PRODUCTS FROM NUCLEAR FUEL REPROCESSING

    SciTech Connect

    Marra, J.; Billings, A.

    2009-06-24

    The Advanced Fuel Cycle Initiative (AFCI) waste management strategy revolves around specific treatment of individual or groups of separated waste streams. A goal for the separations processes is to efficiently manage the waste to be dispositioned as high level radioactive waste. The Advanced Fuel Cycle Initiative (AFCI) baseline technology for immobilization of the lanthanide (Ln) and transition metal fission product (TM) wastes is vitrification into a borosilicate glass. A current interest is to evaluate the feasibility of vitrifying combined waste streams to most cost effectively immobilize the wastes resulting from aqueous fuel reprocessing. Studies showed that high waste loadings are achievable for the Ln only (Option 1) stream. Waste loadings in excess of 60 wt % (on a calcined oxide basis) were demonstrated via a lanthanide borosilicate (LaBS) glass. The resulting glasses had excellent relative durability as determined by the Product Consistency Test (PCT). For a combined Ln and TM waste stream glass (Option 2), noble metal solubility was found to limit waste loading. However, the measured PCT normalized elemental releases for this glass were at least an order of magnitude below that of Environmental Assessment (EA) glass. Current efforts to evaluate the feasibility of vitrifying combined Ln, TM, alkali (Cs is the primary radionuclide of concern) and alkaline earth (Sr is the primary radionuclide of concern) wastes (Option 3) have shown that these approaches are feasible. However, waste loading limitations with respect to heat load (Cs/Sr loading), molybdenum solubility and/or noble metal solubility will likely be realized and must be considered in determining the cost effectiveness of these approaches.

  4. DOE-EMSP Project Report FY 04: Portable Analyzer Based on Microfluidics/Nanoengineered Electrochemical Sensors for In-situ Characterization of Mixed Wastes

    SciTech Connect

    Lin, Yuehe; Yantasee, Wassana; Fryxell, Glen E.; Wang, Zheming; Wang, Joseph

    2004-11-02

    Required characterizations of the DOE's transuranic (TRU) and mixed wastes (MW) before disposing and treatment of the wastes are currently costly and have lengthy turnaround. Research toward developing faster and more sensitive characterization and analysis tools to reduce costs and accelerate throughputs is therefore desirable. This project is aimed at the development of electrochemical sensors, specific to toxic transition metals, uranium, and technetium, that can be integrated into the portable sensor systems. This system development will include fabrication and performance evaluation of electrodes as well as understanding of electrochemically active sites on the electrodes specifically designed for toxic metals, uranium and technetium detection. Subsequently, these advanced measurement units will be incorporated into a microfluidic prototype specifically designed and fabricated for field-deployable characterizations of such species. The electrochemical sensors being investigate d are based on a new class of nanoengineered sorbents, Self-Assembled Monolayer on Mesoporous Supports (SAMMS). SAMMS are highly efficient sorbents due to their interfacial chemistry that can be fine-tuned to selectively sequester a specific target species. Adsorptive stripping voltammetry (AdSV) will be performed on two classes of electrodes: the SAMMS modified carbon paste electrodes, and the SAMMS thin film immobilized on microelectrode arrays. Interfacial chemistry and electrochemistry of metal species on the surfaces of SAMMS-based electrodes will be studied. This fundamental knowledge is required for predicting how the sensors will perform in the real wastes which consist of many interferences/ligands and a spectrum of pH levels. The best electrode for each specific waste constituent will be integrated onto the portable microfluidic platform. Efforts will also be focused on testing the portable microfluidics/electrochemical sensor systems with the selected MW and TRU waste samples

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

    SciTech Connect

    Not Available

    1988-12-01

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

  6. Do the Kepler AGN light curves need reprocessing?

    NASA Astrophysics Data System (ADS)

    Kasliwal, Vishal P.; Vogeley, Michael S.; Richards, Gordon T.; Williams, Joshua; Carini, Michael T.

    2015-10-01

    We gauge the impact of spacecraft-induced effects on the inferred variability properties of the light curve of the Seyfert 1 AGN Zw 229-15 observed by Kepler. We compare the light curve of Zw 229-15 obtained from the Kepler MAST data base with a reprocessed light curve constructed from raw pixel data. We use the first-order structure function, SF(δt), to fit both light curves to the damped power-law PSD (power spectral density) of Kasliwal et al. On short time-scales, we find a steeper log PSD slope (γ = 2.90 to within 10 per cent) for the reprocessed light curve as compared to the light curve found on MAST (γ = 2.65 to within 10 per cent) - both inconsistent with a damped random walk (DRW) which requires γ = 2. The log PSD slope inferred for the reprocessed light curve is consistent with previous results that study the same reprocessed light curve. The turnover time-scale is almost identical for both light curves (27.1 and 27.5 d for the reprocessed and MAST data base light curves). Based on the obvious visual difference between the two versions of the light curve and on the PSD model fits, we conclude that there remain significant levels of spacecraft-induced effects in the standard pipeline reduction of the Kepler data. Reprocessing the light curves will change the model inferenced from the data but is unlikely to change the overall scientific conclusions reached by Kasliwal et al. - not all AGN light curves are consistent with the DRW.

  7. Overview of reductants utilized in nuclear fuel reprocessing/recycling

    SciTech Connect

    Patricia Paviet-Hartmann; Catherine Riddle; Keri Campbell; Edward Mausolf

    2013-10-01

    Most of the aqueous processes developed, or under consideration worldwide for the recycling of used nuclear fuel (UNF) utilize the oxido-reduction properties of actinides to separate them from other radionuclides. Generally, after acid dissolution of the UNF, (essentially in nitric acid solution), actinides are separated from the raffinate by liquid-liquid extraction using specific solvents, associated along the process, with a particular reductant that will allow the separation to occur. For example, the industrial PUREX process utilizes hydroxylamine as a plutonium reductant. Hydroxylamine has numerous advantages: not only does it have the proper attributes to reduce Pu(IV) to Pu(III), but it is also a non-metallic chemical that is readily decomposed to innocuous products by heating. However, it has been observed that the presence of high nitric acid concentrations or impurities (such as metal ions) in hydroxylamine solutions increase the likelihood of the initiation of an autocatalytic reaction. Recently there has been some interest in the application of simple hydrophilic hydroxamic ligands such as acetohydroxamic acid (AHA) for the stripping of tetravalent actinides in the UREX process flowsheet. This approach is based on the high coordinating ability of hydroxamic acids with tetravalent actinides (Np and Pu) compared with hexavalent uranium. Thus, the use of AHA offers a route for controlling neptunium and plutonium in the UREX process by complexant based stripping of Np(IV) and Pu(IV) from the TBP solvent phase, while U(VI) ions are not affected by AHA and remain solvated in the TBP phase. In the European GANEX process, AHA is also used to form hydrophilic complexes with actinides and strip them from the organic phase into nitric acid. However, AHA does not decompose completely when treated with nitric acid and hampers nitric acid recycling. In lieu of using AHA in the UREX + process, formohydroxamic acid (FHA), although not commercially available, hold

  8. National long-term high-level waste-technology program

    NASA Astrophysics Data System (ADS)

    Gray, P. L.

    The national program for long-term management of high level waste (HLW) from nuclear fuels reprocessing is discussed. This covers only DOE defense wastes. Current emphasis is on solidification of waste into a form that, along with additional barriers, may be permanently stored in a repository. An integrated national plan incorporates all the elements of such an overall HLW disposal system. Interim storage is in near-surface tanks at the Hanford and Savannah River sites. At the Idaho site, waste is stored in bins after being calcined. Some Idaho waste is liquid and is also stored in tanks before calcination. Retrieval and immobilization of HLW into a solid, low-release form represent the major elements for which our long-term program has responsibility. Once solidified, the waste will temporarily remain onsite until the final disposal site is prepared for receipt of waste. Currently, a geologic repository is favored for ultimate disposal, although other possibilities such as seabed, icecap, space, and near-surface disposal are also being considered.

  9. Head-end reprocessing equipment remote maintenance demonstration

    SciTech Connect

    Evans, J.H.; Metz, C.F. III.

    1989-01-01

    Prototype equipment for reprocessing breeder reactor nuclear fuel was installed in the Remote Operation and Maintenance Demonstration (ROMD) area of the Consolidated Fuel Reprocessing Program (CFRP) facility at the Oak Ridge National Laboratory (ORNL) in order to evaluate the design of this equipment in a cold mock-up of a remotely maintained hot cell. This equipment included the Remote Disassembly System (RDS) and the Remote Shear System (RSS). These systems were disassembled and reassembled remotely by using the extensive remote handling systems that are installed in this simulated hot-cell environment. 5 refs., 5 figs.

  10. Cleaning of instruments: an absolute requirement for successful reprocessing.

    PubMed

    Wilder, Jonathan A; Roth, Klaus

    2012-01-01

    Cleaning of medical devices, and the validation and verification of this cleaning, has advanced greatly, even in the last five years. It is recognized as an essential part of the reprocessing of reusable medical devices, and is becoming more and more consistent as device manufacturers and the FDA give cleaning more consideration when a device is brought to market. Scrupulous care must be taken to ensure that the cleaning approach for each device is appropriate (but similar items can be reprocessed in a similar manner). There is no replacement for excellent training and maintainance of equipment, but available tests can help ensure that failures are caught earlier rather than later.

  11. External comparisons of reprocessed SBUV/TOMS ozone data

    NASA Astrophysics Data System (ADS)

    Wellemeyer, C. G.; Taylor, S. L.; Singh, R. R.; McPeters, R. D.

    1994-04-01

    Ozone Retrievals from the Solar Backscatter Ultraviolet (SBUV) Instrument on-board the Nimbus-7 Satellite have been reprocessed using an improved internal calibration. The resulting data set covering November, 1978 through January, 1987 has been archived at the National Space Science Data Center in Greenbelt, Maryland. The reprocessed SBUV total ozone data as well as recalibrated Total Ozone Mapping Spectrometer (TOMS) data are compared with total ozone measurements from a network of ground based Dobson spectrophotometers. The SBUV also measures the vertical distribution of ozone, and these measurements are compared with external measurements made by SAGE II, Umkehr, and Ozonesondes. Special attention is paid to long-term changes in ozone bias.

  12. External comparisons of reprocessed SBUV/TOMS ozone data

    NASA Technical Reports Server (NTRS)

    Wellemeyer, C. G.; Taylor, S. L.; Singh, R. R.; Mcpeters, R. D.

    1994-01-01

    Ozone Retrievals from the Solar Backscatter Ultraviolet (SBUV) Instrument on-board the Nimbus-7 Satellite have been reprocessed using an improved internal calibration. The resulting data set covering November, 1978 through January, 1987 has been archived at the National Space Science Data Center in Greenbelt, Maryland. The reprocessed SBUV total ozone data as well as recalibrated Total Ozone Mapping Spectrometer (TOMS) data are compared with total ozone measurements from a network of ground based Dobson spectrophotometers. The SBUV also measures the vertical distribution of ozone, and these measurements are compared with external measurements made by SAGE II, Umkehr, and Ozonesondes. Special attention is paid to long-term changes in ozone bias.

