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Sample records for radioactive salt waste

  1. Characterization of a ceramic waste form encapsulating radioactive electrorefiner salt

    SciTech Connect

    Moschetti, T. L.; Sinkler, W.; DiSanto, T.; Noy, M.; Warren, A. R.; Cummings, D. G.; Johnson, S. G.; Goff, K. M.; Bateman, K. J.; Frank, S. M.

    1999-11-11

    Argonne National Laboratory has developed a ceramic waste form to immobilize radioactive waste salt produced during the electrometallurgical treatment of spent fuel. This study presents the first results from electron microscopy and durability testing of a ceramic waste form produced from that radioactive electrorefiner salt. The waste form consists of two primary phases: sodalite and glass. The sodalite phase appears to incorporate most of the alkali and alkaline earth fission products. Other fission products (rare earths and yttrium) tend to form a separate phase and are frequently associated with the actinides, which form mixed oxides. Seven-day leach test results are also presented.

  2. Membrane Treatment of Liquid Salt Bearing Radioactive Wastes

    SciTech Connect

    Dmitriev, S. A.; Adamovich, D. V.; Demkin, V. I.; Timofeev, E. M.

    2003-02-25

    The main fields of introduction and application of membrane methods for preliminary treatment and processing salt liquid radioactive waste (SLRW) can be nuclear power stations (NPP) and enterprises on atomic submarines (AS) utilization. Unlike the earlier developed technology for the liquid salt bearing radioactive waste decontamination and concentrating this report presents the new enhanced membrane technology for the liquid salt bearing radioactive waste processing based on the state-of-the-art membrane unit design, namely, the filtering units equipped with the metal-ceramic membranes of ''TruMem'' brand, as well as the electrodialysis and electroosmosis concentrators. Application of the above mentioned units in conjunction with the pulse pole changer will allow the marked increase of the radioactive waste concentrating factor and the significant reduction of the waste volume intended for conversion into monolith and disposal. Besides, the application of the electrodialysis units loaded with an ion exchange material at the end polishing stage of the radioactive waste decontamination process will allow the reagent-free radioactive waste treatment that meets the standards set for the release of the decontaminated liquid radioactive waste effluents into the natural reservoirs of fish-farming value.

  3. BLENDING ANALYSIS FOR RADIOACTIVE SALT WASTE PROCESSING FACILITY

    SciTech Connect

    Lee, S.

    2012-05-10

    Savannah River National Laboratory (SRNL) evaluated methods to mix and blend the contents of the blend tanks to ensure the contents are properly blended before they are transferred from the blend tank such as Tank 21 and Tank 24 to the Salt Waste Processing Facility (SWPF) feed tank. The tank contents consist of three forms: dissolved salt solution, other waste salt solutions, and sludge containing settled solids. This paper focuses on developing the computational model and estimating the operation time of submersible slurry pump when the tank contents are adequately blended prior to their transfer to the SWPF facility. A three-dimensional computational fluid dynamics approach was taken by using the full scale configuration of SRS Type-IV tank, Tank 21H. Major solid obstructions such as the tank wall boundary, the transfer pump column, and three slurry pump housings including one active and two inactive pumps were included in the mixing performance model. Basic flow pattern results predicted by the computational model were benchmarked against the SRNL test results and literature data. Tank 21 is a waste tank that is used to prepare batches of salt feed for SWPF. The salt feed must be a homogeneous solution satisfying the acceptance criterion of the solids entrainment during transfer operation. The work scope described here consists of two modeling areas. They are the steady state flow pattern calculations before the addition of acid solution for tank blending operation and the transient mixing analysis during miscible liquid blending operation. The transient blending calculations were performed by using the 95% homogeneity criterion for the entire liquid domain of the tank. The initial conditions for the entire modeling domain were based on the steady-state flow pattern results with zero second phase concentration. The performance model was also benchmarked against the SRNL test results and literature data.

  4. Container materials for isolation of radioactive waste in salt

    SciTech Connect

    Streicher, M.A.; Andrews, A.

    1987-10-01

    The workshop reviewed the extensive data on the corrosion resistance of low-carbon steel in simulated salt repository environments, determined whether these data were sufficient to recommend low-carbon steel for fabrication of the container, and assessed the suitability of other materials under consideration in the SRP. The panelists determined the need for testing and research programs, recommended experimental approaches, and recommended materials based on existing technology. On the first day of the workshop, presentations were made on waste package requirements; the expected corrosion environment; degradation processes, including a review of data from corrosion tests on carbon steel; and rationales for container design and materials, modeling studies, and planned future work. The second day was devoted to a panel caucus, presentation of workshop findings, and open discussion. 76 refs., 2 figs., 3 tabs.

  5. Cement-based waste forms for disposal of Savannah River Plant low-level radioactive salt waste

    SciTech Connect

    Langton, C A; Dukes, M D; Simmons, R V

    1983-01-01

    Defense waste processing at the Savannah River Plant will include decontamination and disposal of approximately 100 million liters of soluble salts containing primarily NaNO/sub 3/, NaOH, NaNO/sub 2/, NaAl(OH)/sub 4/, and Na/sub 2/SO/sub 4/. A cement-based waste form, saltstone, has been designed for disposal of Savannah River Plant low-level radioactive salt waste. Bulk properties of this material have been tailored with respect to salt leach rate, permeability, and compressive strength. Microstructure and mineralogy of leached and unleached specimens were characterized by SEM and x-ray diffraction analyses, respectively. It has been concluded that the salt leach rate can be limited so that amounts of salt and radionuclides in the groundwater at the perimeter of the 100-acre disposal site will not exceed EPA drinking water standards. 7 references, 4 figures, 6 tables.

  6. Stabilization/Solidification of radioactive molten salt waste via gel-route pretreatment.

    PubMed

    Park, Hwan-Seo; Kim, In-Tae; Kim, Hwan-Young; Ryu, Seung-Kon; Kim, Joon-Hyung

    2007-02-15

    The volatilization of radionuclides during the stabilization/solidification of radioactive wastes at high temperatures is one of the major problems to be considered in choosing suitable wasteforms, process, material systems, etc. This paper reports a novel method to convert volatile wastes into nonvolatile compounds via a sol-gel process, which is different from the conventional method using metal-alkoxides and organic solvents. The material system was designed with sodium silicate (Si) as a gelling agent, phosphoric acid (P) as a catalyst/stabilizer, aluminum nitrate (Al) as a property promoter, and H20 as a solvent. A novel structural model for the chemical conversion of molten salt waste, named RPRM (Reaction Product in Reaction Module), was established, and the waste could be solidified with glass matrix via a simple procedure. The leached fraction of Cs and Sr by a PCT leaching method was 0.72% and 0.014%, respectively. In conclusion, the RPRM model isto converttargetwastes into stable and manageable products, not to obtain a specific crystalline product for each radionuclide. This paper suggested a new stabilization/solidification method for salt wastes by establishing the gel-forming material system and showing a practical example, not a new synthesis method of stable crystalline phase. This process, named "gel-route stabilization/solidification (GRSS)", will be a prospective alternative with stable chemical process on the immobilization of salt wastes and various mixed radioactive waste for final disposal. PMID:17593740

  7. Treatment of Liquid Radioactive Waste with High Salt Content by Colloidal Adsorbents - 13274

    SciTech Connect

    Lee, Keun-Young; Chung, Dong-Yong; Kim, Kwang-Wook; Lee, Eil-Hee; Moon, Jei-Kwon

    2013-07-01

    Treatment processes have been fully developed for most of the liquid radioactive wastes generated during the operation of nuclear power plants. However, a process for radioactive liquid waste with high salt content, such as waste seawater generated from the unexpected accident at nuclear power station, has not been studied extensively. In this study, the adsorption efficiencies of cesium (Cs) and strontium (Sr) in radioactive liquid waste with high salt content were investigated using several types of zeolite with different particle sizes. Synthesized and commercial zeolites were used for the treatment of simulated seawater containing Cs and Sr, and the reaction kinetics and adsorption capacities of colloidal zeolites were compared with those of bulk zeolites. The experimental results demonstrated that the colloidal adsorbents showed fast adsorption kinetic and high binding capacity for Cs and Sr. Also, the colloidal zeolites could be successfully applied to the static adsorption condition, therefore, an economical benefit might be expected in an actual processes where stirring is not achievable. (authors)

  8. Saltstone: cement-based waste form for disposal of Savannah River Plant low-level radioactive salt waste

    SciTech Connect

    Langton, C.A.

    1984-01-01

    Defense waste processing at the Savannah River Plant will include decontamination and disposal of approximately 400 million liters of waste containing NaNO/sub 3/, NaOH, Na/sub 2/SO/sub 4/, and NaNO/sub 2/. After decontamination, the salt solution is classified as low-level waste. A cement-based waste form, saltstone, has been designed for disposal of Savannah River Plant low-level radioactive salt waste. Bulk properties of this material have been tailored with respect to salt leach rate, permeability, and compressive strength. Microstructure and mineralogy of leached and unleached specimens were characterized by SEM and x-ray diffraction analyses. The disposal system for the DWPF salt waste includes reconstitution of the crystallized salt as a solution containing 32 wt % solids. This solution will be decontaminated to remove /sup 137/Cs and /sup 90/Sr and then stabilized in a cement-based waste form. Laboratory and field tests indicate that this stabilization process greatly reduces the mobility of all of the waste constitutents in the surface and near-surface environment. Engineered trenches for subsurface burial of the saltstone have been designed to ensure compatibility between the waste form and the environment. The total disposal sytem, saltstone-trench-surrounding soil, has been designed to contain radionuclides, Cr, and Hg by both physical encapsulation and chemical fixation mechanisms. Physical encapsulation of the salts is the mechanism employed for controlling N and OH releases. In this way, final disposal of the SRP low-level waste can be achieved and the quality of the groundwater at the perimeter of the disposal site meets EPA drinking water standards.

  9. Groundwater occurrence and the dissolution of salt at the WIPP Radioactive Waste Repository Site

    NASA Astrophysics Data System (ADS)

    Chaturvedi, Lokesh; Rehfeldt, Kenneth

    1984-04-01

    The Waste Isolation Pilot Plant (WIPP), located about 25 miles east of Carlsbad, in southeastern New Mexico, is slated to be the first deep geologic repository for permanent disposal of radioactive wastes in the United States. The repository will be located in bedded salt of Permian (225 m.y. B.P.) age, at a depth of 655 m below the ground surface. The present mission of WIPP calls for a permanent disposal of approximately 170,000 cubic meters of defense transuranic (TRU) wastes and for temporary, retrievable emplacement of 4.25 cubic meters of experimental high level wastes. The site will not be licensed by the Nuclear Regulatory Commission (NRC) but will comply with the Environmental Protection Agency (EPA) standards and all other federal and state regulations.

  10. Aspects of the thermal and transport properties of crystalline salt in designing radioactive waste storages in halogen formations

    SciTech Connect

    Nikitin, A. N. Pocheptsova, O. A.; Matthies, S.

    2010-05-15

    Some of the properties of natural rock salt are described. This rock is of great practical interest, because, along with its conventional applications in the chemical and food industries, it is promising for use in engineering underground radioactive waste storages and natural gas reservoirs. The results of structural and texture studies of rock salt by neutron diffraction are discussed. The nature of the salt permeability under temperature and stress gradients is theoretically estimated.

  11. 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)

  12. Radioactive Wastes.

    PubMed

    Choudri, B S; Baawain, Mahad

    2016-10-01

    Papers reviewed herein present a general overview of radioactive waste activities around the world in 2015. These include safety assessments, decommission and decontamination of nuclear facilities, fusion facilities, transportation and management solutions for the final disposal of low and high level radioactive wastes (LLW and HLW), interim storage and final disposal options for spent fuel (SF), and tritiated wastes, with a focus on environmental impacts due to the mobility of radionuclides in water, soil and ecosystem alongwith other progress made in the management of radioactive wastes. PMID:27620100

  13. Radioactive Wastes.

    PubMed

    Choudri, B S; Baawain, Mahad

    2015-10-01

    Papers reviewed herein present a general overview of radioactive waste activities around the world in 2014. These include safety assessments, decommission and decontamination of nuclear facilities, fusion facilities, transportation and management solutions for the final disposal of low and high level radioactive wastes (LLW and HLW), interim storage and final disposal options for spent fuel (SF), and tritiated wastes, with a focus on environmental impacts due to the mobility of radionuclides in water, soil and ecosystem alongwith other progress made in the management of radioactive wastes. PMID:26420096

  14. Radioactive waste isolation in salt: geochemistry of brine in rock salt in temperature gradients and gamma-radiation fields - a selective annotated bibliography

    SciTech Connect

    Hull, A.B.; Williams, L.B.

    1985-07-01

    Evaluation of the extensive research concerning brine geochemistry and transport is critically important to successful exploitation of a salt formation for isolating high-level radioactive waste. This annotated bibliography has been compiled from documents considered to provide classic background material on the interactions between brine and rock salt, as well as the most important results from more recent research. Each summary elucidates the information or data most pertinent to situations encountered in siting, constructing, and operating a mined repository in salt for high-level radioactive waste. The research topics covered include the basic geology, depositional environment, mineralogy, and structure of evaporite and domal salts, as well as fluid inclusions, brine chemistry, thermal and gamma-radiation effects, radionuclide migration, and thermodynamic properties of salts and brines. 4 figs., 6 tabs.