  13. Record of Technical Change {number_sign}1 for ''Corrective Action Investigation Plan for Corrective Action Unit 143: Area 25 Contaminated Waste Dumps, Nevada Test Site, Nevada,'' Revision 1, DOE/NV 506

    SciTech Connect

    US DOE Nevada Operations Office

    1999-07-21

    This Record of Technical Change updates the technical information provided in ''Corrective Action Investigation Plan for Corrective Action Unit 143: Area 25 Contaminated Waste Dumps, Nevada Test Site, Nevada,'' Revision 0, DOE/NV-506

  14. 78 FR 15358 - DOE's Preferred Alternative for Certain Tanks Evaluated in the Final Tank Closure and Waste...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-03-11

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY DOE's... Environmental Impact Statement for the Hanford Site, Richland, Washington AGENCY: U.S. Department of Energy (DOE). ACTION: Notice of DOE's preferred alternative. SUMMARY: The U.S. Department of Energy (DOE) is...

  15. Issues for Conceptual Design of AFCF and CFTC LWR Spent Fuel Separations Influencing Next-Generation Aqueous Fuel Reprocessing

    SciTech Connect

    D. Hebditch; R. Henry; M. Goff; K. Pasamehmetoglu; D. Ostby

    2007-09-01

    In 2007, the U.S. Department of Energy (DOE) published the Global Nuclear Energy Partnership (GNEP) strategic plan, which aims to meet US and international energy, safeguards, fuel supply and environmental needs by harnessing national laboratory R&D, deployment by industry and use of international partnerships. Initially, two industry-led commercial scale facilities, an advanced burner reactor (ABR) and a consolidated fuel treatment center (CFTC), and one developmental facility, an advanced fuel cycle facility (AFCF) are proposed. The national laboratories will lead the AFCF to provide an internationally recognized R&D center of excellence for developing transmutation fuels and targets and advancing fuel cycle reprocessing technology using aqueous and pyrochemical methods. The design drivers for AFCF and the CFTC LWR spent fuel separations are expected to impact on and partly reflect those for industry, which is engaging with DOE in studies for CFTC and ABR through the recent GNEP funding opportunity announcement (FOA). The paper summarizes the state-of-the-art of aqueous reprocessing, gives an assessment of engineering drivers for U.S. aqueous processing facilities, examines historic plant capital costs and provides conclusions with a view to influencing design of next-generation fuel reprocessing plants.

  16. Proceedings of the tenth annual DOE low-level waste management conference: Session 2: Site performance assessment

    SciTech Connect

    Not Available

    1988-12-01

    This document contains twelve papers on various aspects of low-level radioactive waste management. Topics of this volume include: performance assessment methodology; remedial action alternatives; site selection and site characterization procedures; intruder scenarios; sensitivity analysis procedures; mathematical models for mixed waste environmental transport; and risk assessment methodology. Individual papers were processed separately for the database. (TEM)

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

    SciTech Connect

    Not Available

    1988-12-01

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

  18. Applications of curium measurements for safeguarding at large-scale reprocessing plants

    SciTech Connect

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

    1997-08-01

    Safeguarding the plutonium passing through a large-scale reprocessing plant (such as one with 800 t of uranium per year) involves nondestructive assay measurements for plutonium at key points. The gamma-ray and neutron signals from the plutonium are generally hidden by the much larger backgrounds from fission products and actinides, so indirect measurements are routinely used. The intense neutron emission rate from spent fuel is from curium. In a spent fuel assembly at the head-end of a plant, the curium neutrons are used to deduce the amount of plutonium present. Coincidence and multiplicity counting are alternative ways to measure neutrons from spent fuel; they have advantages over total neutron counting in certain conditions and offer new opportunities for examining assemblies. New uses for measurements of curium`s neutrons are proposed to safeguard waste streams. From a year`s work at a large-scale plant, 4 to 7 kg of plutonium can remain in leached hulls and 4 to 22 kg of plutonium can remain in the vitrified high-level liquid waste. While the plutonium in these wastes has the safeguards advantage of being dilute, it is important to verify (a) that the many kilograms involved are in fact present and (b) that the declared masses are not higher than the actual amounts so that more concentrated plutonium cannot pass through the plant by masquerading as waste. Curium measurements on spent fuel assemblies, the accountability tank, and leached hulls would form a safeguards system around all the inputs and outputs of a plant`s head-end where the plutonium is always intimately mixed with the curium. A neutron measurement of the vitrified waste would help identify the presence of a diversion path upstream because essentially all of the curium measured in the spent fuel assemblies should also be found in the vitrified waste (on a batch basis). 7 refs., 4 figs.

  19. Glass ceramics containment matrix for insoluble residues coming from spent fuel reprocessing

    NASA Astrophysics Data System (ADS)

    Pinet, O.; Boën, R.

    2014-04-01

    Spent fuel reprocessing by hydrometallurgical process generates insoluble residues waste streams called fines solution. Considering their radioactivity, fines solution could be considered as Intermediate Level Waste. This waste stream is usually mixed with fission products stream before vitrification. Thus fines are incorporated in glass matrix designed for High Level Waste. The withdrawal of fines from high level glass could decrease the volume of high level waste after conditioning. It could also decrease the reaction time between high level waste and additives to obtain a homogeneous melt and then increase the vitrification process capacity. Separated conditioning of fines in glass matrices has been tested. The fines content targeted value is 16 wt%. To achieve this objective, two types of glass ceramic formulations have been tested. 700 g of the two selected glass ceramics have been prepared using simulated fines. Additives used were ground glass. Melting is achieved at 1100 °C. According to the type of glass ceramic, reducing or oxidizing conditions have been performed during melting. Due to their composition and the melting redox conditions, different phases have been observed. These crystalline phases are typically RuO2, metallic Ru, metallic Pd, MoO2 and CaMoO4. In view of melting these matrices in an in can process the corrosiveness of one of the most oxidizing borosilicate glass ceramic formulation has been tested. This one has been remelted at 1100 °C in inconel 601 pot for 3 days. The oxygen fugacity measurement performed in the remelted glass leads to an oxidizing value, indicating that no significant reaction occurred between the inconel pot and the glass melt had occurred.

  20. Study on release and transport of aerial radioactive materials in reprocessing plants

    SciTech Connect

    Amano, Y.; Tashiro, S.; Uchiyama, G.; Abe, H.; Yamane, Y.; Yoshida, K.; Kodama, T.

    2013-07-01

    The release and transport characteristics of radioactive materials at a boiling accident of the high active liquid waste (HALW) in a reprocessing plant have been studied for improving experimental data of source terms of the boiling accident. In the study, a heating test and a thermogravimetry and differential thermal analysis (TG-DTA) test were conducted. In the heating test using a simulated HALW, it was found that ruthenium was mainly released into the air in the form of gas and that non-volatile elements were released into the air in the form of mist. In the TG-DTA test, the rate constants and reaction heat of thermal decomposition of ruthenium nitrosyl nitrate were obtained from TG and DTA curves. (authors)

  1. Progress in evaluation of radionuclide geochemical information developed by DOE high-level nuclear waste repository site projects. Annual report, October 1984-September 1985. Volume 4

    SciTech Connect

    Meyer, R.E.; Arnold, W.D.; Blencoe, J.G.; Jacobs, G.K.; Kelmers, A.D.; Seeley, F.G.; Whatley, S.K.

    1986-05-01

    Information pertaining to the potential geochemical behavior of radionuclides at candidate sites for a high-level radioactive waste repository, which is being developed by projects within the Department of Energy (DOE), is being evaluated by Oak Ridge National Laboratory for the Nuclear Regulatory Commission (NRC). During this report period, emphasis was placed on the evaluation of information pertinent to the Hanford site in southeastern Washington. Results on the sorption/solubility behavior of technetium, neptunium, and uranium in the basalt/water geochemical system are summarized and compared to the results of DOE. Also, summaries of results are reported from two geochemical modeling studies: (1) an evaluation of the information developed by DOE on the native copper deposits of Michigan as a natural analog for the emplacement of copper canisters in a repository in basalt, and (2) calculation of the solubility and speciation of radionuclides for representative groundwaters from the Yucca Mountain site in Nevada.

  2. Analysis of the Reuse of Uranium Recovered from the Reprocessing of Commercial LWR Spent Fuel

    SciTech Connect

    DelCul, Guillermo Daniel; Trowbridge, Lee D; Renier, John-Paul; Ellis, Ronald James; Williams, Kent Alan; Spencer, Barry B; Collins, Emory D

    2009-02-01

    This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reactor). Reuse is an attractive alternative to the current Advanced Fuel Cycle Initiative (AFCI) Global Nuclear Energy Partnership (GNEP) baseline plan, which stores the reprocessed uranium (RU) for an uncertain future or attempts to dispose of it as 'greater-than-Class C' waste. Considering that the open fuel cycle currently deployed in the United States already creates a huge excess quantity of depleted uranium, the closed fuel cycle should enable the recycle of the major components of spent fuel, such as the uranium and the hazardous, long-lived transuranic (TRU) actinides, as well as the managed disposal of fission product wastes. Compared with the GNEP baseline scenario, the reuse of RU in the uranium fuel cycle has a number of potential advantages: (1) avoidance of purchase costs of 11-20% of the natural uranium feed; (2) avoidance of disposal costs for a large majority of the volume of spent fuel that is reprocessed; (3) avoidance of disposal costs for a portion of the depleted uranium from the enrichment step; (4) depending on the {sup 235}U assay of the RU, possible avoidance of separative work costs; and (5) a significant increase in the production of {sup 238}Pu due to the presence of {sup 236}U, which benefits somewhat the transmutation value of the plutonium and also provides some proliferation resistance.

  3. DOE handbook: Guide to good practices for training and qualification of chemical operators

    SciTech Connect

    1996-03-01

    The purpose of this Handbook is to provide contractor training organizations with information that can be used as a reference to refine existing chemical operator training programs, or develop new training programs where no program exists. This guide, used in conjunction with facility-specific job analyses, will provide a framework for training and qualification programs for chemical operators at DOE reactor and nonreactor facilities. Recommendations for qualification are made in four areas: education, experience, physical attributes, and training. Contents include: initial qualification; administrative training; industrial safety training; specialized skills training; on-the-job training; trainee evaluation; continuing training; training effectiveness evaluation; and program records. Two appendices describe Fundamentals training and Process operations. This handbook covers chemical operators in transportation of fuels and wastes, spent fuel receiving and storage, fuel disassembly, fuel reprocessing, and both liquid and solid low-level waste processing.

  4. Pyrochemical treatment of Idaho Chemical Processing Plant high-level waste calcine

    SciTech Connect

    Todd, T.A.; DelDebbio, J.A.; Nelson, L.O.; Sharpsten, M.R.

    1993-06-01

    The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Engineering Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1951 to recover uranium, krypton-85, and isolated fission products for interim treatment and immobilization. The acidic radioactive high-level liquid waste (HLLW) is routinely stored in stainless steel tanks and then, since 1963, calcined to form a dry granular solid. The resulting high-level waste (HLW) calcine is stored in seismically hardened stainless steel bins that are housed in underground concrete vaults. A research and development program has been established to determine the feasibility of treating ICPP HLW calcine using pyrochemical technology.This technology is described.

  5. Consolidated Fuel Reprocessing Program. Progress report for period, April 1-June 30, 1985

    SciTech Connect

    Not Available

    1985-08-01

    All research and development on civilian power reactor fuel reprocessing in the United States is managed under the Consolidated Fuel Reprocessing Program (CFRP) centered at Oak Ridge National Laboratory (ORNL). Technical progress is reported in overview fashion for the following: (1) process and engineering R and D; (2) engineering systems; (3) integrated equipment test facility operations; (4) strategic planning and analysis; (5) breeder reprocessing engineering test project; and HTGR fuel reprocessing.

  6. Health Care: Reprocessed Medical Single-Use Devices in DoD

    DTIC Science & Technology

    2002-09-30

    to handle devices without single-use or multiple-use designations, how MTFs could determine the economic viability of SUDs reprocessing, or...application approvals, the military health system should be able to realize and quantify potential monetary benefits. Economic Viability of Reprocessing...In developing a DoD reprocessing policy and when determining the economic viability of reprocessing at individual MTFs, 10 several issues

  7. 9 CFR 114.18 - Reprocessing of biological products.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    .... 114.18 Section 114.18 Animals and Animal Products ANIMAL AND PLANT HEALTH INSPECTION SERVICE... shall accurately reflect the action taken. (d) Test samples of the reprocessed serial and test reports for all tests conducted shall be submitted to Animal and Plant Health Inspection Service. The licensee...

  8. 9 CFR 114.18 - Reprocessing of biological products.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    .... 114.18 Section 114.18 Animals and Animal Products ANIMAL AND PLANT HEALTH INSPECTION SERVICE... shall accurately reflect the action taken. (d) Test samples of the reprocessed serial and test reports for all tests conducted shall be submitted to Animal and Plant Health Inspection Service. The licensee...

  9. Eye Movement Desensitization and Reprocessing: A Critical Analysis.

    ERIC Educational Resources Information Center

    Erwin, Terry McVannel

    Since Shapiro's introduction of Eye Movement Desensitization and Reprocessing (EMDR) in 1989, it has been a highly controversial therapeutic technique. Critical reviews of Shapiro's initial study have highlighted many methodological shortcomings in her work. And early empirical research that followed Shapiro's original study has been criticized…

  10. Preliminary Performance Assessment for the Waste Management Area C at the Hanford Site in Southeast Washington

    SciTech Connect

    Bergeron, Marcel P.; Singleton, Kristin M.; Eberlein, Susan J.

    2015-01-07

    A performance assessment (PA) of Single-Shell Tank (SST) Waste Management Area C (WMA C) located at the U.S. Department of Energy's (DOE) Hanford Site in southeastern Washington is being conducted to satisfy the requirements of the Hanford Federal Facility Agreement and Consent Order (HFFACO), as well as other Federal requirements and State-approved closure plans and permits. The WMP C PA assesses the fate, transport, and impacts of radionuclides and hazardous chemicals within residual wastes left in tanks and ancillary equipment and facilities in their assumed closed configuration and the subsequent risks to humans into the far future. The part of the PA focused on radiological impacts is being developed to meet the requirements for a closure authorization under DOE Order 435.1 that includes a waste incidental to reprocessing determination for residual wastes remaining in tanks, ancillary equipment, and facilities. An additional part of the PA will evaluate human health and environmental impacts from hazardous chemical inventories in residual wastes remaining in WMA C tanks, ancillary equipment, and facilities needed to meet the requirements for permitted closure under RCRA.

  11. Total System Performance Assessment - Analyses for Disposal of Commercial and DOE Waste Inventories at Yucca Mountain - Input to Final Environmental Impact Statement and Site Suitability Evaluation, Rev. 00

    SciTech Connect

    NA

    2001-09-17

    This Letter Report presents the results of calculations to assess long-term performance of commercial spent nuclear fuel (CSNF), U.S. Department of Energy (DOE) spent nuclear fuel (DSNF), high-level radioactive waste (HLW), and Greater Than Class C (GTCC) radioactive waste and DOE Special Performance Assessment Required (SPAR) radioactive waste at the potential Yucca Mountain repository in Nye County Nevada with respect to the 10,000-year performance period specified in 40 CFR Part 197.30 (66 FR 32074 [DIRS 155216], p. 32134) with regard to radiation-protection standards. The EPA Final Rule 40 CFR Part 197 has three separate standards, individual-protection, human-intrusion, and groundwater-protection standards, all with a compliance timeframe of 10,000 years. These calculations evaluate the dose to receptors for each of these standards. Further, this Letter Report includes the results of simulations to the 1,000,000-year performance period described in 40 CFR Part 197.35 (66 FR 32074 [DIRS 155216], p. 32135) which calls for the calculation of the peak dose to the Reasonably Maximally Exposed Individual (RMEI) that would occur after 10,000 years and within the period of geological stability. In accordance with TSPA-SR the ''period of geologic stability'' is from zero to 1,000,000 years after repository closure. The calculations also present the 5th and 95th percentiles, and the mean and median of the set of probabilistic simulations used to evaluate various disposal scenarios.

  12. Combination and long term stability of the IGS Reprocessing campaign

    NASA Astrophysics Data System (ADS)

    Booker, David; Clarke, Peter J.; Lavallée, David A.

    2010-05-01

    During the relatively short life of the Global Positioning System (GPS) there have been several changes to the analysis procedure, leading to inhomogeneous coordinate time series. Although they have reduced systematic errors in more recent solutions, these changes have modified the apparent periodic signals observed and led to spurious discontinuities. The International GNSS Service (IGS) reprocessing campaign uses the latest operational models and techniques to reprocess the back catalogue of GPS data to produce remove inconsistencies caused by the various model changes, thus producing a homogeneous time series of station coordinates and Earth Rotation Parameters (ERPs). Weekly coordinate and ERP solutions from up to 11 reprocessing analysis centres (ACs) have been aligned to the ITRF and combined using the TANYA software in a rigorous weighted least-squares solution. Analysis of the time series of station coordinates and Helmert transformation parameters between the combined solution and the ITRF shows a at least a 50 percent improvement in the stability of the reprocessed weekly solutions compared with earlier operational products. There is a gradual decrease in the weighted root mean square coordinate difference, both between the combined weekly solutions and the ITRF and between the individual AC solutions and their weekly combination, which reaches a minimum around the end of 2005 with a slight increase thereafter. We observe clear differences in the periodicity of Helmert transformation parameters between the individual AC solutions and the combined solution, which presumably result from variations in AC processing strategy. There is a clear annual or near annual periodic variation in the scale difference between the combined solution and the ITRF05 and some less clear variation between the translation parameters, which needs further analysis as to its cause. Keywords: GPS, ITRF, IGS reprocessing campaign, periodic errors

  13. Combined orbits and clocks from the IGS 2nd reprocessing

    NASA Astrophysics Data System (ADS)

    Griffiths, J.; Ray, J.

    2016-12-01

    In early 2015, the Analysis Centers (ACs) of the International GNSS Service (IGS) completed their second reanalysis of the full history of globally distributed GPS and GLONASS data collected since 1994. The suite of reprocessed AC solutions includes daily product files containing station positions, Earth rotation parameters, satellite orbits and clocks. This second reprocessing—or repro2—provided the IGS contribution to ITRF2014; it follows the successful first reprocessing, which provided the IGS input for ITRF2008. For this poster, we will discuss the newly combined repro2 GPS orbits and clocks. We also revisit our previous analysis of orbit day-boundary discontinuities with several significant changes and improvements: 1) Orbit discontinuities for the contributing ACs were studied in addition to those for the IGS repro2 combined orbits. (2) Apart from homogeneous reprocessing with updated analysis models, the main difference compared to the IGS Final operational products is that NOAA/NGS inputs were not submitted for the IGS reprocessing, yet they contribute heavily in the operational orbits in recent years. (3) Also, during spring 2016, the ESA modified their orbit model so that it is no longer consistent with the one used for reprocessing. A much longer span of orbits was available now, up to 11.2 years for some individual satellites, which allows a far better resolution of spectral features. 4) The procedure to compute orbit discontinuities has been further refined to account for extrapolation edge effects, improved geopotential fields, and to allow for spectral analysis of a longer time series of jumps. The satellite position time series used are complete enough that linear interpolation is necessary for only sparse gaps. So the key results are based on standard FFT power spectra (stacked over the available constellation and lightly smoothed). However, we have also computed Lomb-Scargle periodgrams to provide higher frequency resolution of some spectral

  14. Reprocessing VIIRS sensor data records from the early SNPP mission

    NASA Astrophysics Data System (ADS)

    Blonski, Slawomir; Cao, Changyong

    2016-10-01

    The Visible-Infrared Imaging Radiometer Suite (VIIRS) instrument onboard the Suomi National Polar-orbiting Partnership (SNPP) satellite began acquiring Earth observations in November 2011. VIIRS data from all spectral bands became available three months after launch when all infrared-band detectors were cooled down to operational temperature. Before that, VIIRS sensor data record (SDR) products were successfully generated for the visible and near infrared (VNIR) bands. Although VIIRS calibration has been significantly improved through the four years of the SNPP mission, SDR reprocessing for this early mission phase has yet to be performed. Despite a rapid decrease in the telescope throughput that occurred during the first few months on orbit, calibration coefficients for the VNIR bands were recently successfully generated using an automated procedure that is currently deployed in the operational SDR production system. The reanalyzed coefficients were derived from measurements collected during solar calibration events that occur on every SNPP orbit since the beginning of the mission. The new coefficients can be further used to reprocess the VIIRS SDR products. In this study, they are applied to reprocess VIIRS data acquired over pseudo-invariant calibration sites Libya 4 and Sudan 1 in Sahara between November 2011 and February 2012. Comparison of the reprocessed SDR products with the original ones demonstrates improvements in the VIIRS calibration provided by the reprocessing. Since SNPP is the first satellite in a series that will form the Joint Polar Satellite System (JPSS), calibration methods developed for the SNPP VIIRS will also apply to the future JPSS measurements.