  15. Radioactive wastes

    SciTech Connect

    Devarakonda, M.S.; Hickox, J.A.

    1996-11-01

    This paper provides a review of literature published in 1995 on the subject of radioactive wastes. Topics covered include: national programs; waste repositories; mixed wastes; decontamination and decommissioning; remedial actions and treatment; and environmental occurrence and transport of radionuclides. 155 refs.

  16. Radioactive Waste.

    ERIC Educational Resources Information Center

    Blaylock, B. G.

    1978-01-01

    Presents a literature review of radioactive waste disposal, covering publications of 1976-77. Some of the studies included are: (1) high-level and long-lived wastes, and (2) release and burial of low-level wastes. A list of 42 references is also presented. (HM)

  17. Method to synthesize dense crystallized sodalite pellet for immobilizing halide salt radioactive waste

    DOEpatents

    Koyama, T.

    1992-01-01

    This report describes a method for immobilizing waste chloride salts containing radionuclides such as cesium and strontium and hazardous materials such as barium. A sodalite intermediate is prepared by mixing appropriate amounts of silica, alumina and sodium hydroxide with respect to sodalite and heating the mixture to form the sodalite intermediate and water. Heating is continued to drive off the water to form a water-free intermediate. The water-free intermediate is mixed with either waste salt or waste salt which has been contacted with zeolite to concentrate the radionuclides and hazardous material. The waste salt-intermediate mixture is then compacted and heated under conditions of heat and pressure to form sodalite with the waste salt, radionuclides and hazardous material trapped within the sodalite cage structure. This provides a final product having excellent leach resistant capabilities.

  18. Method to synthesize dense crystallized sodalite pellet for immobilizing halide salt radioactive waste

    DOEpatents

    Koyama, Tadafumi

    1994-01-01

    A method for immobilizing waste chloride salts containing radionuclides such as cesium and strontium and hazardous materials such as barium. A sodalite intermediate is prepared by mixing appropriate amounts of silica, alumina and sodium hydroxide with respect to sodalite and heating the mixture to form the sodalite intermediate and water. Heating is continued to drive off the water to form a water-free intermediate. The water-free intermediate is mixed with either waste salt or waste salt which has been contacted with zeolite to concentrate the radionuclides and hazardous material. The waste salt-intermediate mixture is then compacted and heated under conditions of heat and pressure to form sodalite with the waste salt, radionuclides and hazardous material trapped within the sodalite cage structure. This provides a final product having excellent leach resistant capabilities.

  19. Method to synthesize dense crystallized sodalite pellet for immobilizing halide salt radioactive waste

    DOEpatents

    Koyama, Tadafumi.

    1994-08-23

    A method is described for immobilizing waste chloride salts containing radionuclides such as cesium and strontium and hazardous materials such as barium. A sodalite intermediate is prepared by mixing appropriate amounts of silica, alumina and sodium hydroxide with respect to sodalite and heating the mixture to form the sodalite intermediate and water. Heating is continued to drive off the water to form a water-free intermediate. The water-free intermediate is mixed with either waste salt or waste salt which has been contacted with zeolite to concentrate the radionuclides and hazardous material. The waste salt-intermediate mixture is then compacted and heated under conditions of heat and pressure to form sodalite with the waste salt, radionuclides and hazardous material trapped within the sodalite cage structure. This provides a final product having excellent leach resistant capabilities.

  20. Technetium removal column flow testing with alkaline, high salt, radioactive tank waste

    SciTech Connect

    Blanchard, D.L. Jr.; Kurath, D.E.; Golcar, G.R.; Conradson, S.D.

    1996-09-30

    This report describes two bench-scale column tests conducted to demonstrate the removal of Tc-99 from actual alkaline high salt radioactive waste. The waste used as feed for these tests was obtained from the Hanford double shell tank AW-101, which contains double shell slurry feed (DSSF). The tank sample was diluted to approximately 5 M Na with water, and most of the Cs-137 was removed using crystalline silicotitanates. The tests were conducted with two small columns connected in series, containing, 10 mL of either a sorbent, ABEC 5000 (Eichrom Industries, Inc.), or an anion exchanger Reillex{trademark}-HPQ (Reilly Industries, Inc.). Both materials are selective for pertechnetate anion (TcO{sub 4}{sup -}). The process steps generally followed those expected in a full-scale process and included (1) resin conditioning, (2) loading, (3) caustic wash to remove residual feed and prevent the precipitation of Al(OH){sub 3}, and (4) elution. A small amount of Tc-99m tracer was added as ammonium pertechnetate to the feed and a portable GEA counter was used to closely monitor the process. Analyses of the Tc-99 in the waste was performed using ICP-MS with spot checks using radiochemical analysis. Technetium x-ray absorption spectroscopy (XAS) spectra of 6 samples were also collected to determine the prevalence of non-pertechnetate species [e.g. Tc(IV)].

  1. Radioactive waste isolation in salt: peer review of Office of Nuclear Waste Isolation's Socioeconomic Program Plan

    SciTech Connect

    Winter, R.; Fenster, D.; O'Hare, M.; Zillman, D.; Harrison, W.; Tisue, M.

    1984-02-01

    The ONWI Socioeconomic Program Plan spells out DOE's approach to analyzing the socioeconomic impacts from siting, constructing, and operating radioactive waste repositories and discusses mitigation strategies. The peer review indicated the following modifications should be made to the Plan: encourage active public participation in the decision-making processes leading to repository site selection; clearly define mechanisms for incorporating the concerns of local residents, state and local governments, and other potentially interested parties into the early stages of the site selection process; place significantly greater emphasis on using primary socioeconomic data during the site selection process, reversing the current overemphasis on secondary data collection, description of socioeconomic conditions at potential locations, and development of analytical methodologies; recognize that mitigation mechanisms other than compensation and incentives may be effective; as soon as potential sites are identified, the US Department of Energy (DOE) should begin discussing impact mitigation agreements with local officials and other interested parties; and comply fully with the pertinent provisions of NWPA.

  2. Radioactive Waste Isolation in Salt: Peer review of documents dealing with geophysical investigations

    SciTech Connect

    McGinnis, L.D.; Bowen, R.H.

    1987-03-01

    The Salt Repository Project, a US Department of Energy program to develop a mined repository in salt for high-level radioactive waste, is governed by a complex and sometimes inconsistent array of laws, administrative regulations, guidelines, and position papers. In conducting multidisciplinary peer reviews of contractor documents in support of this project, Argonne National Laboratory has needed to inform its expert reviewers of these governmental mandates, with particular emphasis on the relationship between issues and the technical work undertaken. This report acquaints peer review panelists with the regulatory framework as it affects their reviews of site characterization plans and related documents, including surface-based and underground test plans. Panelists will be asked to consider repository performance objectives and issues as they judge the adequacy of proposed geophysical testing. All site-specific discussions relate to the Deaf Smith County site in Texas, which was approved for site characterization by the President in May 1986. Natural processes active at the Deaf Smith County site and the status of geophysical testing near the site are reviewed briefly. 25 refs., 4 figs., 5 tabs.

  3. SAVANNAH RIVER SITE INCIPIENT SLUDGE MIXING IN RADIOACTIVE LIQUID WASTE STORAGE TANKS DURING SALT SOLUTION BLENDING

    SciTech Connect

    Leishear, R.; Poirier, M.; Lee, S.; Steeper, T.; Fowley, M.; Parkinson, K.

    2011-01-12

    This paper is the second in a series of four publications to document ongoing pilot scale testing and computational fluid dynamics (CFD) modeling of mixing processes in 85 foot diameter, 1.3 million gallon, radioactive liquid waste, storage tanks at Savannah River Site (SRS). Homogeneous blending of salt solutions is required in waste tanks. Settled solids (i.e., sludge) are required to remain undisturbed on the bottom of waste tanks during blending. Suspension of sludge during blending may potentially release radiolytically generated hydrogen trapped in the sludge, which is a safety concern. The first paper (Leishear, et. al. [1]) presented pilot scale blending experiments of miscible fluids to provide initial design requirements for a full scale blending pump. Scaling techniques for an 8 foot diameter pilot scale tank were also justified in that work. This second paper describes the overall reasons to perform tests, and documents pilot scale experiments performed to investigate disturbance of sludge, using non-radioactive sludge simulants. A third paper will document pilot scale CFD modeling for comparison to experimental pilot scale test results for both blending tests and sludge disturbance tests. That paper will also describe full scale CFD results. The final paper will document additional blending test results for stratified layers in salt solutions, scale up techniques, final full scale pump design recommendations, and operational recommendations. Specifically, this paper documents a series of pilot scale tests, where sludge simulant disturbance due to a blending pump or transfer pump are investigated. A principle design requirement for a blending pump is UoD, where Uo is the pump discharge nozzle velocity, and D is the nozzle diameter. Pilot scale test results showed that sludge was undisturbed below UoD = 0.47 ft{sup 2}/s, and that below UoD = 0.58 ft{sup 2}/s minimal sludge disturbance was observed. If sludge is minimally disturbed, hydrogen will not be

  4. Molten salt oxidation of mixed waste: Preliminary bench-scale experiments without radioactivity

    SciTech Connect

    Haas, P.A.; Rudolph, J.C.; Bell, J.T.

    1994-06-01

    Molten salt oxidation (MSO) is a process in which organic wastes are oxidized by sparging them with air through a bed of molten sodium carbonate (bp 851 {degrees}C) at {ge} 900{degrees}C. This process is readily applicable to the mixed waste because acidic products from Cl, S, P, etc., in the waste, along with most metals and most radionuclides, are retained within the melt as oxides or salts. Rockwell International has studied the application of MSO to various wastes, including some mixed waste. A unit used by Rockwell to study the mixed waste treatment is presently in use at Oak Ridge National Laboratory (ORNL). ORNL`s studies to date have concentrated on chemical flowsheet questions. Concerns that were studied included carbon monoxide (CO) emissions, NO{sub x}, emissions, and metal retention under a variety of conditions. Initial experiments show that CO emissions increase with increasing NaCl content in the melt, increasing temperature, and increasing airflow. Carbon monoxide content is especially high (> 2000 ppm) with high chlorine content (> 10%). Thermal NO{sub x}, emissions are relatively low ( < 5 ppm) at temperatures < 1000{degrees}C. However, most (85--100%) of the nitrogen in the feed as organic nitrate or amine was released as NO{sub x}, The metal contents of the melt and of knockout pot samples of condensed salt show high volatilities of Cs as CsCl. Average condensed salt concentrations were 60% for barium and 100% for strontium and cobalt. The cerium disappeared -- perhaps from deposition on the alumina reactor walls.

  5. Radioactive waste isolation in salt: peer review of Office of Nuclear Waste Isolation's Socioeconomic Program Plan

    SciTech Connect

    Winter, R.; Fenster, D.; O'Hare, M.; Zillman, D.; Harrison, W.; Tisue, M.

    1984-07-01

    The following recommendations have been abstracted from the body of this report. The Office of Nuclear Waste Isolation's Socioeconomic Program Plan for the Establishment of Mined Geologic Repositories to Isolate Nuclear Waste should be modified to: (1) encourage active public participation in the decision-making processes leading to repository site selection; (2) clearly define mechanisms for incorporating the concerns of local residents, state and local governments, and other potentially interested parties into the early stages of the site selection process. In addition, the Office of Nuclear Waste Isolation should carefully review the overall role that these persons and groups, including local pressure groups organized in the face of potential repository development, will play in the siting process; (3) place significantly greater emphasis on using primary socioeconomic data during the site selection process, reversing the current overemphasis on secondary data collection, description of socioeconomic conditions at potential locations, and development of analytical methodologies; (4) include additional approaches to solving socioeconomic problems. For example, a reluctance to acknowledge that solutions to socioeconomic problems need to be found jointly with interested parties is evident in the plan; (5) recognize that mitigation mechanisms other than compensation and incentives may be effective; (6) as soon as potential sites are identified, the US Department of Energy (DOE) should begin discussing impact mitigation agreements with local officials and other interested parties; and (7) comply fully with the pertinent provisions of NWPA.

  6. Dechlorination and stabilization of radioactive chloride salt waste in a molten state

    SciTech Connect

    In-Tae Kim; Hwan-Seo Park; Yong-Jun Cho; Hwan-Young Kim; Seong-Won Park; Eung-Ho Kim

    2007-07-01

    This study suggests a new method to stabilize the molten salt wastes generated from he pyro-processing of a LWR spent fuel. Using a conventional sol-gel process, an inorganic material (SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}, SAP) reactive to metal chlorides was prepared. In this paper, the reactivity of the SAP on the metal chlorides at 650-850 deg. C, the thermal stability of the reaction products and their leach-resistance under the PCT-A leach test were investigated. In the SAP, three different kinds of chains are available; Si-O-Si (main chain), Si-O-Al (side chain) and Al-O-P/P-O-P (reactive chain). Alkali metal chlorides were converted into metal aluminosilicate (Li{sub x}Al{sub x}Si{sub 1-x}O{sub 2-x}) and metal phosphate(Li{sub 3}PO{sub 4} and Cs{sub 2}AlP{sub 3}O{sub 10}) while the alkaline earth and rare earth chlorides were changed into only metal phosphates (Sr{sub 5}(PO{sub 4}){sub 3}Cl and CePO{sub 4}). The conversion rate was about 96% at a salt waste/SAP weight ratio of 0.5 and a weight loss up to 1100 deg. C measured by the thermo-gravimetric analysis was below 1 Wt%. The leach rates of Cs and Sr under the PCT-A leaching condition were about 10{sup -2} and 10{sup -4} g/m{sup 3}.day, respectively. From these results, it could be concluded that the SAP developed in this study can be considered as an effective stabilizer for metal chlorides and the method of using the SAP could provide a chance to minimize the final waste volume to be disposed off. (authors)

  7. Radioactive waste isolation in salt: special advisory report on the status of the Office of Nuclear Waste Isolation's plans for repository performance assessment

    SciTech Connect

    Ditmars, J.D.; Walbridge, E.W.; Rote, D.M.; Harrison, W.; Herzenberg, C.L.