  15. Flowsheet Analysis of U-Pu Co-Crystallization Process as a New Reprocessing System

    SciTech Connect

    Shunji Homma; Jun-ichi Ishii; Jiro Koga; Shiro Matsumoto; Toshiaki Kikuchi; Takahiro Chikazawa; Atsuhiro Shibata

    2006-07-01

    A new fuel reprocessing system by U-Pu co-crystallization process is proposed and examined by flowsheet analysis. This reprocessing system is based on the fact that hexavalent plutonium in nitric acid solution is co-crystallized with uranyl nitrate, whereas it is not crystallized when uranyl nitrate does not exist in the solution. The system consists of five steps: dissolution of spent fuel, plutonium oxidation, U-Pu co-crystallization as a co-decontamination, re-dissolution of the crystals, and U re-crystallization as a U-Pu separation. The system requires a recycling of the mother liquor from the U-Pu co-crystallization step and the appropriate recycle ratio is determined by flowsheet analysis such that the satisfactory decontamination is achieved. Further flowsheet study using four different compositions of LWR spent fuels demonstrates that the constant ratio of plutonium to uranium in mother liquor from the re-crystallization step is achieved for every composition by controlling the temperature. It is also demonstrated by comparing to the Purex process that the size of the plant based on the proposed system is significantly reduced. (authors)

  16. Development of biological and chemical methods for environmental monitoring of DOE waste disposal and storage facilities. Final report

    SciTech Connect

    1989-04-01

    Hazardous chemicals in the environment have received ever increasing attention in recent years. In response to ongoing problems with hazardous waste management, Congress enacted the Resource Conservation and Recovery Act (RCRA) in 1976. In 1980, Congress adopted the Comprehensive Environmental Response Compensation, and Liability Act (CERCLA), commonly called Superfund to provide for emergency spill response and to clean up closed or inactive hazardous waste sites. Scientists and engineers have begun to respond to the hazardous waste challenge with research and development on treatment of waste streams as well as cleanup of polluted areas. The magnitude of the problem is just now beginning to be understood. The U.S. Environmental Protection Agency (USEPA) National Priorities List as of September 13 1985, contained 318 proposed sites and 541 final sites (USEPA, 1985). Estimates of up to 30,000 sites containing hazardous wastes (1,200 to 2,000 of which present a serious threat to public health) have been made (Public Law 96-150). In addition to the large number of sites, the costs of cleanup using available technology are phenomenal. For example, a 10-acre toxic waste site in Ohio is to be cleaned up by removing chemicals from the site and treating the contaminated groundwater. The federal government has already spent more than $7 million to remove the most hazardous wastes and the groundwater decontamination alone is expected to take at least 10 years and cost $12 million. Another example of cleanup costs comes from the State of California Commission for Economic Development which predicts a bright economic future for the state except for the potential outlay of $40 billion for hazardous waste cleanup mandated by federal and state laws.

  17. Recovery of transplutonium elements from nuclear reactor waste

    DOEpatents

    Campbell, David O.; Buxton, Samuel R.

    1977-05-24

    A method of separating actinide values from nitric acid waste solutions resulting from reprocessing of irradiated nuclear fuels comprises oxalate precipitation of the major portion of actinide and lanthanide values to provide a trivalent fraction suitable for subsequent actinide/lanthanide partition, exchange of actinide and lanthanide values in the supernate onto a suitable cation exchange resin to provide an intermediate-lived raffinate waste stream substantially free of actinides, and elution of the actinide values from the exchange resin. The eluate is then used to dissolve the trivalent oxalate fraction prior to actinide/lanthanide partition or may be combined with the reprocessing waste stream and recycled.

  18. Progress in evaluation of radionuclide geochemical information developed by DOE high-level nuclear waste repository site projects: report for January-March 1985. Volume 2

    SciTech Connect

    Kelmers, A.D.; Seeley, F.G.; Arnold, W.D.; Blencoe, J.G.; Meyer, R.E.; Jacobs, G.K.; Whatley, S.K.

    1985-09-01

    Geochemical information relevant to the retention of radionuclides by the Hanford Site (in basalt) and the Yucca Mountain site (in tuff), candidate high-level nuclear waste geologic repositories being developed by US Department of Energy (DOE) projects, is being evaluated by Oak Ridge National Laboratory (ORNL) for the US Nuclear Regulatory Commission (NRC). Our evaluation of the sorption of technetium by basalt/groundwater systems was essentially completed this quarter and the results summarized; we conclude that the experimental methodology and results reported by the DOE for the Hanford Site have not conclusively established that significant retardation of technetium migration may be provided by phases present in the basalts of the Hanford Site. We have shown that sodium boltwoodite is the saturating uranium solid phase in two basalt/groundwater systems. Because thermodynamic data are not available for sodium boltwoodite, calculated solubilities for uranium are erroneous in these systems. Results of radionuclide solubility/speciation calculations, published by the DOE for the Yucca Mountain site, were evaluated this quarter under our geochemical modeling task. We express concerns relative to the inherent limitations of such calculations. Samples of Yucca Mountain tuff and J-13 well water were received for use in our planned radionuclide sorption/solubility experiments. These Yucca Mountain materials will be used to evaluate radionuclide sorption and apparent concentration limit values published by the Nevada Nuclear Waste Storage Investigations (NNWSI) project. 40 refs., 5 figs., 16 tabs.

  19. Radioactive Waste Management Basis

    SciTech Connect

    Perkins, B K

    2009-06-03

    The purpose of this Radioactive Waste Management Basis is to describe the systematic approach for planning, executing, and evaluating the management of radioactive waste at LLNL. The implementation of this document will ensure that waste management activities at LLNL are conducted in compliance with the requirements of DOE Order 435.1, Radioactive Waste Management, and the Implementation Guide for DOE Manual 435.1-1, Radioactive Waste Management Manual. Technical justification is provided where methods for meeting the requirements of DOE Order 435.1 deviate from the DOE Manual 435.1-1 and Implementation Guide.

  20. Using a Consensus Conference to Characterize Regulatory Concerns Regarding Bioremediation of Radionuclides and Heavy Metals in Mixed Waste at DOE Sites

    SciTech Connect

    Denise Lach; Stephanie Sanford

    2006-09-01

    A consensus workshop was developed and convened with ten state regulators to characterize concerns regarding emerging bioremediation technology to be used to clean-up radionuclides and heavy metals in mixed wastes at US DOE sites. Two questions were explored: integrated questions: (1) What impact does participation in a consensus workshop have on the knowledge, attitudes, and practices of state regulators regarding bioremediation technology? (2) How effective is a consensus workshop as a strategy for eliciting and articulating regulators’ concerns regarding the use of bioremediation to clean up radionuclides and heavy metals in mixed wastes at U.S. Department of Energy Sites around the county? State regulators met together for five days over two months to learn about bioremediation technology and develop a consensus report of their recommendations regarding state regulatory concerns. In summary we found that panel members: - quickly grasped the science related to bioremediation and were able to effectively interact with scientists working on complicated issues related to the development and implementation of the technology; - are generally accepting of in situ bioremediation, but concerned about costs, implementation (e.g., institutional controls), and long-term effectiveness of the technology; - are concerned equally about technological and implementation issues; and - believed that the consensus workshop approach to learning about bioremediation was appropriate and useful. Finally, regulators wanted decision makers at US DOE to know they are willing to work with DOE regarding innovative approaches to clean-up at their sites, and consider a strong relationship between states and the DOE as critical to any effective clean-up. They do not want perceive themselves to be and do not want others to perceive them as barriers to successful clean-up at their sites.

  1. Method for calcining nuclear waste solutions containing zirconium and halides

    DOEpatents

    Newby, Billie J.

    1979-01-01

    A reduction in the quantity of gelatinous solids which are formed in aqueous zirconium-fluoride nuclear reprocessing waste solutions by calcium nitrate added to suppress halide volatility during calcination of the solution while further suppressing chloride volatility is achieved by increasing the aluminum to fluoride mole ratio in the waste solution prior to adding the calcium nitrate.

  2. Development of iodine waste forms using low-temperature sintering glass.

    SciTech Connect

    Krumhansl, James Lee; Nenoff, Tina Maria; Garino, Terry J.; Rademacher, David

    2010-06-01

    This presentation will describe our recent work on the use of low temperature-sintering glass powders mixed with either AgI or AgI-zeolite to produce a stable waste form. Radioactive iodine ({sup 129}I, half-life of 1.6 x 10{sup 7} years) is generated in the nuclear fuel cycle and is of particular concern due to its extremely long half-life and its effects on human health. As part of the DOE/NE Advanced Fuel Cycle Initiative (AFCI), the separation of {sup 129}I from spent fuel during fuel reprocessing is being studied. In the spent fuel reprocessing scheme under consideration, the iodine is released in gaseous form and collected using Ag-loaded zeolites, to form AgI. Although AgI has extremely low solubility in water, it has a relatively high vapor pressure at moderate temperatures (>550 C), thus limiting the thermal processing. Because of this, immobilization using borosilicate glass is not feasible. Therefore, a bismuth oxide-based glasses are being studied due to the low solubility of bismuth oxide in aqueous solution at pH > 7. These waste forms were processed at 500 C, where AgI volatility is low but the glass powder is able to first densify by viscous sintering and then crystallize. Since the glass is not melted, a more chemically stable glass can be used. The AgI-glass mixture was found to have high iodine leach resistance in these initial studies.

  3. 1999 Annual Report on Waste Generation and Pollution Prevention Progress as Required by DOE Order 5400.1

    SciTech Connect

    SEGALL, P.

    2000-03-01

    Hanford's missions are to safely clean-up and manage the site's legacy wastes, and to develop and deploy science and technology. Through these missions Hanford will contribute to economic diversification of the region. Hanford's environmental management or clean-up mission is to protect the health and safety of the public, workers, and the environment; control hazardous materials; and utilize the assets (people, infrastructure, and site) for other missions. Hanford's science and technology mission is to develop and deploy science and technology in the service of the nation including stewardship of the Hanford Site. Pollution Prevention is a key to the success of these missions by reducing the amount of waste to be managed and identifying/implementing cost effective waste reduction projects. Hanford's original mission, the production of nuclear materials for the nation's defense programs, lasted more than 40 years, and like most manufacturing operations, Hanford's operations generated large quantities of waste and pollution. However, the by-products from Hanford operations pose unique problems like radiation hazards, vast volumes of contaminated water and soil, and many contaminated structures including reactors, chemical plants and evaporation ponds. The clean-up activity is an immense and challenging undertaking. Including characterization and decommissioning of 149 single shell storage tanks, treating 28 double shell tanks, safely disposing of over 2,100 metric tons of spent nuclear fuel stored on site, removing numerous structures, and dealing with significant solid waste, ground water, and land restoration issues.

  4. CORAL: a stepping stone for establishing the Indian fast reactor fuel reprocessing technology

    SciTech Connect

    Venkataraman, M.; Natarajan, R.; Raj, Baldev

    2007-07-01

    The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR) spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)

  5. Scoping evaluation of the technical capabilities of DOE sites for disposal of hazardous metals in mixed low-level waste

    SciTech Connect

    Gruebel, M.M.; Waters, R.D.; Langkopf, B.S.

    1997-05-01

    A team of analysts designed and conducted a scoping evaluation to estimate the technical capabilities of fifteen Department of Energy sites for disposal of the hazardous metals in mixed low-level waste (i.e., waste that contains both low-level radioactive materials and hazardous constituents). Eight hazardous metals were evaluated: arsenic, barium, cadmium, chromium, lead, mercury, selenium, and silver. The analysis considered transport only through the groundwater pathway. The results are reported as site-specific estimates of maximum concentrations of each hazardous metal in treated mixed low-level waste that do not exceed the performance measures established for the analysis. Also reported are site-specific estimates of travel times of each hazardous metal to the point of compliance.