    1983-10-01

    Repository performance assessment is analysis that identifies events and processes that might affect a repository system for isolation of radioactive waste, examines their effects on barriers to waste migration, and estimates the probabilities of their occurrence and their consequences. In 1983 Battelle Memorial Institute's Office of Nuclear Waste Isolation (ONWI) prepared two plans - one for performance assessment for a waste repository in salt and one for verification and validation of performance assessment technology. At the request of the US Department of Energy's Salt Repository Project Office (SRPO), Argonne National Laboratory reviewed those plans and prepared this report to advise SRPO of specific areas where ONWI's plans for performance assessment might be improved. This report presents a framework for repository performance assessment that clearly identifies the relationships among the disposal problems, the processes underlying the problems, the tools for assessment (computer codes), and the data. In particular, the relationships among important processes and 26 model codes available to ONWI are indicated. A common suggestion for computer code verification and validation is the need for specific and unambiguous documentation of the results of performance assessment activities. A major portion of this report consists of status summaries of 27 model codes indicated as potentially useful by ONWI. The code summaries focus on three main areas: (1) the code's purpose, capabilities, and limitations; (2) status of the elements of documentation and review essential for code verification and validation; and (3) proposed application of the code for performance assessment of salt repository systems. 15 references, 6 figures, 4 tables.

  8. Efficacy of backfilling and other engineered barriers in a radioactive waste repository in salt

    SciTech Connect

    Claiborne, H.C.

    1982-09-01

    In the United States, investigation of potential host geologic formations was expanded in 1975 to include hard rocks. Potential groundwater intrusion is leading to very conservative and expensive waste package designs. Recent studies have concluded that incentives for engineered barriers and 1000-year canisters probably do not exist for reasonable breach scenarios. The assumption that multibarriers will significantly increase the safety margin is also questioned. Use of a bentonite backfill for surrounding a canister of exotic materials was developed in Sweden and is being considered in the US. The expectation that bentonite will remain essentially unchanged for hundreds of years for US repository designs may be unrealistic. In addition, thick bentonite backfills will increase the canister surface temperature and add much more water around the canister. The use of desiccant materials, such as CaO or MgO, for backfilling seems to be a better method of protecting the canister. An argument can also be made for not using backfill material in salt repositories since the 30-cm-thick space will provide for hole closure for many years and will promote heat transfer via natural convection. It is concluded that expensive safety systems are being considered for repository designs that do not necessarily increase the safety margin. It is recommended that the safety systems for waste repositories in different geologic media be addressed individually and that cost-benefit analyses be performed.

  9. Simulated waste package test in salt

    SciTech Connect

    Kalia, H.N.

    1994-03-01

    The Salt Repository Site Characterization Project Office (SRPO), of the US Department of Energy (DOE) Office of the Civilian Radioactive Waste Management (OCRWM), in cooperation with Federal Republic of Germany (FRG), simulated a waste package test at Asse Salt Mine (Asse). The purpose of this test was to determine the effect of heat produced by the decay of High-Level Radioactive Waste (HLW) on: Migration of brine moisture; Thermomechanical response of the salt; Geomechanical response of the room mined in salt; Corrosion on potential HLW waste package container materials; and Generation of gases. This paper describes the these performed, results obtained, and the performance of instruments and data acquisition system deployed.

  10. Ceramicrete stabilization of radioactive-salt-containing liquid waste and sludge water. Final CRADA report.

    SciTech Connect

    Ehst, D.; Nuclear Engineering Division

    2010-08-04

    It was found that the Ceramicrete Specimens incorporated the Streams 1 and 2 sludges with the adjusted loading about 41.6 and 31.6%, respectively, have a high solidity. The visible cracks in the matrix materials and around the anionite AV-17 granules included could not obtain. The granules mentioned above fixed by Ceramicrete matrix very strongly. Consequently, we can conclude that irradiation of Ceramecrete matrix, goes from the high radioactive elements, not result the structural degradation. Based on the chemical analysis of specimens No.462 and No.461 used it was shown that these matrix included the formation elements (P, K, Mg, O), but in the different samples their correlations are different. These ratios of the content of elements included are about {+-} 10%. This information shows a great homogeneity of matrix prepared. In the list of the elements founded, expect the matrix formation elements, we detected also Ca and Si (from the wollastonite - the necessary for Ceramicrete compound); Na, Al, S, O, Cl, Fe, Ni also have been detected in the Specimen No.642 from the waste forms: NaCl, Al(OH){sub 3}, Na{sub 2}SO{sub 4}. Fe(OH){sub 3}, nickel ferrocyanide and Ni(NO{sub 3})2. The unintelligible results also were found from analysis of an AV-17 granules, in which we obtain the great amount of K. The X-ray radiographs of the Ceramicrete specimens with loading 41.4 % of Stream 1 and 31.6% of Stream 2, respectively showed that the realization of the advance technology, created at GEOHKI, leads to formation of excellent ceramic matrix with high amount of radioactive streams up to 40% and more. Really, during the interaction with start compounds MgO and KH{sub 2}PO{sub 4} with the present of H{sub 3}BO{sub 3} and Wollastonite this process run with high speed under the controlled regimes. That fact that the Ceramicrete matrix with 30-40% of Streams 1 and 2 have a crystalline form, not amorphous matter, allows to permit that these matrix should be very stable, reliable

  11. Radioactive waste isolation in salt: rationale and methodology for Argonne-conducted reviews of site characterization programs

    SciTech Connect

    Harrison, W.; Ditmars, J.D.; Tisue, M.W.; Hambley, D.F.; Fenster, D.F.; Rote, D.M.

    1985-07-01

    Both regulatory and technical concerns must be addressed in Argonne-conducted peer reviews of site characterization programs for individual sites for a high-level radioactive waste repository in salt. This report describes the regulatory framework within which reviews must be conducted and presents background information on the structure and purpose of site characterization programs as found in US Nuclear Regulatory Commission (NRC) Regulatory Guide 4.17 and Title 10, Part 60, of the Code of Federal Regulations. It also presents a methodology to assist reviewers in addressing technical concerns relating to their respective areas of expertise. The methodology concentrates on elements of prime importance to the US Department of Energy's advocacy of a given salt repository system during the NRC licensing process. Instructions are given for reviewing 12 site characterization program elements, starting with performance objectives, performance issues, and levels of performance of repository subsystem components; progressing through performance assessment; and ending with plans for data acquisition and evaluation. The success of a site characterization program in resolving repository performance issues will be determined by judging the likelihood that the proposed data acquisition activities will reduce uncertainties in the performance predictions. 8 refs., 3 figs., 5 tabs.

  12. 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.

  13. Radioactive Wastes. Revised.

    ERIC Educational Resources Information Center

    Fox, Charles H.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. This booklet deals with the handling, processing and disposal of radioactive wastes. Among the topics discussed are: The Nature of Radioactive Wastes; Waste Management; and Research and Development. There are…

  14. Technetium in alkaline, high-salt, radioactive tank waste supernate: Preliminary characterization and removal

    SciTech Connect

    Blanchard, D.L. Jr.; Brown, G.N.; Conradson, S.D.

    1997-01-01

    This report describes the initial work conducted at Pacific Northwest National Laboratory to study technetium (Tc) removal from Hanford tank waste supernates and Tc oxidation state in the supernates. Filtered supernate samples from four tanks were studied: a composite double shell slurry feed (DSSF) consisting of 70% from Tank AW-101, 20% from AP-106, and 10% from AP-102; and three complexant concentrate (CC) wastes (Tanks AN-107, SY-101, ANS SY-103) that are distinguished by having a high concentration of organic complexants. The work included batch contacts of these waste samples with Reillex{trademark}-HPQ (anion exchanger from Reilly Industries) and ABEC 5000 (a sorbent from Eichrom Industries), materials designed to effectively remove Tc as pertechnetate from tank wastes. A short study of Tc analysis methods was completed. A preliminary identification of the oxidation state of non-pertechnetate species in the supernates was made by analyzing the technetium x-ray absorption spectra of four CC waste samples. Molybdenum (Mo) and rhenium (Re) spiked test solutions and simulants were tested with electrospray ionization-mass spectrometry to evaluate the feasibility of the technique for identifying Tc species in waste samples.

  15. Radioactive mixed waste disposal

    SciTech Connect

    Jasen, W.G.; Erpenbeck, E.G.

    1993-02-01

    Various types of waste have been generated during the 50-year history of the Hanford Site. Regulatory changes in the last 20 years have provided the emphasis for better management of these wastes. Interpretations of the Atomic Energy Act of 1954 (AEA), the Resource Conservation and Recovery Act of 1976 (RCRA), and the Hazardous and Solid Waste Amendments (HSWA) have led to the definition of radioactive mixed wastes (RMW). The radioactive and hazardous properties of these wastes have resulted in the initiation of special projects for the management of these wastes. Other solid wastes at the Hanford Site include low-level wastes, transuranic (TRU), and nonradioactive hazardous wastes. This paper describes a system for the treatment, storage, and disposal (TSD) of solid radioactive waste.

  16. Radioactive waste disposal package

    DOEpatents

    Lampe, Robert F.

    1986-11-04

    A radioactive waste disposal package comprising a canister for containing vitrified radioactive waste material and a sealed outer shell encapsulating the canister. A solid block of filler material is supported in said shell and convertible into a liquid state for flow into the space between the canister and outer shell and subsequently hardened to form a solid, impervious layer occupying such space.

  17. Radioactive waste disposal package

    DOEpatents

    Lampe, Robert F.

    1986-01-01

    A radioactive waste disposal package comprising a canister for containing vitrified radioactive waste material and a sealed outer shell encapsulating the canister. A solid block of filler material is supported in said shell and convertible into a liquid state for flow into the space between the canister and outer shell and subsequently hardened to form a solid, impervious layer occupying such space.

  18. Radioactive waste isolation in salt: Peer review of the Golder Associates draft test plan for in situ testing in an exploratory shaft in salt

    SciTech Connect

    Hambley, D.F.; Mraz, D.Z.; Unterberter, R.R.; Stormont, J.C.; Neuman, S.P.; Russell, J.E.; Jacoby, C.H.; Hull, A.B.; Brady, B.H.G.; Ditmars, J.D.

    1987-01-01

    This report documents the peer review conducted by Argonne National Laboratory of a document entitled ''Draft Test Plan for In Situ Testing in an Exploratory Shaft in Salt,'' prepared for Battelle Memorial Institute's Office of Nuclear Waste Isolation by Golder Associates, Inc. In general, the peer review panelists found the test plan to be technically sound, although some deficiencies were identified. Recommendations for improving the test plan are presented in this review report. A microfiche copy of the following unpublished report is attached to the inside back cover of this report: ''Draft Test Plan for In Situ Testing in an Exploratory Shaft in Salt,'' prepared by Golder Associates, Inc., for Office of Nuclear Waste Isolation, Battelle Memorial Institute, Columbus, Ohio (March 1985).

  19. Organic waste processing using molten salt oxidation

    SciTech Connect

    Adamson, M. G., LLNL

    1998-03-01

    Molten Salt Oxidation (MSO) is a thermal means of oxidizing (destroying) the organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. The U. S. Department of Energy`s Office of Environmental Management (DOE/EM) is currently funding research that will identify alternatives to incineration for the treatment of organic-based mixed wastes. (Mixed wastes are defined as waste streams which have both hazardous and radioactive properties.) One such project is Lawrence Livermore National Laboratory`s Expedited Technology Demonstration of Molten Salt Oxidation (MSO). The goal of this project is to conduct an integrated demonstration of MSO, including off-gas and spent salt treatment, and the preparation of robust solid final forms. Livermore National Laboratory (LLNL) has constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are presently being performed under carefully controlled (experimental) conditions. The system consists of a MSO process vessel with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. In this paper we describe the integrated system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is to identify the most suitable waste streams and waste types for MSO treatment.

  20. Radioactive waste isolation in salt: peer review of the D'Appolonia report on Schematic Designs for Penetration Seals for a Repository in the Permian Basin, Texas

    SciTech Connect

    Hambley, D.F.; Stormont, J.C.; Russell, J.E.; Edgar, D.E.; Fenster, D.F.; Harrison, W.; Tisue, M.W.

    1984-09-01

    Argonne made the following recommedations for improving the reviewed reports. The authors of the report should: state the major assumptions of the study in Sec. 1.1 rather than later in the report; consider using salt for the shaft seals in salt horizons; reconsider whether keys are needed for the bulkheads; provide for interface grouting because use of expansive cement will not guarantee that interfaces will be impermeable; discuss the sealing schedule and, where appropriate, consider what needs to be done to ensure that emplaced radioactive waste could be retrieved if necessary; describe in more detail the sealing of the Dockum and Ogallala aquifers; consider an as low as reasonably achievable approach to performance requirements for the initial design phase; address the concerns in the 1983 US Nuclear Regulatory Commission document entitled Draft Technical Position: Borehole and Shaft Sealing of High-Level Nuclear Waste Repositories; cite the requirements for release of radioactivity by referring to specific clauses in the regulations of the US Environmental Protection Agency; and provide further explanation in the outline of future activities about materials development and verification testing. More emphasis on development of accelerated testing programs is also required.

  1. Geohydrology of the northern Louisiana salt-dome basin pertinent to the storage of radioactive wastes; a progress report

    USGS Publications Warehouse

    Hosman, R.L.

    1978-01-01

    Salt domes in northern Louisiana are being considered as possible storage sites for nuclear wastes. The domes are in an area that received regional sedimentation through early Tertiary (Eocene) time with lesser amounts of Quaternary deposits. The Cretaceous-Tertiary accumulation is a few thousand feet thick; the major sands are regional aquifers that extend far beyond the boundaries of the salt-dome basin. Because of multiple aquifers, structural deformation, and variations in the hydraulic characteristics of cap rock, the ground-water hydrology around a salt dome may be highly complex. The Sparta Sand is the most productive and heavily used regional aquifer. It is either penetrated by or overlies most of the domes. A fluid entering the Sparta flow system would move toward one of the pumping centers, all at or near municipalities that pump from the Sparta. Movement could be toward surface drainage where local geologic and hydrologic conditions permit leakage to the surface or to a surficial aquifer. (Woodard-USGS)

  2. Method for calcining radioactive wastes

    DOEpatents

    Bjorklund, William J.; McElroy, Jack L.; Mendel, John E.

    1979-01-01

    This invention relates to a method for the preparation of radioactive wastes in a low leachability form by calcining the radioactive waste on a fluidized bed of glass frit, removing the calcined waste to melter to form a homogeneous melt of the glass and the calcined waste, and then solidifying the melt to encapsulate the radioactive calcine in a glass matrix.

  3. Salton Sea Geothermal Field, California, as a near-field natural analog of a radioactive waste repository in salt

    SciTech Connect

    Elders, W.A.; Cohen, L.H.

    1983-11-01

    Since high concentrations of radionuclides and high temperatures are not normally encountered in salt domes or beds, finding an exact geologic analog of expected near-field conditions in a mined nuclear waste repository in salt will be difficult. The Salton Sea Geothermal Field, however, provides an opportunity to investigate the migration and retardation of naturally occurring U, Th, Ra, Cs, Sr and other elements in hot brines which have been moving through clay-rich sedimentary rocks for up to 100,000 years. The more than thirty deep wells drilled in this field to produce steam for electrical generation penetrate sedimentary rocks containing concentrated brines where temperatures reach 365/sup 0/C at only 2 km depth. The brines are primarily Na, K, Ca chlorides with up to 25% of total dissolved solids; they also contain high concentrations of metals such as Fe, Mn, Li, Zn, and Pb. This report describes the geology, geophysics and geochemistry of this system as a prelude to a study of the mobility of naturally occurring radionuclides and radionuclide analogs within it. The aim of this study is to provide data to assist in validating quantitative models of repository behavior and to use in designing and evaluating waste packages and engineered barriers. 128 references, 33 figures, 13 tables.

  4. Radioactive waste material disposal

    DOEpatents

    Forsberg, Charles W.; Beahm, Edward C.; Parker, George W.

    1995-01-01

    The invention is a process for direct conversion of solid radioactive waste, particularly spent nuclear fuel and its cladding, if any, into a solidified waste glass. A sacrificial metal oxide, dissolved in a glass bath, is used to oxidize elemental metal and any carbon values present in the waste as they are fed to the bath. Two different modes of operation are possible, depending on the sacrificial metal oxide employed. In the first mode, a regenerable sacrificial oxide, e.g., PbO, is employed, while the second mode features use of disposable oxides such as ferric oxide.

  5. Radioactive waste material disposal

    DOEpatents

    Forsberg, C.W.; Beahm, E.C.; Parker, G.W.

    1995-10-24

    The invention is a process for direct conversion of solid radioactive waste, particularly spent nuclear fuel and its cladding, if any, into a solidified waste glass. A sacrificial metal oxide, dissolved in a glass bath, is used to oxidize elemental metal and any carbon values present in the waste as they are fed to the bath. Two different modes of operation are possible, depending on the sacrificial metal oxide employed. In the first mode, a regenerable sacrificial oxide, e.g., PbO, is employed, while the second mode features use of disposable oxides such as ferric oxide. 3 figs.

  6. Disposal of NORM waste in salt caverns

    SciTech Connect

    Veil, J.A.; Smith, K.P.; Tomasko, D.; Elcock, D.; Blunt, D.; Williams, G.P.

    1998-07-01

    Some types of oil and gas production and processing wastes contain naturally occurring radioactive materials (NORM). If NORM is present at concentrations above regulatory levels in oil field waste, the waste requires special disposal practices. The existing disposal options for wastes containing NORM are limited and costly. This paper evaluates the legality, technical feasibility, economics, and human health risk of disposing of NORM-contaminated oil field wastes in salt caverns. Cavern disposal of NORM waste is technically feasible and poses a very low human health risk. From a legal perspective, there are no fatal flaws that would prevent a state regulatory agency from approving cavern disposal of NORM. On the basis of the costs charged by caverns currently used for disposal of nonhazardous oil field waste (NOW), NORM waste disposal caverns could be cost competitive with existing NORM waste disposal methods when regulatory agencies approve the practice.

  7. Radioactive waste isolation in salt: Peer review of the Office of Nuclear Waste Isolation's draft report on an issues hierarchy and data needs for site characterization

    SciTech Connect

    Harrison, W.; Fenster, D.F.; Ditmars, J.D.; Paddock, R.A.; Rote, D.M.; Hambley, D.F.; Seitz, M.G.; Hull, A.B.

    1986-12-01

    At the request of the Salt Repository Project (SRPO), Argonne National Laboratory conducted an independent peer review of a report by the Battelle Office of Nuclear Waste Isolation entitled ''Salt Repository Project Issues Hierarchy and Data Needs for Site Characterization (Draft).'' This report provided a logical structure for evaluating the outstanding questions (issues) related to selection and licensing of a site as a high-level waste repository. It also provided a first estimate of the information and data necessary to answer or resolve those questions. As such, this report is the first step in developing a strategy for site characterization. Microfiche copies of ''Draft Issues Hierarchy, Resolution Strategy, and Information Needs for Site Characterization and Environmental/Socioeconomic Evaluation - July, 1986'' and ''Issues Hierarchy and Data Needs for Site Characterization - February, 1985'' are included in the back pocket of this report.

  8. Radioactive waste isolation in salt: peer review of Westinghouse Electric Corporation's report on reference conceptual designs for a repository waste package

    SciTech Connect

    Rote, D.M.; Hull, A.B.; Was, G.S.; Macdonald, D.D.; Wilde, B.E.; Russell, J.E.; Kruger, J.; Harrison, W.; Hambley, D.F.

    1985-10-01

    This report documents the findings of the peer panel constituted by Argonne National Laboratory to review Region A of Westinghouse Electric Corporation's report entitled Waste Package Reference Conceptual Designs for a Repository in Salt. The panel determined that the reviewed report does not provide reasonable assurance that US Nuclear Regulatory Commission (NRC) requirements for waste packages will be met by the proposed design. It also found that it is premature to call the design a ''reference design,'' or even a ''reference conceptual design.'' This review report provides guidance for the preparation of a more acceptable design document.

  9. Long-term cement corrosion in chloride-rich solutions relevant to radioactive waste disposal in rock salt - Leaching experiments and thermodynamic simulations

    NASA Astrophysics Data System (ADS)

    Bube, C.; Metz, V.; Bohnert, E.; Garbev, K.; Schild, D.; Kienzler, B.

    Low- and intermediate-level radioactive wastes are frequently solidified in a cement matrix. In a potential repository for nuclear wastes, the cementitious matrix is altered upon contact with solution and the resulting secondary phases may provide for significant retention of the radionuclides incorporated in the wastes. In order to assess the secondary phases formed upon corrosion in chloride-rich solutions, which are relevant for nuclear waste disposal in rock salt, leaching experiments were performed. Conventional laboratory batch experiments using powdered hardened cement paste in MgCl2-rich solutions were left to equilibrate for up to three years and full-scale cemented waste products were exposed to NaCl-rich and MgCl2-rich solutions for more than twenty years, respectively. Solid phase analyses revealed that corrosion of hardened cement in MgCl2-rich solutions advanced faster than in NaCl-rich solutions due to the extensive exchange of Mg from solution against Ca from the cementitious solid. Thermodynamic equilibrium simulations compared well to results at the final stages of the respective experiments indicating that close to equilibrium conditions were reached. At high cement product to brine ratios (>0.65 g mL-1), the solution composition in the laboratory-scale experiments was close to that of the full-scale experiments (cement to brine ratio of 2.5 g mL-1) in the MgCl2 systems. The present study demonstrates the applicability of thermodynamic methods used in this approach to adequately describe full-scale long-term experiments with cemented waste simulates.

  10. PROCESSING OF RADIOACTIVE WASTE

    DOEpatents

    Johnson, B.M. Jr.; Barton, G.B.

    1961-11-14

    A process for treating radioactive waste solutions prior to disposal is described. A water-soluble phosphate, borate, and/or silicate is added. The solution is sprayed with steam into a space heated from 325 to 400 deg C whereby a powder is formed. The powder is melted and calcined at from 800 to 1000 deg C. Water vapor and gaseous products are separated from the glass formed. (AEC)

  11. 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.

  12. Radioactive waste isolation in salt: Peer review of the Fluor Technology, Inc. , report and position paper concerning waste emplacement mode and its effect on repository conceptual design

    SciTech Connect

    Hambley, D.F.; Russell, J.E.; Whitfield, R.G.; McGinnis, L.D.; Harrison, W.; Jacoby, C.H.; Bump, T.R.; Mraz, D.Z.; Busch, J.S.; Fischer, L.E.

    1987-02-01

    Recommendations for revising the Fluor Technology, Inc., draft position paper entitled Evaluation of Waste Emplacement Mode and the final report entitled Waste Package/Repository Impact Study include: reevaluate the relative rankings for the various emplacement modes; delete the following want objectives: maximize ability to locate the package horizon because sufficient flexibility exists to locate rooms in the relatively clean San Andres Unit 4 Salt and maximize far-field geologic integrity during retrieval because by definition the far field will be unaffected by thermal and stress perturbations caused by remining; give greater emphasis to want objectives regarding cost and use of present technology; delete the following statements from pages 1-1 and 1-2 of the draft position paper: ''No thought or study was given to the impacts of this configuration (vertical emplacement) on repository construction or short and long-term performance of the site'' and ''Subsequent salt repository designs adopted the vertical emplacement configuration as the accepted method without further evaluation.''; delete App. E and lines 8-17 of page 1-4 of the draft position paper because they are inappropriate; adopt a formal decision-analysis procedure for the 17 identified emplacement modes; revise App. F of the impact study to more accurately reflect current technology; consider designing the underground layout to take advantage of stress-relief techniques; consider eliminating reference to fuel assemblies <10 yr ''out-of-reactor''; model the temperature distribution, assuming that the repository is constructed in an infinitely large salt body; state that the results of creep analyses must be considered tentative until they can be validated by in situ measurements; and reevaluate the peak radial stresses on the waste package so that the calculated stress conditions more closely approximate expected in situ conditions.

  13. Radioactive waste isolation in salt: peer review of the Office of Nuclear Waste Isolation's reports on multifactor life testing of waste package materials

    SciTech Connect

    McPheeters, C.C.; Harrison, W.; Ditmars, J.D.; Lerman, A.; Rote, D.M.; Edgar, D.E.; Hambley, D.F.

    1984-09-01

    Two documents that provide the approaches in designing a test program to investigate uniform corrosion of low-carbon cash steel in a salt repository environment were reviewed. Recommendations are made by the Peer Review Panel for improving the two reports.

  14. Radioactive waste processing apparatus

    DOEpatents

    Nelson, Robert E.; Ziegler, Anton A.; Serino, David F.; Basnar, Paul J.

    1987-01-01

    Apparatus for use in processing radioactive waste materials for shipment and storage in solid form in a container is disclosed. The container includes a top, and an opening in the top which is smaller than the outer circumference of the container. The apparatus includes an enclosure into which the container is placed, solution feed apparatus for adding a solution containing radioactive waste materials into the container through the container opening, and at least one rotatable blade for blending the solution with a fixing agent such as cement or the like as the solution is added into the container. The blade is constructed so that it can pass through the opening in the top of the container. The rotational axis of the blade is displaced from the center of the blade so that after the blade passes through the opening, the blade and container can be adjusted so that one edge of the blade is adjacent the cylindrical wall of the container, to insure thorough mixing. When the blade is inside the container, a substantially sealed chamber is formed to contain vapors created by the chemical action of the waste solution and fixant, and vapors emanating through the opening in the container.

  15. Radioactive Waste Management

    NASA Astrophysics Data System (ADS)

    Baisden, P. A.; Atkins-Duffin, C. E.