  6. Engineered Osmosis for Energy Efficient Separations: Optimizing Waste Heat Utilization FINAL SCIENTIFIC REPORT DOE F 241.3

    SciTech Connect

    NATHAN HANCOCK

    2013-01-13

    The purpose of this study is to design (i) a stripper system where heat is used to strip ammonia (NH{sub 3}) and carbon dioxide (CO{sub 2}) from a diluted draw solution; and (ii) a condensation or absorption system where the stripped NH{sub 3} and CO{sub 2} are captured in condensed water to form a re-concentrated draw solution. This study supports the Industrial Technologies Program of the DOE Office of Energy Efficiency and Renewable Energy and their Industrial Energy Efficiency Grand Challenge award solicitation. Results from this study show that stimulated Oasys draw solutions composed of a complex electrolyte solution associated with the dissolution of NH{sub 3} and CO{sub 2} gas in water can successfully be stripped and fully condensed under standard atmospheric pressure. Stripper bottoms NH{sub 3} concentration can reliably be reduced to < 1 mg/L, even when starting with liquids that have an NH{sub 3} mass fraction exceeding 6% to stimulate diluted draw solution from the forward osmosis membrane component of the process. Concentrated draw solution produced by fully condensing the stripper tops was show to exceed 6 M-C with nitrogen-to-carbon (N:C) molar ratios on the order of two. Reducing the operating pressure of the stripper column serves to reduce the partial vapor pressure of both NH{sub 3} and CO{sub 2} in solution and enables lower temperature operation towards integration of industrial low-grade of waste heat. Effective stripping of solutes was observed with operating pressures as low as 100 mbar (3-inHg). Systems operating at reduced pressure and temperature require additional design considerations to fully condense and absorb these constituents for reuse within the Oasys EO system context. Comparing empirical data with process stimulation models confirmed that several key parameters related to vapor-liquid equilibrium and intrinsic material properties were not accurate. Additional experiments and refinement of material property databases within the

  7. View graph presentations of the sixth DOE industry/university/lab forum on robotics for environmental restoration and waste management

    SciTech Connect

    1995-10-01

    The mission of the Robotics Technology Development Program involves the following: develop robotic systems where justified by safety, cost, and/or efficiency arguments; integrate the best talent from National Labs, industry, and universities in focused teams addressing complex-wide problems; and involve customers in the identification and development of needs driven technologies. This presentation focuses on five areas. They are: radioactive tank waste remediation (Richland); mixed waste characterization, treatment, and disposal (Idaho Falls); decontamination and decommissioning (Morgantown); landfill stabilization (Savannah River); and contaminant plumes containment and remediation (Savannah River).

  8. Reprocessing of LiH in Molten Chlorides

    NASA Astrophysics Data System (ADS)

    Masset, Patrick J.; Gabriel, Armand; Poignet, Jean-Claude

    2008-06-01

    LiH was used as inactive material to stimulate the reprocessing of lithium tritiate in molten chlorides. The electrochemical properties (diffusion coefficients, apparent standard potentials) were measured by means of transient electrochemical techniques (cyclic voltammetry and chronopotentiometry). At 425 ºC the diffusion coefficient and the apparent standard potential were 2.5 · 10-5 cm2 s-1 and -1.8 V vs. Ag/AgCl, respectively. For the process design the LiH solubility was measured by means of DTA to optimize the LiH concentration in the molten phase. In addition electrolysis tests were carried out at 460 ºC with current densities up to 1 A cm-2 over 24 h. These results show that LiH may be reprocessed in molten chlorides consisting in the production of hydrogen gas at the anode and molten metallic lithium at the cathode.

  9. On the Reprocessing and Reanalysis of Observations for Climate

    NASA Technical Reports Server (NTRS)

    Bosilovich, Michael G.; Kennedy, John; Dee, Dick; ONeill, Alan

    2012-01-01

    The long observational record is critical to our understanding of the Earth s climate, but most observing systems were not developed with a climate objective in mind. As a result, tremendous efforts have gone into assessing and reprocessing the data records to improve their usefulness in climate studies. Many challenges remain, such as tracking the improvement of processing algorithms and limited spatial coverage. Reanalyses have fostered significant research, yet reliable global trends in many physical fields are not yet attainable, despite significant advances in data assimilation and numerical modeling. Communication of the strengths, limitations and uncertainties of reprocessed observations and reanalysis data, not only among the community of developers, but also with the extended research community, including the new generations of researchers and the decision makers is crucial for further advancement of the observational data records. WCRP provides the means to bridge the different motivating objectives on which national efforts focus.

  10. ATLAS level-1 calorimeter trigger: Monitoring and data reprocessing

    NASA Astrophysics Data System (ADS)

    Dimond, David; Hong, Tae; Carlson, Benjamin; Atlas Collaboration

    2017-01-01

    We present the monitoring and data reprocessing for the calorimeter-based hardware level-1 trigger system (L1Calo) for the ATLAS experiment. This trigger system was upgraded after the Run-1 data taking period (2009-2012) to prepare for Run-2 (2015-current), which allowed better control the event rates for algorithms based on jets and/or missing energy. Monitoring tools for the upgraded system is described. We also present a new offline tool to reprocess previous data samples with altered L1Calo settings, such as calibration constants and noise cuts. The samples are used to study the dependence of the event rates and signal efficiencies on the settings. The studies can help plan the appropriate L1Calo settings for upcoming data taking periods as well as for future runs.

  11. Principles of infection prevention and reprocessing in ENT endoscopy

    PubMed Central

    Kramer, Axel; Kohnen, Wolfgang; Israel, Susanne; Ryll, Sylvia; Hübner, Nils-Olaf; Luckhaupt, Horst; Hosemann, Werner

    2015-01-01

    This article gives an overview on the principles of reprocessing of rigid and flexible endoscopes used in ENT units including structural and spatial requirements based on general and ENT-specific risks of infection associated with diagnostic and therapeutic endoscopy. The underlying legal principles as well as recommendations from scientific societies will be exemplified in order to give a practical guidance to the otorhinolaryngologist. Preliminary results of a small nation-wide survey on infection control standards based on data of 29 ENT practices in Germany reveal current deficits of varying degree concerning infection control management including reprocessing of endoscopes. The presented review aims to give support to the establishment of a structured infection control management program including the evaluation of results by means of a prospective surveillance. PMID:26770284

  12. The sterilization efficacy of reprocessed single use diathermy pencils.

    PubMed

    Batista Neto, Simone; Graziano, Kazuko Uchikawa; Padoveze, Maria Clara; Kawagoe, Júlia Yaeko

    2010-01-01

    In Brazil, single use diathermy pencils (SUDP) are among the most common reused devices. This study assesses the sterilization efficacy of reprocessing SUDP using two cleansing methods (manual or automated), followed by one of three of the low-temperature sterilization methods: Hydrogen Peroxide Plasma (HPP), Ethylene Oxide (ETO) or Low-Temperature Steam Formaldehyde (LTSF). The sample was composed of 360 SUDP after their first use. The probability of sterilization failure was estimated considering the number of positive microbiological results obtained by cultures of the studied devices. The overall sterilization failure probability for SUDP was 0.26. The sterilization method, which presented the lowest failure probability was the LTSF (0.01), followed by ETO (0.21) and HPP (0.56). Automated cleansing obtained a better result than manual cleansing. This trial demonstrated that the probability of sterilization in reprocessed SUDP is highly dependent on both the type of cleansing and the sterilization method applied.

  13. Candidate waste forms for immobilisation of waste chloride salt from pyroprocessing of spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Vance, E. R.; Davis, J.; Olufson, K.; Chironi, I.; Karatchevtseva, I.; Farnan, I.

    2012-01-01

    Sodalite/glass bodies prepared by hot isostatic pressing (HIPing) at ˜850 °C/100 MPa are candidates for immobilising fission product-bearing waste KCl-LiCl pyroprocessing salts. To study the capacity of sodalite to structurally incorporate such pyroprocessing salts, K, Li, Cs, Sr, Ba and La were individually targeted for substitution in a Na site in sodalite (Na vacancies targeted as charge compensators for alkaline and rare earths) and studied by X-ray diffraction and scanning electron microscopy after sintering in the range of 800-1000 °C. K and Li appeared to enter the sodalite, but Cs, Sr and Ba formed aluminosilicate phases and La formed an oxyapatite phase. However these non-sodalite phases have reasonable resistance to water leaching. Pure chlorapatite gives superior leach resistance to sodalite, and alkalis, alkaline and rare earth ions are generally known to enter chlorapatite, but attempts to incorporate simulated waste salt formulations into HIPed chlorapatite-based preparations or to substitute Cs alone into the structure of Ca-based chlorapatite were not successful on the basis of scanning electron microscopy. The materials exhibited severe water leachability, mainly in regard to Cs release. Attempts to substitute Cs into Ba- and Sr-based chlorapatites also did not look encouraging. Consequently the use of apatite alone to retain fission product-bearing waste pyroprocessing salts from electrolytic nuclear fuel reprocessing is problematical, but chlorapatite glass-ceramics may be feasible, albeit with reduced waste loadings. Spodiosite, Ca 2(PO 4)Cl, does not appear to be suitable for incorporation of Cl-bearing waste containing fission products.

  14. Process centrifuge operating problems and equipment failures in canyon reprocessing facilities at the Savannah River Site

    SciTech Connect

    Durant, W.S.; Baughman, D.F.

    1990-03-01

    The Savannah River Laboratory (SRL) maintains a compilation of operating problems and equipment failures that have occurred in the fuel reprocessing areas of the Savannah River Site (SRS). At present, the data bank contains more than 230,000 entries ranging from minor equipment malfunctions to incidents with the potential for injury or contamination of personnel, or for economic loss. The data bank has been used extensively for a wide variety of purposes, such as failure analyses, trend analyses, and preparation of safety analyses. Typical of the data are problems associated with the canyon process centrifuges. This report contains a compilation of the centrifuge operating problems and equipment failures primarily as an aid to organizations with related equipment. Publication of these data was prompted by a number of requests for this information by other Department of Energy (DOE) sites. 11 refs., 2 figs., 4 tabs.

  15. Process agitator operating problems and equipment failures, F-Canyon Reprocessing Facility

    SciTech Connect

    Durant, W.S.; Starks, J.B.; Low, J.M.; Galloway, W.D.

    1988-09-01

    The Savannah River Laboratory (SRL) maintains a compilation of operating problems and equipment failures that have occurred in the fuel reprocessing areas of the Savannah River Plant (SRP). At present, the data bank contains more than 200,000 entries ranging from minor equipment malfunctions to incidents with the potential for injury or contamination of personnel, or for economic loss. The data bank has been used extensively for a wide variety of purposes, such as failure analyses, trend analyses, and preparation of safety analyses. Typical of the data are problems associated with the F-Canyon process agitators. This report contains a compilation of the agitator operating problems and equipment failures primarily as an aid to organizations with related equipment. Publication of these data was prompted by a number of requests for this information by other Department of Energy (DOE) sites. 4 figs., 4 tabs.