    Issues related to the management of radioactive wastes are presented with specific emphasis on high-level wastes generated as a result of energy and materials production using nuclear reactors. The final disposition of these high-level wastes depends on which nuclear fuel cycle is pursued, and range from once-through burning of fuel in a light water reactor followed by direct disposal in a geologic repository to more advanced fuel cycles (AFCs) where the spent fuel is reprocessed or partitioned to recover the fissile material (primarily 235U and 239Pu) as well as the minor actinides (MAs) (neptunium, americium, and curium) and some long-lived fission products (e.g., 99Tc and 129I). In the latter fuel cycle, the fissile materials are recycled through a reactor to produce more energy, the short-lived fission products are vitrified and disposed of in a geologic repository, and the minor actinides and long-lived fission products are converted to less radiotoxic or otherwise stable nuclides by a process called transmutation. The advantages and disadvantages of the various fuel cycle options and the challenges to the management of nuclear wastes they represent are discussed.

  16. Method for immobilizing mixed waste chloride salts containing radionuclides and other hazardous wastes

    DOEpatents

    Lewis, Michele A.; Johnson, Terry R.

    1993-01-01

    The invention is a method for the encapsulation of soluble radioactive waste chloride salts containing radionuclides such as strontium, cesium and hazardous wastes such as barium so that they may be permanently stored without future threat to the environment. The process consists of contacting the salts containing the radionuclides and hazardous wastes with certain zeolites which have been found to ion exchange with the radionuclides and to occlude the chloride salts so that the resulting product is leach resistant.

  17. Method for immobilizing mixed waste chloride salts containing radionuclides and other hazardous wastes

    DOEpatents

    Lewis, Michele A.; Johnson, Terry R.

    1993-09-07

    The invention is a method for the encapsulation of soluble radioactive waste chloride salts containing radionuclides such as strontium, cesium and hazardous wastes such as barium so that they may be permanently stored without future threat to the environment. The process consists of contacting the salts containing the radionuclides and hazardous wastes with certain zeolites which have been found to ion exchange with the radionuclides and to occlude the chloride salts so that the resulting product is leach resistant.

  18. High-Level Radioactive Waste.

    ERIC Educational Resources Information Center

    Hayden, Howard C.

    1995-01-01

    Presents a method to calculate the amount of high-level radioactive waste by taking into consideration the following factors: the fission process that yields the waste, identification of the waste, the energy required to run a 1-GWe plant for one year, and the uranium mass required to produce that energy. Briefly discusses waste disposal and…

  19. Radioactive waste processing apparatus

    DOEpatents

    Nelson, R.E.; Ziegler, A.A.; Serino, D.F.; Basnar, P.J.

    1985-08-30

    Apparatus for use in processing radioactive waste materials for shipment and storage in solid form in a container is disclosed. The container includes a top, and an opening in the top which is smaller than the outer circumference of the container. The apparatus includes an enclosure into which the container is placed, solution feed apparatus for adding a solution containing radioactive waste materials into the container through the container opening, and at least one rotatable blade for blending the solution with a fixing agent such as cement or the like as the solution is added into the container. The blade is constructed so that it can pass through the opening in the top of the container. The rotational axis of the blade is displaced from the center of the blade so that after the blade passes through the opening, the blade and container can be adjusted so that one edge of the blade is adjacent the cylindrical wall of the container, to insure thorough mixing. When the blade is inside the container, a substantially sealed chamber is formed to contain vapors created by the chemical action of the waste solution and fixant, and vapors emanating through the opening in the container. The chamber may be formed by placing a removable extension over the top of the container. The extension communicates with the apparatus so that such vapors are contained within the container, extension and solution feed apparatus. A portion of the chamber includes coolant which condenses the vapors. The resulting condensate is returned to the container by the force of gravity.

  20. Polyethylene encapsulation of molten salt oxidation mixed low-level radioactive salt residues

    SciTech Connect

    Lageraaen, P.R.; Kalb, P.D.; Grimmett, D.L.; Gay, R.L.; Newman, C.D.

    1995-10-01

    A limited scope treatability study was conducted for polyethylene encapsulation of salt residues generated by a Molten Salt Oxidation (MSO) technology demonstration at the Energy Technology Engineering Center (ETEC), operated by Rockwell International for the US Department of Energy (DOE). During 1992 and 1993, ETEC performed a demonstration with a prototype MSO unit and treated approximately 50 gallons of mixed waste comprised of radioactively contaminated oils produced by hot cell operations. A sample of the mixed waste contaminated spent salt was used during the BNL polyethylene encapsulation treatability study. A nominal waste loading of 50 wt % was successfully processed and waste form test specimens were made for Toxicity Characteristic Leaching Procedure (TCLP) testing. The encapsulated product was compared with base-line TCLP results for total chromium and was found to be well within allowable EPA guidelines.

  1. PROCESSING OF RADIOACTIVE WASTE

    DOEpatents

    Allemann, R.T.; Johnson, B.M. Jr.

    1961-10-31

    A process for concentrating fission-product-containing waste solutions from fuel element processing is described. The process comprises the addition of sugar to the solution, preferably after it is made alkaline; spraying the solution into a heated space whereby a dry powder is formed; heating the powder to at least 220 deg C in the presence of oxygen whereby the powder ignites, the sugar is converted to carbon, and the salts are decomposed by the carbon; melting the powder at between 800 and 900 deg C; and cooling the melt. (AEC) antidiuretic hormone from the blood by the liver. Data are summarized from the following: tracer studies on cardiovascular functions; the determination of serum protein-bound iodine; urinary estrogen excretion in patients with arvanced metastatic mammary carcinoma; the relationship between alheroclerosis aad lipoproteins; the physical chemistry of lipoproteins; and factors that modify the effects of densely ionizing radia

  2. Crystallization of sodium nitrate from radioactive waste

    SciTech Connect

    Krapukhin, V.B.; Krasavina, E.P. Pikaev, A.K.

    1997-07-01

    From the 1940s to the 1980s, the Institute of Physical Chemistry of the Russian Academy of Sciences (IPC/RAS) conducted research and development on processes to separate acetate and nitrate salts and acetic acid from radioactive wastes by crystallization. The research objective was to decrease waste volumes and produce the separated decontaminated materials for recycle. This report presents an account of the IPC/RAS experience in this field. Details on operating conditions, waste and product compositions, decontamination factors, and process equipment are described. The research and development was generally related to the management of intermediate-level radioactive wastes. The waste solutions resulted from recovery and processing of uranium, plutonium, and other products from irradiated nuclear fuel, neutralization of nuclear process solutions after extractant recovery, regeneration of process nitric acid, equipment decontamination, and other radiochemical processes. Waste components include nitric acid, metal nitrate and acetate salts, organic impurities, and surfactants. Waste management operations generally consist of two stages: volume reduction and processing of the concentrates for storage, solidification, and disposal. Filtration, coprecipitation, coagulation, evaporation, and sorption were used to reduce waste volume. 28 figs., 40 tabs.

  3. Sorting of solid radioactive wastes

    SciTech Connect

    Marek, J.; Pecival, I.; Hejtman, J.; Wildman, J.; Cechak, T.

    1993-12-31

    In the nuclear power plants solid radioactive wastes are produced during regular operation and during small repairs. It is necessary to sort them into the highly contaminated wastes for which a special procedure for storage is necessary and waste that is not radioactive and can be stored in the environment under specific regulations. The aim of the project was to propose and to construct equipment, which is able to sort the waste with a high degree of reliability and to distinguish highly contaminated wastes from wastes which are less dangerous to the environment. The sensitivity of the detection system was tested by a mathematical model. The radioactive wastes from the primary part of the nuclear power plant can have three composition types. Details of the composition of the radioisotopes mixture are presented.

  4. Radioactive waste material melter apparatus

    DOEpatents

    Newman, Darrell F.; Ross, Wayne A.

    1990-01-01

    An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another.

  5. Radioactive waste material melter apparatus

    DOEpatents

    Newman, D.F.; Ross, W.A.

    1990-04-24

    An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another. 8 figs.

  6. Geochemical aspects of radioactive waste disposal

    SciTech Connect

    Brookins, D.G.

    1984-01-01

    The book addresses various topics related to the geochemistry of waste disposal: natural radioactivity, kinds of radioactive waste, details of possible disposal sites, low-level waste, uranium mill tailing, natural analogs, waste forms, and engineered barriers. Emphasis throughout is on the importance of natural analogs, the behavior of elements resembling those to be put in a waste repository as they occur in natural situations where the temperature, pressure, and movement of ground water are similar to those expected near a repository. The author is convinced that conclusions drawn from the study of analog elements are directly applicable to predictions about radionuclide behavior, and that the observed near-immobility of most of these elements in comparable geologic environments is good evidence that radioactive waste can be disposed of underground with negligible effects on the biosphere. Much of his own research has been in this area, and the best parts of the book are the descriptions of his work on trace elements in the salt minerals at the Waste Isolation Pilot Plant in southeastern New Mexico, on the movement of radionuclides and their daughter elements from the famous Precambrian reactor at Oklahoma in Gabon, and on the distribution of analog elements in rocks near the contacts of igneous intrusions.

  7. Mixed Waste Salt Encapsulation Using Polysiloxane - Final Report

    SciTech Connect

    Miller, C.M.; Loomis, G.G.; Prewett, S.W.

    1997-11-01

    A proof-of-concept experimental study was performed to investigate the use of Orbit Technologies polysiloxane grouting material for encapsulation of U.S. Department of Energy mixed waste salts leading to a final waste form for disposal. Evaporator pond salt residues and other salt-like material contaminated with both radioactive isotopes and hazardous components are ubiquitous in the DOE complex and may exceed 250,000,000 kg of material. Current treatment involves mixing low waste percentages (less than 10% by mass salt) with cement or costly thermal treatment followed by cementation to the ash residue. The proposed technology involves simple mixing of the granular salt material (with relatively high waste loadings-greater than 50%) in a polysiloxane-based system that polymerizes to form a silicon-based polymer material. This study involved a mixing study to determine optimum waste loadings and compressive strengths of the resultant monoliths. Following the mixing study, durability testing was performed on promising waste forms. Leaching studies including the accelerated leach test and the toxicity characteristic leaching procedure were also performed on a high nitrate salt waste form. In addition to this testing, the waste form was examined by scanning electron microscope. Preliminary cost estimates for applying this technology to the DOE complex mixed waste salt problem is also given.

  8. SELF SINTERING OF RADIOACTIVE WASTES

    DOEpatents

    McVay, T.N.; Johnson, J.R.; Struxness, E.G.; Morgan, K.Z.

    1959-12-29

    A method is described for disposal of radioactive liquid waste materials. The wastes are mixed with clays and fluxes to form a ceramic slip and disposed in a thermally insulated container in a layer. The temperature of the layer rises due to conversion of the energy of radioactivity to heat boillng off the liquid to fomn a dry mass. The dry mass is then covered with thermal insulation, and the mass is self-sintered into a leach-resistant ceramic cake by further conversion of the energy of radioactivity to heat.

  9. Defense waste salt disposal at the Savannah River Plant. [Cement-based waste form, saltstone

    SciTech Connect

    Langton, C A; Dukes, M D

    1984-01-01

    A cement-based waste form, saltstone, has been designed for disposal of Savannah River Plant low-level radioactive salt waste. The disposal process includes emplacing the saltstone in engineered trenches above the water table but below grade at SRP. Design of the waste form and disposal system limits the concentration of salts and radionuclides in the groundwater so that EPA drinking water standards will not be exceeded at the perimeter of the disposal site. 10 references, 4 figures, 3 tables.

  10. Radioactive waste shredding: Preliminary evaluation

    SciTech Connect

    Soelberg, N.R.; Reimann, G.A.

    1994-07-01

    The critical constraints for sizing solid radioactive and mixed wastes for subsequent thermal treatment were identified via a literature review and a survey of shredding equipment vendors. The types and amounts of DOE radioactive wastes that will require treatment to reduce the waste volume, destroy hazardous organics, or immobilize radionuclides and/or hazardous metals were considered. The preliminary steps of waste receipt, inspection, and separation were included because many potential waste treatment technologies have limits on feedstream chemical content, physical composition, and particle size. Most treatment processes and shredding operations require at least some degree of feed material characterization. Preliminary cost estimates show that pretreatment costs per unit of waste can be high and can vary significantly, depending on the processing rate and desired output particle size.

  11. 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.

  12. Radioactive waste isolation in salt:

    SciTech Connect

    Ditmars, J.D.; Baecher, G.B.; Edgar, D.E.; Dowding, C.H.

    1988-03-01

    The approach presumes that measurements are undertaken to support performance predictions. A quantitative performance objective like groundwater travel time is compared with performance predictions. The approach recognizes that such predictions are uncertain because the measurements upon which they are based are uncertain. The effectiveness of measurement activities is quantified by an index, ..beta.., that reflects the number of standard deviations separating the best estimate of performance from the performance objective. Measurements that reduce the uncertainty in predictions lead to increased values of ..beta... Evaluating ..beta.. for a particular measurement scheme requires identifying the measured quantities that significantly affect prediction uncertainty. Sources of uncertainty are spatial variation, noise, estimation error, and measurement bias. Changing the measurement scheme to increase ..beta.. increases the likelihood of a performance objective being achieved or exceeded. The application of the ..beta..-index method to the Richton dome site in Mississippi focuses on uncertainties in hydraulic conductivity data in relation to groundwater travel time predictions. The ..beta.. values for four different measurement schemes for hydraulic conductivity are determined. 44 refs., 14 figs., 15 tabs.

  13. Effects of Heat Generation on Nuclear Waste Disposal in Salt

    NASA Astrophysics Data System (ADS)

    Clayton, D. J.

    2008-12-01

    Disposal of nuclear waste in salt is an established technology, as evidenced by the successful operations of the Waste Isolation Pilot Plant (WIPP) since 1999. The WIPP is located in bedded salt in southeastern New Mexico and is a deep underground facility for transuranic (TRU) nuclear waste disposal. There are many advantages for placing radioactive wastes in a geologic bedded-salt environment. One desirable mechanical characteristic of salt is that it flows plastically with time ("creeps"). The rate of salt creep is a strong function of temperature and stress differences. Higher temperatures and deviatoric stresses increase the creep rate. As the salt creeps, induced fractures may be closed and eventually healed, which then effectively seals the waste in place. With a backfill of crushed salt emplaced around the waste, the salt creep can cause the crushed salt to reconsolidate and heal to a state similar to intact salt, serving as an efficient seal. Experiments in the WIPP were conducted to investigate the effects of heat generation on the important phenomena and processes in and around the repository (Munson et al. 1987; 1990; 1992a; 1992b). Brine migration towards the heaters was induced from the thermal gradient, while salt creep rates showed an exponential dependence on temperature. The project "Backfill and Material Behavior in Underground Salt Repositories, Phase II" (BAMBUS II) studied the crushed salt backfill and material behavior with heat generation at the Asse mine located near Remlingen, Germany (Bechthold et al. 2004). Increased salt creep rates and significant reconsolidation of the crushed salt were observed at the termination of the experiment. Using the data provided from both projects, exploratory modeling of the thermal-mechanical response of salt has been conducted with varying thermal loading and waste spacing. Increased thermal loading and decreased waste spacing drive the system to higher temperatures, while both factors are desired to

  14. 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)

  15. Vitrification of radioactive wastes

    SciTech Connect

    McIntosh, T.; Bibler, N.; Weber, T.

    1993-12-31

    The United States and Russia are conducting joint technology exchanges on the vitrification of high-level wastes at the Chelyabinsk-65 site in Russia and the Savannah River site in the United States. These activities are part of a larger program of exchanges provided for by a Memorandum of Cooperation (MOC) between the US and the USSR in the fields of Environmental Restoration and Waste Management that was signed in the fall of 1990. An agreement to exchange information in the vitrification area was concluded during a visit by a DOE delegation to Chelyabinsk-65 in the fall of 1991. Samples of simulated high-level waste glass and the leach test procedures used by each country were to be exchanged. The results of studies based on leaching the simulated high-level waste glasses, and potential future work in the area of vitrification, to include advanced concepts, will be discussed.

  16. Slag cement-low level radioactive waste forms at Savannah River Plant

    SciTech Connect

    Malek, R.I.A.; Roy, D.M.; Langton, C.A.

    1986-12-01

    A hydrated ceramic waste form, ''salt-stone,'' was designed for solidification and stabilization of Savannah River Plant (SRP) low level radioactive defense waste. This waste is a concentrated salt solution containing mainly sodium nitrate, nitrite, aluminate, sulfate, and hydroxide and has radioactivity. Ground, granulated blast furnace slag (a byproduct from the steel industry) was identified as a potential hydraulic ingredient for saltstone since its reactivity was found to be enhanced by the high alkalinity of the waste solution.

  17. Radioactive waste isolation in salt: peer review of the Office of Nuclear Waste Isolation's report on the Organic Geochemistry of Deep Groundwaters from the Palo Duro Basin, Texas

    SciTech Connect

    Fenster, D.F.; Brookins, D.G.; Harrison, W.; Seitz, M.G.; Lerman, A.; Stamoudis, V.C.

    1984-08-01

    This report summarizes Argonne's review of the Office of Nuclear Waste Isolation's (ONWI's) final report entitled The Organic Geochemistry of Deep Ground Waters from the Palo Duro Basin, Texas, dated September 1983. Recommendations are made for improving the ONWI report. The main recommendation is to make the text consistent with the title and with the objective of the project as stated in the introduction. Three alternatives are suggested to accomplish this.

  18. Public attitudes about radioactive waste

    SciTech Connect

    Bisconti, A.S.

    1992-12-31

    Public attitudes about radioactive waste are changeable. That is my conclusion from eight years of social science research which I have directed on this topic. The fact that public attitudes about radioactive waste are changeable is well-known to the hands-on practitioners who have opportunities to talk with the public and respond to their concerns-practitioners like Ginger King, who is sharing the podium with me today. The public`s changeability and open-mindedness are frequently overlooked in studies that focus narrowly on fear and dread. Such studies give the impression that the outlook for waste disposal solutions is dismal. I believe that impression is misleading, and I`d like to share research findings with you today that give a broader perspective.

  19. Radioactive Waste Management BasisApril 2006

    SciTech Connect

    Perkins, B K

    2011-08-31

    This Radioactive Waste Management Basis (RWMB) documents radioactive waste management practices adopted at Lawrence Livermore National Laboratory (LLNL) pursuant to Department of Energy (DOE) Order 435.1, Radioactive Waste Management. 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. Radioactive waste treatment technologies and environment

    SciTech Connect

    HORVATH, Jan; KRASNY, Dusan

    2007-07-01

    The radioactive waste treatment and conditioning are the most important steps in radioactive waste management. At the Slovak Electric, plc, a range of technologies are used for the processing of radioactive waste into a form suitable for disposal in near surface repository. These technologies operated by JAVYS, PLc. Nuclear and Decommissioning Company, PLc. Jaslovske Bohunice are described. Main accent is given to the Bohunice Radwaste Treatment and Conditioning Centre, Bituminization plant, Vitrification plant, and Near surface repository of radioactive waste in Mochovce and their operation. Conclusions to safe and effective management of radioactive waste in the Slovak Republic are presented. (authors)

  1. Testing of stripping columns for the removal of benzene from aqueous radioactive salt solution

    SciTech Connect

    Georgeton, G.K.; Taylor, G.A.; Gaughan, T.P.

    1995-06-27

    Radioactive high level wastes (HLW) generated from production of special nuclear materials at the Savannah River Site (SRS) are held in interim storage in 51 underground, million gallon tanks. Radioactive cesium ({sup 137}Cs) is segregated by evaporation of aqueous waste solution for interim storage in a salt matrix comprised of Na and K salts or in concentrated salt solution. The saltcake will be dissolved and {sup 137}Cs will be separated from the nonradioactive salts in solution in the In-Tank Precipitation (ITP) Process. The cesium will be combined with other radioactive species and glass formers to be melted and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). The salt solution remaining after decontamination in the ITP process will be incorporated into grout for disposal at the site`s Saltstone facility. In the ITP facility, sodium tetraphenylborate (STPB) will be added to precipitate the cesium. Potassium in the waste solution also reacts with STPB and precipitates. Due to radiolytic and chemical degradation of the tetraphenylborate (TPB) precipitate, benzene is generated. The benzene dissolves into the decontaminated salt solution (DSS) and into water (WW) used to {open_quotes}wash{close_quotes} the precipitate to lower the soluble salt content of the slurry. Safety and processing requirements for disposal of the DSS and for temporary storage of the WW dictate that the benzene concentration be reduced.

  2. Vitrification of hazardous and radioactive wastes

    SciTech Connect

    Bickford, D.F.; Schumacher, R.

    1995-12-31

    Vitrification offers many attractive waste stabilization options. Versatility of waste compositions, as well as the inherent durability of a glass waste form, have made vitrification the treatment of choice for high-level radioactive wastes. Adapting the technology to other hazardous and radioactive waste streams will provide an environmentally acceptable solution to many of the waste challenges that face the public today. This document reviews various types and technologies involved in vitrification.

  3. Blending Of Radioactive Salt Solutions In Million Gallon Tanks

    SciTech Connect

    Leishear, Robert A.; Lee, Si Y.; Fowley, Mark D.; Poirier, Michael R.

    2012-12-10

    Research was completed at Savannah River National Laboratory (SRNL) to investigate processes related to the blending of radioactive, liquid waste, salt solutions in 4920 cubic meter, 25.9 meter diameter storage tanks. One process was the blending of large salt solution batches (up to 1135 ? 3028 cubic meters), using submerged centrifugal pumps. A second process was the disturbance of a settled layer of solids, or sludge, on the tank bottom. And a third investigated process was the settling rate of sludge solids if suspended into slurries by the blending pump. To investigate these processes, experiments, CFD models (computational fluid dynamics), and theory were applied. Experiments were performed using simulated, non-radioactive, salt solutions referred to as supernates, and a layer of settled solids referred to as sludge. Blending experiments were performed in a 2.44 meter diameter pilot scale tank, and flow rate measurements and settling tests were performed at both pilot scale and full scale. A summary of the research is presented here to demonstrate the adage that, ?One good experiment fixes a lot of good theory?. Experimental testing was required to benchmark CFD models, or the models would have been incorrectly used. In fact, CFD safety factors were established by this research to predict full-scale blending performance. CFD models were used to determine pump design requirements, predict blending times, and cut costs several million dollars by reducing the number of required blending pumps. This research contributed to DOE missions to permanently close the remaining 47 of 51 SRS waste storage tanks.

  4. Blending of Radioactive Salt Solutions in Million Gallon Tanks - 13002

    SciTech Connect

    Leishear, Robert A.; Lee, Si Y.; Fowley, Mark D.; Poirier, Michael R.

    2013-07-01

    Research was completed at Savannah River National Laboratory (SRNL) to investigate processes related to the blending of radioactive, liquid waste, salt solutions in 4920 cubic meter, 25.9 meter diameter storage tanks. One process was the blending of large salt solution batches (up to 1135 - 3028 cubic meters), using submerged centrifugal pumps. A second process was the disturbance of a settled layer of solids, or sludge, on the tank bottom. And a third investigated process was the settling rate of sludge solids if suspended into slurries by the blending pump. To investigate these processes, experiments, CFD models (computational fluid dynamics), and theory were applied. Experiments were performed using simulated, non-radioactive, salt solutions referred to as supernates, and a layer of settled solids referred to as sludge. Blending experiments were performed in a 2.44 meter diameter pilot scale tank, and flow rate measurements and settling tests were performed at both pilot scale and full scale. A summary of the research is presented here to demonstrate the adage that, 'One good experiment fixes a lot of good theory'. Experimental testing was required to benchmark CFD models, or the models would have been incorrectly used. In fact, CFD safety factors were established by this research to predict full-scale blending performance. CFD models were used to determine pump design requirements, predict blending times, and cut costs several million dollars by reducing the number of required blending pumps. This research contributed to DOE missions to permanently close the remaining 47 of 51 SRS waste storage tanks. (authors)

  5. Radioactive Waste Management BasisSept 2001

    SciTech Connect

    Goodwin, S S

    2011-08-31

    This Radioactive Waste Management Basis (RWMB) documents radioactive waste management practices adopted at Lawrence Livermore National Laboratory (LLNL) pursuant to Department of Energy (DOE) Order 435.1, Radioactive Waste Management. The purpose of this RWMB 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 meeeting the requirements of DOE Order 435.1 deviate from the DOE Manual 435.1-1 and Implementation Guide.

  6. System for radioactive waste cementation

    SciTech Connect

    Dmitriev, S.A.; Barinov, A.S.; Varlakov, A.P.; Volkov, A.S.; Karlin, S.V.

    1995-12-31

    NPP, research reactors and radiochemical enterprises produce a great amount of liquid radioactive waste (LRW). One of the methods of LRW solidification is cementation. The recent investigations demonstrated possible inclusion of sufficient amount of waste in the cement matrix (up to 20--30 mass% on dry residue). In this case the cementation process becomes competitive with bituminization process, where the matrix can include 40--50 mass% and the solidified product volume is equal to the volume, obtained by cementation. Additionally, the cement matrix in contrast with the bituminous one is unburnable. Many countries are investigating the cementation process. The main idea governing technological process is the waste and cement mixing method and type of mixer. In world practice some principal types of cementation systems are used. The paper describes the SIA Radon industrial plant in Moscow.

  7. Method for solidifying radioactive wastes

    SciTech Connect

    Dippel, T.; Loida, A.

    1985-08-13

    A process is claimed for solidifying radioactive wastes by producing compact blocks which are to be disposed in transporting or permanent storage containers. The compact blocks are produced from prefabricated ceramic tablets which contain radioactive substances and a matrix which continuously surrounds these ceramic tablets and is solid in its final state. Glass powder or a mixture of oxidic non-clay minerals or a mixture of both is used as the matrix material. The ceramic tablets and the matrix material are filled into the container and are compressed. The resulting compressed mixture is heated to a temperature in the range from 1423/sup 0/ K. to 1623/sup 0/ K., is held at this temperate range for one to three hours, and is finally gradually cooled to room temperature.

  8. [Microbiological Aspects of Radioactive Waste Storage].

    PubMed

    Safonov, A V; Gorbunova, O A; German, K E; Zakharova, E V; Tregubova, V E; Ershov, B G; Nazina, T N

    2015-01-01

    The article gives information about the microorganisms inhabiting in surface storages of solid radioactive waste and deep disposal sites of liquid radioactive waste. It was shown that intensification of microbial processes can lead to significant changes in the chemical composition and physical state of the radioactive waste. It was concluded that the biogeochemical processes can have both a positive effect on the safety of radioactive waste storages (immobilization of RW macrocomponents, a decreased migration ability of radionuclides) and a negative one (biogenic gas production in subterranean formations and destruction of cement matrix). PMID:26310021

  9. Microbiological treatment of radioactive wastes

    SciTech Connect

    Francis, A.J.