  16. On the Reprocessing and Reanalysis of Observations for Climate

    NASA Technical Reports Server (NTRS)

    Bosilovich, Michael G.; Kennedy, John; Dee, Dick; Allan, R.; O'Neill, Alan

    2013-01-01

    The long observational record is critical to our understanding of the Earths climate, but most observing systems were not developed with a climate objective in mind. As a result, tremendous efforts have gone into assessing and reprocessing the data records to improve their usefulness in climate studies. The purpose of this paper is to both review recent progress in reprocessing and reanalyzing observations, and to summarize the challenges that must be overcome in order to improve our understanding of climate and variability. Reprocessing improves data quality through more scrutiny and improved retrieval techniques for individual observing systems, while reanalysis merges many disparate observations with models through data assimilation, yet both aim to provide an climatology of Earth processes. Many challenges remain, such as tracking the improvement of processing algorithms and limited spatial coverage. Reanalyses have fostered significant research, yet reliable global trends in many physical fields are not yet attainable, despite significant advances in data assimilation and numerical modeling. Oceanic reanalyses have made significant advances in recent years, but will only be discussed here in terms of progress toward integrated Earth system analyses. Climate data sets are generally adequate for process studies and large-scale climate variability. Communication of the strengths, limitations and uncertainties of reprocessed observations and reanalysis data, not only among the community of developers, but also with the extended research community, including the new generations of researchers and the decision makers is crucial for further advancement of the observational data records. It must be emphasized that careful investigation of the data and processing methods are required to use the observations appropriately.

  17. Legal implications of single-use medical device reprocessing.

    PubMed

    Larose, Emily

    2013-01-01

    Over 10 years ago, the Public Health Agency of Canada released the results of a nation-wide survey of hospitals that demonstrated that the reuse of single-use medical devices was widespread in Canadian healthcare institutions. In this article, the author discusses the reuse and reprocessing of these devices, as well as the risks this practice presents. She then goes on to outline the legal implications of reusing single-use devices.

  18. Reprocessing of metallurgical slag into materials for the building industry

    SciTech Connect

    Pioro, L.S.; Pioro, I.L

    2004-07-01

    Several methods of reprocessing metallurgical (blast furnace) slag into materials for the building industry, based on melting aggregates with submerged combustion, were developed and tested. The first method involves melting hot slag with some additives directly in a slag ladle with a submerged gas-air burner, with the objective of producing stabilized slag or glass-ceramic. The second method involves direct draining of melted slag from a ladle into the slag receiver, with subsequent control of the slag draining into the converter where special charging materials are added to the melt, with the objective of producing glass-ceramic. A third method involves melting cold slag with some additives inside a melting converter with submerged gas-air burners, with the objective of producing glass-ceramic fillers for use in road construction. Specific to the melting process is the use of a gas-air mixture with direct combustion inside the melt. This feature provides melt bubbling to help achieve maximum heat transfer from combustion products to the melt, improve mixing (and therefore homogeneity of the melt), and increases the rate of chemical reactions. The experimental data for different aspects of the proposed methods are presented. The reprocessed blast-furnace slag in the form of granules can be used as fillers for concretes, asphalts, and as additives in the production of cement, bricks and other building materials. As well, reprocessed blast-furnace slag can be poured into forms for the production of glass-ceramic tiles.

  19. Methodology for estimating reprocessing costs for nuclear fuels

    SciTech Connect

    Carter, W. L.; Rainey, R. H.

    1980-02-01

    A technological and economic evaluation of reprocessing requirements for alternate fuel cycles requires a common assessment method and a common basis to which various cycles can be related. A methodology is described for the assessment of alternate fuel cycles utilizing a side-by-side comparison of functional flow diagrams of major areas of the reprocessing plant with corresponding diagrams of the well-developed Purex process as installed in the Barnwell Nuclear Fuel Plant (BNFP). The BNFP treats 1500 metric tons of uranium per year (MTU/yr). Complexity and capacity factors are determined for adjusting the estimated facility and equipment costs of BNFP to determine the corresponding costs for the alternate fuel cycle. Costs of capacities other than the reference 1500 MT of heavy metal per year are estimated by the use of scaling factors. Unit costs of reprocessed fuel are calculated using a discounted cash flow analysis for three economic bases to show the effect of low-risk, typical, and high-risk financing methods.

  20. Noble Gas Measurement and Analysis Technique for Monitoring Reprocessing Facilities

    SciTech Connect

    Charlton, William S

    1999-09-01

    An environmental monitoring technique using analysis of stable noble gas isotopic ratios on-stack at a reprocessing facility was developed. This technique integrates existing technologies to strengthen safeguards at reprocessing facilities. The isotopic ratios are measured using a mass spectrometry system and are compared to a database of calculated isotopic ratios using a Bayesian data analysis method to determine specific fuel parameters (e.g., burnup, fuel type, fuel age, etc.). These inferred parameters can be used by investigators to verify operator declarations. A user-friendly software application (named NOVA) was developed for the application of this technique. NOVA included a Visual Basic user interface coupling a Bayesian data analysis procedure to a reactor physics database (calculated using the Monteburns 3.01 code system). The integrated system (mass spectrometry, reactor modeling, and data analysis) was validated using on-stack measurements during the reprocessing of target fuel from a U.S. production reactor and gas samples from the processing of EBR-II fast breeder reactor driver fuel. These measurements led to an inferred burnup that matched the declared burnup with sufficient accuracy and consistency for most safeguards applications. The NOVA code was also tested using numerous light water reactor measurements from the literature. NOVA was capable of accurately determining spent fuel type, burnup, and fuel age for these experimental results. Work should continue to demonstrate the robustness of this system for production, power, and research reactor fuels.

  1. Concept for dismantling the Hllw treatment facility on the Former Wak Reprocessing Site

    SciTech Connect

    Birringer, K.J.; Fleisch, J.; Graffunder, I.; Pfeifer, W.

    2007-07-01

    The German pilot reprocessing plant WAK was operated until 1990 and processed about 200 tons of nuclear fuels from test and power reactors. In late 1991, the Federal Republic of Germany, the State of Baden-Wuerttemberg, and the utilities decided to shut down the WAK and to dismantle it completely to the green field. In the years 2000/2001, remote-controlled dismantling of the process cells in the reprocessing building was completed. Part of the building has already been subjected to release measurement and released from the obligations under the German Atomic Energy Act. However, a major prerequisite for the complete dismantling of the WAK is the management of the 60 m{sup 3} high-level liquid waste (HLLW) with an activity of 8.0 E 17 Bq resulting from reprocessing. For this purpose, the Karlsruhe vitrification plant (VEK) was constructed and is now under commissioning /1/. Hot operation is foreseen for the years 2007/2008. Following vitrification operation, dismantling of the four HLLW tanks in the storage building will be a particularly challenging task in terms of radiology. The HLLW tanks are located in thick-walled concrete cells that require remote- controlled horizontal access. For this purpose, a new access building, the southern extension, was built. It serves to bring in and operate the remote handling tools and allows for the contamination-safe removal and measurement of the MAW drums. In contrast to the crane in the process building, the manipulator carrier system used here is an 8 Mg excavator. All tools, including the wall cutter, chisel, cutting disk, scissors, and the electric master-slave manipulator (EMSM), can be docked to this excavator. The VEK installations shall be dismantled parallel to the HLLW storage tanks. Due to the dose rates expected after operation, two dismantling areas have to be distinguished in the VEK: The core area with the HLLW transfer cell, melter cell, and exhaust gas cell requires remote dismantling. All remaining cells

  2. Progress and experiences from the decommissioning of the Eurochemic reprocessing plant

    SciTech Connect

    Gills, R.; Lewandowski, P.; Ooms, B.; Reusen, N.; Van Laer, W.; Walthery, R.

    2007-07-01

    Belgoprocess started the industrial decommissioning of the main process building of the former EUROCHEMIC reprocessing plant in 1990, after completion of a pilot project in which two buildings were emptied and decontaminated to background levels. The remaining structures were demolished and the concrete debris was disposed of as industrial waste and green field conditions restored. The Eurochemic reprocessing plant operated from 1966 to 1974 to process fuel from power reactors and research reactors. The main building is a large concrete structure, comprising a surface area of 55,000 m{sup 2}, concrete volume 12,500 m{sup 3}, and 1,500 Mg of metal components. The building is divided into multiple cells. About 106 individual cell structures have to be dismantled, involving the removal and decontamination of equipment from each cell, the decontamination of the cell walls, ceilings and floors, the dismantling of the ventilation system. Most of the work involves hands-on operations under protective clothing tailored to each specific task. Tool automation and automatic positioning systems are successfully applied. In view of the final demolition of the main process building, the main process building is divided into three parts - each part is isolated from the others. In the middle of 2008, after the removal of the NDA-IPAN/GEA installation, the eastern part will be demolished. The paper presents a status overview of the decommissioning and decontamination activities at the main process building of the former Eurochemic reprocessing plant on the nuclear site of Dessel in Belgium. The specific BELGOPROCESS approach will be highlighted, in which the decommissioning activities are carried out on an industrial scale with special emphasis on cost minimisation, the use of technology on an industrial representative scale and the specific alpha contamination of equipment and building surfaces, requiring that the decommissioning work is done with adequate protective clothing

  3. Recovery of iron, carbon and zinc from steel plant waste oxides using the AISI-DOE postcombustion smelting technology

    SciTech Connect

    Sarma, B.; Downing, K.B.; Aukrust, E.

    1996-09-01

    This report describes a process to recover steel plant waste oxides to be used in the production of hot metal. The process flowsheet used at the pilot plant. Coal/coke breeze and iron ore pellets/waste oxides are charged into the smelting reactor. The waste oxides are either agglomerated into briquettes (1 inch) using a binder or micro-agglomerated into pellets (1/4 inch) without the use of a binder. The iron oxides dissolve in the slag and are reduced by carbon to produce molten iron. The gangue oxides present in the raw materials report to the slag. Coal charged to the smelter is both the fuel as well as the reductant. Carbon present in the waste oxides is also used as the fuel/reductant resulting in a decrease in the coal requirement. Oxygen is top blown through a central, water-cooled, dual circuit lance. Nitrogen is injected through tuyeres at the bottom of the reactor for stirring purposes. The hot metal and slag produced in the smelting reactor are tapped at regular intervals through a single taphole using a mudgun and drill system. The energy requirements of the process are provided by (i) the combustion of carbon to carbon monoxide, referred to as primary combustion and (ii) the combustion of CO and H{sub 2} to CO{sub 2} and H{sub 2}O, known as postcombustion.

  4. Treatment of air pollution control residues with iron rich waste sulfuric acid: does it work for antimony (Sb)?

    PubMed

    Okkenhaug, Gudny; Breedveld, Gijs D; Kirkeng, Terje; Lægreid, Marit; Mæhlum, Trond; Mulder, Jan

    2013-03-15

    Antimony (Sb) in air pollution control (APC) residues from municipal solid waste incineration has gained increased focus due to strict Sb leaching limits set by the EU landfill directive. Here we study the chemical speciation and solubility of Sb at the APC treatment facility NOAH Langøya (Norway), where iron (Fe)-rich sulfuric acid (∼3.6M, 2.3% Fe(II)), a waste product from the industrial extraction of ilmenite, is used for neutralization. Antimony in water extracts of untreated APC residues occurred exclusively as pentavalent antimonate, even at low pH and Eh values. The Sb solubility increased substantially at pH<10, possibly due to the dissolution of ettringite (at alkaline pH) or calcium (Ca)-antimonate. Treated APC residues, stored anoxically in the laboratory, simulating the conditions at the NOAH Langøya landfill, gave rise to decreasing concentrations of Sb in porewater, occurring exclusively as Sb(V). Concentrations of Sb decreased from 87-918μgL(-1) (day 3) to 18-69μgL(-1) (day 600). We hypothesize that an initial sorption of Sb to Fe(II)-Fe(III) hydroxides (green rust) and eventually precipitation of Ca- and Fe-antimonates (tripuhyite; FeSbO4) occurred. We conclude that Fe-rich, sulfuric acid waste is efficient to immobilize Sb in APC residues from waste incineration. Copyright © 2013 Elsevier B.V. All rights reserved.

  5. Advanced reprocessing developments in Europe contribution of European projects ACSEPT and ACTINET-I3

    SciTech Connect

    Bourg, S.; Poinssot, C.; Geist, A.; Cassayre, L.; Rhodes, C.; Ekberg, C.

    2012-07-01

    Nuclear energy has more than ever to demonstrate that it can contribute safely and on a sustainable way to answer the international increase in energy needs. Actually, in addition to an increased safety of the reactors themselves, its acceptance is still closely associated to our capability to reduce the lifetime of the nuclear waste, to manage them safely and to propose options for a better use of the natural resources. Spent fuel reprocessing can help to reach these objectives. But this cannot be achieved only by optimizing industrial processes through engineering studies. It is of a primary importance to increase our fundamental knowledge in actinide sciences in order to build the future of nuclear energy on reliable and scientifically-founded results, and therefore meet the needs of the future fuel cycles in terms of fabrication and performance of fuels, reprocessing and waste management. At the European level, both the collaborative project ACSEPT and the Integrated Infrastructure Initiative ACTINET-I3 work together to improve our knowledge in actinides chemistry and therefore develop advanced separation processes. These tools are complementary and work in close connection on some specific issues such as the understanding of the selectivity of extracting organic ligands. By offering trans-national access to the main nuclear research facility in Europe, ACTINET-I3 aims at increasing the knowledge in actinide sciences by gathering all the expertise available in European nuclear research institutes or university and giving them the opportunity to come and work in hot-labs (ITU, Atalante...) or beamlines (ESFR, ANKA, PSI) ACSEPT is focused on the development of advanced separation processes, both aqueous and pyrochemical. Head-end steps, fuel re-fabrication, solvent treatment, waste management are also taken into account. In aqueous process development, the SANEX and innovative SANEX flowsheets demonstration were successfully achieved. Chemical systems were

  6. Control of radio-iodine at the German reprocessing plant WAK during operation and after shutdown

    SciTech Connect

    Herrmann, F.J.; Herrmann, B.; Kuhn, K.D.

    1997-08-01

    During 20 years of operation 207 metric tons of oxide fuel from nuclear power reactors with 19 kg of iodine-129 had been reprocessed in the WAK plant near Karlsruhe. In January 1991 the WAK Plant was shut down. During operation iodine releases of the plant as well as the iodine distribution over the liquid and gaseous process streams had been determined. Most of the iodine is evolved into the dissolver off-gas in volatile form. The remainder is dispersed over many aqueous, organic and especially gaseous process and waste streams. After shut down of the plant in January 1991, iodine measurements in the off-gas streams have been continued up to now. Whereas the iodine-129 concentration in the dissolver off-gas dropped during six months after shutdown by three orders of magnitude, the iodine concentrations in the vessel ventilation system of the PUREX process and the cell vent system decreased only by a factor of 10 during the same period. Iodine-129 releases of the liquid high active waste storage tanks did not decrease distinctly. The removal efficiencies of the silver impregnated iodine filters in the different off-gas streams of the WAK plant depend on the iodine concentration in the off-gas. The reason of the observed dependence of the DF on the iodine-129 concentration might be due to the presence of organic iodine compounds which are difficult to remove. 13 refs., 3 figs.

  7. Clean tailing reclamation: Tailing reprocessing for sulfide removal and vegetation establishment

    SciTech Connect

    Jennings, S.R.; Kruegar, J.

    1997-12-31

    Mine wastes exhibiting elevated heavy metal concentrations are widespread causes of resource degradation in the western US and elsewhere. This problem is further exacerbated by the presence of pyrite that oxidizes upon exposure to the atmosphere resulting in acid generation. Since pyrite was not recovered as a mineral of economic value during mining, it was disposed of in waste piles and tailing ponds that are now a source of acid generation and release of metals to the environment. Tailing cleaning, or sulfide mineral recovery through reprocessing, was evaluated as an innovative reclamation technology. Tailing materials, from both operational and abandoned mines, were collected to evaluate the feasibility of sulfide mineral recovery. Successful mineral separation was performed resulting in a low volume metal sulfide concentrate and a high volume cleaned silicate media. Total metal concentrations were decreased in the cleaned tailing material and elevated in the sulfide concentrate compared with the original tailing chemistry. In greenhouse trials, vegetation establishment in cleaned tailing material was compared with plant growth in topsoil and lime-amended tailings. While vegetation performance was best in the topsoil control, both lime-amended and cleaned tailings displayed adequate plant growth.

  8. Safe storage time for reprocessed flexible endoscopes: a systematic review.

    PubMed

    Schmelzer, Marilee; Daniels, Glenda; Hough, Helen

    2015-09-01

    Flexible endoscopes are used to diagnose and treat gastrointestinal and pulmonary diseases. They have narrow, internal channels which are used to insert instruments, air and water into hollow organs and to remove tissues and secretions. Since endoscopes are contaminated during use and have heat sensitive components that cannot be sterilized, they are reprocessed with cleaning and high-level disinfection to destroy microorganisms. Knowing how long reprocessed endoscopes can be safely stored is essential for preventing infection and decreasing unnecessary costs. The objective was to systematically review the best available evidence related to safe storage time (in days) of flexible endoscopes that have undergone reprocessing in order to determine when they can be safely used again without posing any risk of contamination from pathogens. The types of equipment were flexible endoscopes that had been reprocessed according to recommended guidelines, stored for a specified period of time, and tested for microorganisms.The intervention was the length of time (in days) that reprocessed endoscopes were appropriately stored before use.This review included non-randomized controlled trials and prospective cohort studies.This review considered studies that included the outcome measure: microbial growth on endoscopes which was measured with microbiological cultures. The search strategy aimed to find studies in English and Spanish and included published and unpublished studies from 1990 to 2014. An initial search of CINAHL, MEDLINE/PUBMed and EMBASE was conducted followed by an analysis of the text words contained in the title and abstract and index terms used to describe the articles. Next, a search using all identified keywords and index terms was undertaken across all included databases. Then, the reference lists of all identified reports and articles were searched for additional studies. A citation search was performed in order to find additional studies that cited those

  9. Removal efficiency of silver impregnated filter materials and performance of iodie filters in the off-gas of the Karlsruhe reprocessing plant WAK

    SciTech Connect

    Herrmann, F.J.; Herrmann, B.; Hoeflich, V.

    1997-08-01

    An almost quantitative retention of iodine is required in reprocessing plants. For the iodine removal in the off-gas streams of a reprocessing plant various sorption materials had been tested under realistic conditions in the Karlsruhe reprocessing plant WAK in cooperation with the Karlsruhe research center FZK. The laboratory results achieved with different iodine sorption materials justified long time performance tests in the WAK Plant. Technical iodine filters and sorption materials for measurements of iodine had been tested from 1972 through 1992. This paper gives an overview over the most important results, Extended laboratory, pilot plant, hot cell and plant experiences have been performed concerning the behavior and the distribution of iodine-129 in chemical processing plants. In a conventional reprocessing plant for power reactor fuel, the bulk of iodine-129 and iodine-127 is evolved into the dissolver off-gas. The remainder is dispersed over many aqueous, organic and gaseous process and waste streams of the plant. Iodine filters with silver nitrate impregnated silica were installed in the dissolver off-gas of the Karlsruhe reprocessing plant WAK in 1975 and in two vessel vent systems in 1988. The aim of the Karlsruhe iodine research program was an almost quantitative evolution of the iodine during the dissolution process to remove as much iodine with the solid bed filters as possible. After shut down of the WAK plant in December 1990 the removal efficiency of the iodine filters at low iodine concentrations had been investigated during the following years. 12 refs., 2 figs., 2 tabs.

  10. Waste management progress report

    SciTech Connect

    1997-06-01

    During the Cold War era, when DOE and its predecessor agencies produced nuclear weapons and components, and conducted nuclear research, a variety of wastes were generated (both radioactive and hazardous). DOE now has the task of managing these wastes so that they are not a threat to human health and the environment. This document is the Waste Management Progress Report for the U.S. Department of Energy dated June 1997. This progress report contains a radioactive and hazardous waste inventory and waste management program mission, a section describing progress toward mission completion, mid-year 1997 accomplishments, and the future outlook for waste management.

  11. DEVELOPMENT OF VISUAL CRITERIA FOR EVALUATION OF CORRODED TRANSURANIC (TRU) WASTE DRUMS AT THE DEPARTMENT OF ENERGY (US/DOE) HANFORD SITE

    SciTech Connect

    CANNELL, G.R.