    1992-12-31

    The ability of microorganisms which are ubiquitous throughout nature to bring about information of organic and inorganic compounds in radioactive wastes has been recognized. Unlike organic contaminants, metals cannot be destroyed, but must be either removed or converted to a stable form. Radionuclides and toxic metals in wastes may be present initially in soluble form or, after disposal may be converted to a soluble form by chemical or microbiological processes. The key microbiological reactions include (i) oxidation/reduction; (ii) change in pH and Eh which affects the valence state and solubility of the metal; (iii) production of sequestering agents; and (iv) bioaccumulation. All of these processes can mobilize or stabilize metals in the environment.

  10. Radioactive liquid waste treatment facility

    SciTech Connect

    Black, R.L.

    1984-07-01

    The Radioactive Liquid Waste Treatment Facility (RLWTF) at Argonne National Laboratory-West (ANL-W) in Idaho provides improved treatment for low-level aqueous waste compared to conventional systems. A unique, patented evaporated system is used in the RLWTF. SHADE (shielded hot air drum evaporator, US Patent No. 4,305,780) is a low-cost disposable unit constructed from standard components and is self-shielded. The results of testing and recent operations indicate that evaporation rates of 2 to 6 gph (8 to 23 L/h) can be achieved with a single unit housed in a standard 30-gal (114-L) drum container. The operating experience has confirmed the design evaporation rate of 60,000 gal (227,000 L) per year, using six SHADE's. 2 references, 2 figures, 2 tables.

  11. Molten salt processing of mixed wastes with offgas condensation

    SciTech Connect

    Cooper, J.F.; Brummond, W.; Celeste, J.; Farmer, J.; Hoenig, C.; Krikorian, O.H.; Upadhye, R. ); Gay, R.L.; Stewart, A.; Yosim, S. . Energy Systems Group)

    1991-05-13

    We are developing an advanced process for treatment of mixed wastes in molten salt media at temperatures of 700--1000{degrees}C. Waste destruction has been demonstrated in a single stage oxidation process, with destruction efficiencies above 99.9999% for many waste categories. The molten salt provides a heat transfer medium, prevents thermal surges, and functions as an in situ scrubber to transform the acid-gas forming components of the waste into neutral salts and immobilizes potentially fugitive materials by a combination of particle wetting, encapsulation and chemical dissolution and solvation. Because the offgas is collected and assayed before release, and wastes containing toxic and radioactive materials are treated while immobilized in a condensed phase, the process avoids the problems sometimes associated with incineration processes. We are studying a potentially improved modification of this process, which treats oxidizable wastes in two stages: pyrolysis followed by catalyzed molten salt oxidation of the pyrolysis gases at ca. 700{degrees}C. 15 refs., 5 figs., 1 tab.

  12. Integrated demonstration of molten salt oxidation with salt recycle for mixed waste treatment

    SciTech Connect

    Hsu, P.C.

    1997-11-01

    Molten Salt Oxidation (MSO) is a thermal, nonflame process that has the inherent capability of completely destroying organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. Lawrence Livermore National Laboratory (LLNL) has prepared a facility and constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are performed under carefully controlled (experimental) conditions. The system consists of a MSO processor with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. This integrated system was designed and engineered based on laboratory experience with a smaller engineering-scale reactor unit and extensive laboratory development on salt recycle and final forms preparation. In this paper we present design and engineering details of the system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is identification of the most suitable waste streams and waste types for MSO treatment.

  13. Oil field waste disposal in salt caverns: An information website

    SciTech Connect

    Tomasko, D.; Veil, J. A.

    1999-12-10

    Argonne National Laboratory has completed the construction of a Website for the US Department of Energy (DOE) that provides detailed information on salt caverns and their use for disposing of nonhazardous oil field wastes (NOW) and naturally occurring radioactive materials (NORM). Specific topics in the Website include the following: descriptions of salt deposits and salt caverns within the US, salt cavern construction methods, potential types of wastes, waste emplacement, regulatory issues, costs, carcinogenic and noncarcinogenic human health risks associated with postulated cavern release scenarios, new information on cavern disposal (e.g., upcoming meetings, regulatory issues, etc.), other studies supported by the National Petroleum Technology Office (NPTO) (e.g., considerations of site location, cavern stability, development issues, and bedded salt characterization in the Midland Basin), and links to other associated Web sites. In addition, the Website allows downloadable access to reports prepared on the topic that were funded by DOE. Because of the large quantities of NOW and NORM wastes generated annually by the oil industry, information presented on this Website is particularly interesting and valuable to project managers, regulators, and concerned citizens.

  14. 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.

  15. 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.

  16. Technology applications for radioactive waste minimization

    SciTech Connect

    Devgun, J.S.

    1994-07-01

    The nuclear power industry has achieved one of the most successful examples of waste minimization. The annual volume of low-level radioactive waste shipped for disposal per reactor has decreased to approximately one-fifth the volume about a decade ago. In addition, the curie content of the total waste shipped for disposal has decreased. This paper will discuss the regulatory drivers and economic factors for waste minimization and describe the application of technologies for achieving waste minimization for low-level radioactive waste with examples from the nuclear power industry.

  17. Development of iron phosphate ceramic waste form to immobilize radioactive waste solution

    SciTech Connect

    Choi, Jongkwon; Um, Wooyong; Choung, Sungwook

    2014-05-09

    The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl-KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions were 16 MPa and 19 MPa for LiCl-KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4,136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl-KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl-KCl eutectic salt and off-gas scrubber solutions with significant cost savings.

  18. Development of iron phosphate ceramic waste form to immobilize radioactive waste solution

    NASA Astrophysics Data System (ADS)

    Choi, Jongkwon; Um, Wooyong; Choung, Sungwook

    2014-09-01

    The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl-KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions were 16 MPa and 19 MPa for LiCl-KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl-KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl-KCl eutectic salt and off-gas scrubber solutions with significant cost savings.

  19. Laboratory scale vitrification of low-level radioactive nitrate salts and soils from the Idaho National Engineering Laboratory

    SciTech Connect

    Shaw, P.; Anderson, B.; Davis, D.

    1993-07-01

    INEL has radiologically contaminated nitrate salt and soil waste stored above and below ground in Pad A and the Acid Pit at the Radioactive Waste Management Complex. Pad A contain uranium and transuranic contaminated potassium and sodium nitrate salts generated from dewatered waste solutions at the Rocky Flats Plant. The Acid Pit was used to dispose of liquids containing waste mineral acids, uranium, nitrate, chlorinated solvents, and some mercury. Ex situ vitrification is a high temperature destruction of nitrates and organics and immobilizes hazardous and radioactive metals. Laboratory scale melting of actual radionuclides containing INEL Pad A nitrate salts and Acid Pit soils was performed. The salt/soil/additive ratios were varied to determine the range of glass compositions (resulted from melting different wastes); maximize mass and volume reduction, durability, and immobilization of hazardous and radioactive metals; and minimize viscosity and offgas generation for wastes prevalent at INEL and other DOE sites. Some mixtures were spiked with additional hazardous and radioactive metals. Representative glasses were leach tested and showed none. Samples spiked with transuranic showed low nuclide leaching. Wasteforms were two to three times bulk densities of the salt and soil. Thermally co-processing soils and salts is an effective remediation method for destroying nitrate salts while stabilizing the radiological and hazardous metals they contain. The measured durability of these low-level waste glasses approached those of high-level waste glasses. Lab scale vitrification of actual INEL contaminated salts and soils was performed at General Atomics Laboratory as part of the INEL Waste Technology Development and Environmental Restoration within the Buried Waste Integrated Demonstration Program.

  20. 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.

  1. Phosphate bonded solidification of radioactive incinerator wastes

    SciTech Connect

    Walker, B. W.; Langton, C. A.; Singh, D.

    1999-12-03

    The incinerator at the Department of Energy Savannah River Site burns low level radioactive and hazardous waste. Ash and scrubber system waste streams are generated during the incineration process. Phosphate Ceramic technology is being tested to verify the ash and scrubber waste streams can be stabilized using this solidification method. Acceptance criteria for the solid waste forms include leachability, bleed water, compression testing, and permeability. Other testing on the waste forms include x-ray diffraction and scanning electron microscopy.

  2. Hanford Site annual dangerous waste report: Volume 2, Generator dangerous waste report, radioactive mixed waste

    SciTech Connect

    1994-12-31

    This report contains information on radioactive mixed wastes at the Hanford Site. Information consists of shipment date, physical state, chemical nature, waste description, waste number, waste designation, weight, and waste designation.

  3. Hanford Site annual dangerous waste report: Volume 4, Waste Management Facility report, Radioactive mixed waste

    SciTech Connect

    1994-12-31

    This report contains information on radioactive mixed wastes at the Hanford Site. Information consists of shipment date, physical state, chemical nature, waste description, handling method and containment vessel, waste number, waste designation and amount of waste.

  4. Supplemental Cooling for Nitrate Salt Waste

    SciTech Connect

    Goldberg, Mitchell S.

    2015-08-19

    In July 2015, Los Alamos National Laboratory completed installation of a supplemental cooling system in the structure where remediated nitrate salt waste drums are stored. Although the waste currently is in a safe configuration and is monitored daily,controlling the temperature inside the structure adds another layer of protection for workers, the public,and the environment.This effort is among several layers of precautions designed to secure the waste.

  5. MIXING MODELING ANALYSIS FOR SRS SALT WASTE DISPOSITION

    SciTech Connect

    Lee, S.

    2011-01-18

    Nuclear waste at Savannah River Site (SRS) waste tanks consists of three different types of waste forms. They are the lighter salt solutions referred to as supernate, the precipitated salts as salt cake, and heavier fine solids as sludge. The sludge is settled on the tank floor. About half of the residual waste radioactivity is contained in the sludge, which is only about 8 percentage of the total waste volume. Mixing study to be evaluated here for the Salt Disposition Integration (SDI) project focuses on supernate preparations in waste tanks prior to transfer to the Salt Waste Processing Facility (SWPF) feed tank. The methods to mix and blend the contents of the SRS blend tanks were evalutaed to ensure that the contents are properly blended before they are transferred from the blend tank such as Tank 50H to the SWPF feed tank. The work consists of two principal objectives to investigate two different pumps. One objective is to identify a suitable pumping arrangement that will adequately blend/mix two miscible liquids to obtain a uniform composition in the tank with a minimum level of sludge solid particulate in suspension. The other is to estimate the elevation in the tank at which the transfer pump inlet should be located where the solid concentration of the entrained fluid remains below the acceptance criterion (0.09 wt% or 1200 mg/liter) during transfer operation to the SWPF. Tank 50H is a Waste Tank that will be used to prepare batches of salt feed for SWPF. The salt feed must be a homogeneous solution satisfying the acceptance criterion of the solids entrainment during transfer operation. The work described here consists of two modeling areas. They are the mixing modeling analysis during miscible liquid blending operation, and the flow pattern analysis during transfer operation of the blended liquid. The modeling results will provide quantitative design and operation information during the mixing/blending process and the transfer operation of the blended

  6. Salt caverns for oil field waste disposal.

    SciTech Connect

    Veil, J.; Ford, J.; Rawn-Schatzinger, V.; Environmental Assessment; RMC, Consultants, Inc.

    2000-07-01

    Salt caverns used for oil field waste disposal are created in salt formations by solution mining. When created, caverns are filled with brine. Wastes are introduced into the cavern by pumping them under low pressure. Each barrel of waste injected to the cavern displaces a barrel of brine to the surface. The brine is either used for drilling mud or is disposed of in an injection well. Figure 8 shows an injection pump used at disposal cavern facilities in west Texas. Several types of oil field waste may be pumped into caverns for disposal. These include drilling muds, drill cuttings, produced sands, tank bottoms, contaminated soil, and completion and stimulation wastes. Waste blending facilities are constructed at the site of cavern disposal to mix the waste into a brine solution prior to injection. Overall advantages of salt cavern disposal include a medium price range for disposal cost, large capacity and availability of salt caverns, limited surface land requirement, increased safety, and ease of establishment of individual state regulations.

  7. Fuel production from wastes using molten salts

    SciTech Connect

    Gay, R.L.; Barclay, K.M.; Grantham, L.F.; Yosim, S.J.

    1980-01-01

    The Rockwell International molten salt process for gasification of wastes with resource recovery has been shown here to be well-suited for the processing of a variety of wastes. A variety of waste forms may be processed, that is, solids, liquids, and solid-liquid mixtures. The process is suitable for applications which involve either small or large throughputs. The gasification medium, sodium carbonate, is stable, non-volatile, inexpensive, and nontoxic. Sulfur-containing pollutants are retained in the melt when sulfur-containing wastes are gasified. In the same manner, halogen-containing pollutants are retained during gasification of halogen-containing wastes. The gasification of a high-nitrogen-content waste (leather scraps) produces very little NO/sub x/ in the off-gas. Valuable minerals may be recovered by processing of the salt after gasification of mineral-laden wastes. In general, the molten salt process is best applied to waste materials involving potential pollutants (such as sulfur or chromium) or to wastes where gasification and resource recovery are important (such as the recovery of silver with simultaneous gasification of x-ray film).

  8. Evaluation of Terrorist Interest in Radioactive Wastes

    SciTech Connect

    McFee, J.N.; Langsted, J.M.; Young, M.E.; Day, J.E.