    2004-11-01

    Fluor Hanford, Inc., at the Department of Energy (DOE) Hanford Site, has recently begun retrieving some 37,000 contact-handled, suspect-Transuranic or ''Retrievably Stored Waste'' (CH-TRU) waste drums from its Low Level Burial Grounds (LLBG). The drums are being retrieved, processed and prepared for eventual shipment to the DOE Waste Isolation Pilot Plant (WIPP). Immediately upon retrieval, the drums are visually inspected against requirements identified in the facility Authorization Basis to ensure they are safe for handling and fit for on-site transfer. A number of the retrieved drums did not meet specified corrosion criteria and as such required structural evaluation by Ultrasonic Test (UT) thickness checking (including mechanical surface prep) or overpacking into a Conex-type container prior to transfer. The additional evaluation and overpacking increases personnel exposure to the radioactive waste and reduces efficiency of the retrieval process. Based on historic Hanford CH-TRU waste drum corrosion data, showing very low general corrosion rates, there was reason to believe that existing Hanford site-transfer corrosion criteria were more conservative than needed. In an effort to demonstrate this belief, a corrosion investigation was performed. Eleven CH-TRU waste drums not meeting the corrosion criteria were included in the investigation and from these, 92 separate locations, or areas of corrosion, were evaluated. Each of these locations was visually characterized and evaluated for thickness using the UT method. Visual characterization consisted of ranking photographs for each location on a scale from 1 to 6, representing an increasing level of corrosion attack. UT thickness measurements were then plotted against the visual ratings to identify any significant correlation. Analysis of the data indicated that as the corrosion rating increased, wall thickness decreased. It was concluded that drum surfaces characterized by a corrosion rating of 1-4 could be

  12. Idaho Waste Vitrification Facilities Project Vitrified Waste Interim Storage Facility

    SciTech Connect

    Bonnema, Bruce Edward

    2001-09-01

    This feasibility study report presents a draft design of the Vitrified Waste Interim Storage Facility (VWISF), which is one of three subprojects of the Idaho Waste Vitrification Facilities (IWVF) project. The primary goal of the IWVF project is to design and construct a treatment process system that will vitrify the sodium-bearing waste (SBW) to a final waste form. The project will consist of three subprojects that include the Waste Collection Tanks Facility, the Waste Vitrification Facility (WVF), and the VWISF. The Waste Collection Tanks Facility will provide for waste collection, feed mixing, and surge storage for SBW and newly generated liquid waste from ongoing operations at the Idaho Nuclear Technology and Engineering Center. The WVF will contain the vitrification process that will mix the waste with glass-forming chemicals or frit and turn the waste into glass. The VWISF will provide a shielded storage facility for the glass until the waste can be disposed at either the Waste Isolation Pilot Plant as mixed transuranic waste or at the future national geological repository as high-level waste glass, pending the outcome of a Waste Incidental to Reprocessing determination, which is currently in progress. A secondary goal is to provide a facility that can be easily modified later to accommodate storage of the vitrified high-level waste calcine. The objective of this study was to determine the feasibility of the VWISF, which would be constructed in compliance with applicable federal, state, and local laws. This project supports the Department of Energy’s Environmental Management missions of safely storing and treating radioactive wastes as well as meeting Federal Facility Compliance commitments made to the State of Idaho.

  13. Radioactive Waste Management, its Global Implication on Societies, and Political Impact

    NASA Astrophysics Data System (ADS)

    Matsui, Kazuaki

    2009-05-01

    Reprocessing plant in Rokkasho, Japan is under commissioning at the end of 2008, and it starts soon to reprocess about 800 Mt of spent fuel per annum, which have been stored at each nuclear power plant sites in Japan. Fission products together with minor actinides separated from uranium and plutonium in the spent fuel contain almost all radioactivity of it and will be vitrified with glass matrix, which then will fill the canisters. The canisters with the high level radioactive waste (HLW) are so hot in both thermal and radiological meanings that they have to be cooled off for decades before bringing out to any destination. Where is the final destination for HLW in Japan, which is located at the rim of the Pacific Ocean with volcanoes? Although geological formation in Japan is not so static and rather active as the other parts of the planet, experts concluded with some intensive studies and researches that there will be a lot of variety of geological formations even in Japan which can host the HLW for so long times of more than million years. Then an organization to implement HLW disposal program was set up and started to campaign for volunteers to accept the survey on geological suitability for HLW disposal. Some local governments wanted to apply, but were crashed down by local and neighbor governments and residents. The above development is not peculiar only to Japan, but generally speaking more or less common for those with radioactive waste programs. This is why the radioactive waste management is not any more science and technology issue but socio-political one. It does not mean further R&D on geological disposal is not any more necessary, but rather we, each of us, should face much more sincerely the societal and political issues caused by the development of the science and technology. Second topic might be how effective partitioning and transformation technology may be to reduce the burden of waste disposal and denature the waste toxicity? The third one might

  14. Central Waste Complex (CWC) Waste Analysis Plan

    SciTech Connect

    ELLEFSON, M.D.

    1999-12-01

    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 at the Central Waste Complex (CWC), which is located in the 200 West Area of the Hanford Facility, Richland, Washington. Because dangerous waste does not include the source, special nuclear, and by-product material components of mixed waste, radionuclides are not within the scope of this documentation. The information on radionuclides is provided only for general knowledge.

  15. Effect of peracetic acid reprocessing on the transport characteristics of polysulfone hemodialyzers.

    PubMed

    Wolff, Susanne H; Zydney, Andrew L

    2005-02-01

    Peracetic acid is used extensively for reprocessing hemodialyzers, despite several indications that reprocessing alters the dialyzer transport characteristics. The objective of this study was to obtain quantitative data for the effects of peracetic acid reprocessing on the clearance and sieving coefficients of urea, vitamin B12, and polydisperse dextrans using Fresenius F80A polysulfone dialyzers. Reprocessing restored the urea and vitamin B12 clearance to close to their original values. However, the reprocessed dialyzers had substantially lower clearance of the larger molecular weight dextrans, which was attributed to reductions in the effective pore size caused by residual plasma proteins within the membrane. Storage in peracetic acid provided some additional removal of residual proteins, although the clearance and sieving coefficients of the larger dextrans remained well below their original values. Peracetic acid caused no degradation of the membrane polymer, in sharp contrast to results obtained with bleach reprocessing.

  16. Improved methods for reprocessing of GNSS data for climate monitoring over Poland

    NASA Astrophysics Data System (ADS)

    Stepniak, Katarzyna; Bock, Olivier; Wielgosz, Pawel

    2016-04-01

    The goal of this work is to determine the most accurate and homogeneous processing strategy to reprocess ground-based GNSS data for climate monitoring applications (analysis of trends and variability of Zenith Total Delay, ZTD, and Integrated Water Vapor, IWV). Namely, we investigate the impact of network design strategy and tropospheric modeling approach on the quality and homogeneity of both relative (double difference) and absolute (PPP) solutions. A network of 138 GNSS stations (including 33 stations from the EUREF Permanent Network, EPN, and 105 stations from ASG-EUPOS in Poland) is reprocessed for year 2014 using Bernese 5.2 GNSS software with the final IGS (International GNSS Service) orbits and clocks. First a standard (the shortest) "star" baseline design strategy is used in which the EPN stations are connected together defining a reference network and every ASG-EUPOS station is connected to the nearest EPN station. The initial network is modified automatically by the Bernese software every day depending on the availability of observations at the EPN stations. We show that in case of sub-daily gaps in the measurements of the reference stations, small clusters of stations can be disconnected from the main reference network. As a result, offsets of a few centimeters in ZTD estimates and spikes in formal errors can appear. These offsets and spikes cannot always be detected. This phenomenon is quite frequent in a large network such as considered in this study. It is also responsible for significant discontinuities in the estimated ZTD series which are detrimental to climate monitoring applications. We developed a new baseline design strategy algorithm to circumvent this event and assure that all the stations remain connected to the main reference network. It is shown that using this strategy, the reprocessed ZTD series are much more continuous and homogeneous in comparison to the standard strategy. The results are further validated against a Precise Point

  17. Process to separate transuranic elements from nuclear waste

    DOEpatents

    Johnson, Terry R.; Ackerman, John P.; Tomczuk, Zygmunt; Fischer, Donald F.

    1989-01-01

    A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR).

  18. High-level disinfection, sterilization, and antisepsis: current issues in reprocessing medical and surgical instruments.

    PubMed

    Seavey, Rose

    2013-05-01

    Technology is rapidly changing many aspects of health care. The intricate design of instruments, the configuration of instrument trays, and evidence-based practice have resulted in the need for complicated and specific reprocessing recommendations from instrument manufacturers. Patient safety depends on instruments that are appropriately cared for and adequately reprocessed. This article covers current issues that sterile processing and operating room professionals must deal with regarding reprocessing of medical and surgical instruments.

  19. Corrosion studies in fuel element reprocessing environments containing nitric acid

    SciTech Connect

    Beavers, J A; White, R R; Berry, W E; Griess, J C

    1982-04-01

    Nitric acid is universally used in aqueous fuel element reprocessing plants; however, in the processing scheme being developed by the Consolidated Fuel Reprocessing Program, some of the equipment will be exposed to nitric acid under conditions not previously encountered in fuel element reprocessing plants. A previous report presented corrosion data obtained in hyperazeotropic nitric acid and in concentrated magnesium nitrate solutions used in its preparation. The results presented in this report are concerned with the following: (1) corrosion of titanium in nitric acid; (2) corrosion of nickel-base alloys in a nitric acid-hydrofluoric acid solution; (3) the formation of Cr(VI), which enhances corrosion, in nitric acid solutions; and (4) corrosion of mechanical pipe connectors in nitric acid. The results show that the corrosion rate of titanium increased with the refreshment rate of boiling nitric acid, but the effect diminished rapidly as the temperature decreased. The addition of iodic acid inhibited attack. Also, up to 200 ppM of fluoride in 70% HNO/sub 3/ had no major effect on the corrosion of either titanium or tantalum. In boiling 8 M HNO/sub 3/-0.05 M HF, Inconel 671 was more resistant than Inconel 690, but both alloys experienced end-grain attack. In the case of Inconel 671, heat treatment was very important; annealed and quenched material was much more resistant than furnace-cooled material.The rate of oxidation of Cr(III) to Cr(VI) increased significantly as the nitric acid concentration increased, and certain forms of ruthenium in the solution seemed to accelerate the rate of formation. Mechanical connectors of T-304L stainless steel experienced end-grain attack on the exposed pipe ends, and seal rings of both stainless steel and a titanium alloy (6% Al-4% V) underwent heavy attack in boiling 8 M HNO/sub 3/.

  20. High-level disinfection of gastrointestinal endoscope reprocessing

    PubMed Central

    Chiu, King-Wah; Lu, Lung-Sheng; Chiou, Shue-Shian

    2015-01-01

    High level disinfection (HLD) of the gastrointestinal (GI) endoscope is not simply a slogan, but rather is a form of experimental monitoring-based medicine. By definition, GI endoscopy is a semicritical medical device. Hence, such medical devices require major quality assurance for disinfection. And because many of these items are temperature sensitive, low-temperature chemical methods, such as liquid chemical germicide, must be used rather than steam sterilization. In summarizing guidelines for infection prevention and control for GI endoscopy, there are three important steps that must be highlighted: manual washing, HLD with automated endoscope reprocessor, and drying. Strict adherence to current guidelines is required because compared to any other medical device, the GI endoscope is associated with more outbreaks linked to inadequate cleaning or disinfecting during HLD. Both experimental evaluation on the surveillance bacterial cultures and in-use clinical results have shown that, the monitoring of the stringent processes to prevent and control infection is an essential component of the broader strategy to ensure the delivery of safe endoscopy services, because endoscope reprocessing is a multistep procedure involving numerous factors that can interfere with its efficacy. Based on our years of experience in the surveillance of culture monitoring of endoscopic reprocessing, we aim in this study to carefully describe what details require attention in the GI endoscopy disinfection and to share our experience so that patients can be provided with high quality and safe medical practices. Quality management encompasses all aspects of pre- and post-procedural care including the efficiency of the endoscopy unit and reprocessing area, as well as the endoscopic procedure itself. PMID:25699232