    2006-07-01

    Since September 11, 2001, intelligence gathered from Al Qaeda training camps in Afghanistan, and the ensuing terrorist activities, indicates nuclear material security concerns are valid. This paper reviews available information on sealed radioactive sources thought to be of interest to terrorists, and then examines typical wastes generated during environmental management activities to compare their comparative 'attractiveness' for terrorist diversion. Sealed radioactive sources have been evaluated in numerous studies to assess their security and attractiveness for use as a terrorist weapon. The studies conclude that tens of thousands of curies in sealed radioactive sources are available for potential use in a terrorist attack. This risk is mitigated by international efforts to find lost and abandoned sources and bring them under adequate security. However, radioactive waste has not received the same level of scrutiny to ensure security. This paper summarizes the activity and nature of radioactive sources potentially available to international terrorists. The paper then estimates radiation doses from use of radioactive sources as well as typical environmental restoration or decontamination and decommissioning wastes in a radioactive dispersal device (RDD) attack. These calculated doses indicate that radioactive wastes are, as expected, much less of a health risk than radioactive sources. The difference in radiation doses from wastes used in an RDD are four to nine orders of magnitude less than from sealed sources. We then review the International Atomic Energy Agency (IAEA) definition of 'dangerous source' in an adjusted comparison to common radioactive waste shipments generated in environmental management activities. The highest waste dispersion was found to meet only category 1-3.2 of the five step IAEA scale. A category '3' source by the IAEA standard 'is extremely unlikely, to cause injury to a person in the immediate vicinity'. The obvious conclusion of the

  9. Radioactive waste isolation in salt: Peer review of the Office of Nuclear Waste Isolation's draft report on a multifactor test design to investigate uniform corrosion of low-carbon steel

    SciTech Connect

    Paddock, R.A.; Lerman, A.; Ditmars, J.D.; Macdonald, D.D.; Peerenboom, J.P.; Was, G.S.; Harrison, W.

    1987-01-01

    This report documents Argonne National Laboratory's review of an internal technical memorandum prepared by Battelle Memorial Institute's Office of Nuclear Waste Isolation (ONWI) entitled Multifactor Test Design to Investigate Uniform Corrosion of Low-Carbon Steel in a Nuclear Waste Salt Repository Environment. The several major areas of concern identified by peer review panelists are important to the credibility of the test design proposed in the memorandum and are to adequately addressed there. These areas of concern, along with specific recommendations to improve their treatment, are discussed in detail in Sec. 2 of this report. The twenty recommendations, which were abstracted from those discussions, are presented essentially in the order in which they are introduced in Sec. 2.

  10. Chemical species of plutonium in Hanford radioactive tank waste

    SciTech Connect

    Barney, G.S.

    1997-10-22

    Large quantities of radioactive wastes have been generated at the Hanford Site over its operating life. The wastes with the highest activities are stored underground in 177 large (mostly one million gallon volume) concrete tanks with steel liners. The wastes contain processing chemicals, cladding chemicals, fission products, and actinides that were neutralized to a basic pH before addition to the tanks to prevent corrosion of the steel liners. Because the mission of the Hanford Site was to provide plutonium for defense purposes, the amount of plutonium lost to the wastes was relatively small. The best estimate of the amount of plutonium lost to all the waste tanks is about 500 kg. Given uncertainties in the measurements, some estimates are as high as 1,000 kg (Roetman et al. 1994). The wastes generally consist of (1) a sludge layer generated by precipitation of dissolved metals from aqueous wastes solutions during neutralization with sodium hydroxide, (2) a salt cake layer formed by crystallization of salts after evaporation of the supernate solution, and (3) an aqueous supernate solution that exists as a separate layer or as liquid contained in cavities between sludge or salt cake particles. The identity of chemical species of plutonium in these wastes will allow a better understanding of the behavior of the plutonium during storage in tanks, retrieval of the wastes, and processing of the wastes. Plutonium chemistry in the wastes is important to criticality and environmental concerns, and in processing the wastes for final disposal. Plutonium has been found to exist mainly in the sludge layers of the tanks along with other precipitated metal hydrous oxides. This is expected due to its low solubility in basic aqueous solutions. Tank supernate solutions do not contain high concentrations of plutonium even though some tanks contain high concentrations of complexing agents. The solutions also contain significant concentrations of hydroxide which competes with other

  11. Mixed Waste Encapsulation in Polyester Resins. Treatment for Mixed Wastes Containing Salts. Mixed Waste Focus Area. OST Reference #1685

    SciTech Connect

    None, None

    1999-09-01

    Throughout the Department of Energy (DOE) complex there are large inventories of homogeneous solid mixed wastes, such as treatment residues, fly ashes, and sludges that contain relatively high concentrations (greater than 15% by weight) of salts. The inherent solubility of nitrate, sulfate, and chloride salts makes traditional cement stabilization of these waste streams difficult, expensive, and challenging. Salts can effect the setting rate of cements and can react with cement hydration products to form expansive and cement damaging compounds. Many of these salt wastes are in a dry granular form and are the by-product of treating spent acidic and metal solutions used to recover and reformulate nuclear weapons materials over the past 50 years. At the Idaho National Engineering and Environmental Laboratory (INEEL) alone, there is approximately 8,000 cubic meters of nitrate salts (potassium and sodium nitrate) stored above ground with an earthen cover. Current estimates indicate that over 200 million kg of contaminated salt wastes exist at various DOE sites. Continued primary treatment of waste water coupled with the use of mixed waste incinerators may generate an additional 5 million kg of salt-containing, mixed waste residues each year. One of the obvious treatment solutions for these salt-containing wastes is to immobilize the hazardous components to meet Environmental Protection Agency/Resource Conservation and Recovery Act (EPA/RCRA) Land Disposal Restrictions (LDR), thus rendering the mixed waste to a radioactive waste only classification. One proposed solution is to use thermal treatment via vitrification to immobilize the hazardous component and thereby substantially reduce the volume, as well as provide exceptional durability. However, these melter systems involve expensive capital apparatus with complicated off-gas systems. In addition, the vitrification of high salt waste may cause foaming and usually requires extensive development to specify glass

  12. The safe disposal of radioactive wastes

    PubMed Central

    Kenny, A. W.

    1956-01-01

    A comprehensive review is given of the principles and problems involved in the safe disposal of radioactive wastes. The first part is devoted to a study of the basic facts of radioactivity and of nuclear fission, the characteristics of radioisotopes, the effects of ionizing radiations, and the maximum permissible levels of radioactivity for workers and for the general public. In the second part, the author describes the different types of radioactive waste—reactor wastes and wastes arising from the use of radioisotopes in hospitals and in industry—and discusses the application of the maximum permissible levels of radioactivity to their disposal and treatment, illustrating his discussion with an account of the methods practised at the principal atomic energy establishments. PMID:13374534

  13. Immobilization of IFR salt wastes in mortar

    SciTech Connect

    Fischer, D.F.; Johnson, T.R.

    1988-01-01

    Portland cement-base mortars are being considered for immobilizing chloride salt wastes produced by the fuel cycles of Integral Fast Reactors (IFR). The IFR is a sodium-cooled fast reactor with metal alloy fuels. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500/degree/C. This cell has a liquid cadmium anode in which the fuels are dissolved and a liquid salt electrolyte. The salt will be a mixture of either lithium, potassium, and sodium chlorides or lithium, calcium, barium, and sodium chlorides. One method being considered for immobilizing the treated nontransuranic salt waste is to disperse the salt in a portland cement-base mortar that will be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canister-molds where it will solidify into a strong, leach-resistant material. The set times must be longer than a few hours to allow sufficient time for processing, and the mortar must reach a reasonable compressive strength (/approximately/7 MPa) within three days to permit handling. Because fission product heating will be high, about 0.6 W/kg for a mortar containing 10% waste salt, the effects of elevated temperatures during curing and storage on mortar properties must be considered.

  14. Microwave remediation of hazardous and radioactive wastes

    SciTech Connect

    Wicks, G.G.

    2000-04-28

    A team from the Westinghouse Savannah River Technology Center (WSRC - a DOE Laboratory), and the University of Florida (UF - academia), has been active for about a decade in development of microwave technology for specialized waste management applications. This interaction has resulted in the development of unique equipment and uses of microwave energy for a variety of important applications for remediation of hazardous and radioactive wastes. Discussed are results of this unique technology for processing of electronic circuitry and components, medical wastes, discarded tires, and transuranic radioactive wastes.

  15. Method for storing radioactive combustible waste

    DOEpatents

    Godbee, H.W.; Lovelace, R.C.

    1973-10-01

    A method is described for preventing pressure buildup in sealed containers which contain radioactively contaminated combustible waste material by adding an oxide getter material to the container so as to chemically bind sorbed water and combustion product gases. (Official Gazette)

  16. Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

    SciTech Connect

    Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

    1997-07-07

    Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE`s mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies.

  17. Mobile plant for low-level radioactive waste reprocessing

    SciTech Connect

    Sobolev, I.A.; Panteleyev, V.I.; Demkin, V.I.

    1993-12-31

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

  18. Molten salt oxidation for treating low-level mixed wastes

    SciTech Connect

    Adamson, M G; Ford, T D; Foster, K G; Hipple, D L; Hopper, R W; Hsu, P C

    1998-12-10

    MS0 is a promising alternative to incineration for the treatment of a variety of organic wastes. Lawrence Livermore National Laboratory (LLNL) has prepared a facility (please see the photo attached) in which an integrated pilot-scale MS0 treatment system is being tested and demonstrated. The system consists of a MS0 vessel with a dedicated off-gas treatment system, a salt recycle system, feed preparation equipment, and a ceramic final waste forms immobilization system. The MSO/off-gas system has been operational since December 1997. The salt recycle system and the ceramic final forms immobilization became operational in May and August 1998, respectively. We have tested the MS0 facility with various organic feeds, including chlorinated solvents; tributyl phosphate/kerosene, PCB-contaminated waste oils & solvents, booties, plastic pellets, ion exchange resins, activated carbon, radioactive-spiked organics, and well-characterized low- level liquid mixed wastes. MS0 is a versatile technology for hazardous waste treatment and may be a solution to many waste disposal problems. In this paper we will present our operational experience with MS0 and also discuss its process capabilities as well as performance data with different feeds.

  19. Hazardous chemical and radioactive wastes at Hanford

    SciTech Connect

    Keller, J.F.; Stewart, T.L.

    1991-07-01

    The Hanford Site was established in 1944 to produce plutonium for defense. During the past four decades, a number of reactors, processing facilities, and waste management facilities have been built at Hanford for plutonium production. Generally, Hanford`s 100 Area was dedicated to reactor operation; the 200 Area to fuel reprocessing, plutonium recovery, and waste management; and the 300 Area to fuel fabrication and research and development. Wastes generated from these operations included highly radioactive liquid wastes, which were discharged to single- and double-shell tanks; solid wastes, including both transuranic (TRU) and low-level wastes, which were buried or discharged to caissons; and waste water containing low- to intermediate-level radioactivity, which was discharged to the soil column via near-surface liquid disposal units such as cribs, ponds, and retention basins. Virtually all of the wastes contained hazardous chemical as well as radioactive constituents. This paper will focus on the hazardous chemical components of the radioactive mixed waste generated by plutonium production at Hanford. The processes, chemicals used, methods of disposition, fate in the environment, and actions being taken to clean up this legacy are described by location.

  20. Hazardous chemical and radioactive wastes at Hanford

    SciTech Connect

    Keller, J.F.; Stewart, T.L.

    1991-07-01

    The Hanford Site was established in 1944 to produce plutonium for defense. During the past four decades, a number of reactors, processing facilities, and waste management facilities have been built at Hanford for plutonium production. Generally, Hanford's 100 Area was dedicated to reactor operation; the 200 Area to fuel reprocessing, plutonium recovery, and waste management; and the 300 Area to fuel fabrication and research and development. Wastes generated from these operations included highly radioactive liquid wastes, which were discharged to single- and double-shell tanks; solid wastes, including both transuranic (TRU) and low-level wastes, which were buried or discharged to caissons; and waste water containing low- to intermediate-level radioactivity, which was discharged to the soil column via near-surface liquid disposal units such as cribs, ponds, and retention basins. Virtually all of the wastes contained hazardous chemical as well as radioactive constituents. This paper will focus on the hazardous chemical components of the radioactive mixed waste generated by plutonium production at Hanford. The processes, chemicals used, methods of disposition, fate in the environment, and actions being taken to clean up this legacy are described by location.

  1. Method for solidification of radioactive and other hazardous waste

    DOEpatents

    Anshits, Alexander G.; Vereshchagina, Tatiana A.; Voskresenskaya, Elena N.; Kostin, Eduard M.; Pavlov, Vyacheslav F.; Revenko, Yurii A.; Tretyakov, Alexander A.; Sharonova, Olga M.; Aloy, Albert S.; Sapozhnikova, Natalia V.; Knecht, Dieter A.; Tranter, Troy J.; Macheret, Yevgeny

    2002-01-01

    Solidification of liquid radioactive waste, and other hazardous wastes, is accomplished by the method of the invention by incorporating the waste into a porous glass crystalline molded block. The porous block is first loaded with the liquid waste and then dehydrated and exposed to thermal treatment at 50-1,000.degree. C. The porous glass crystalline molded block consists of glass crystalline hollow microspheres separated from fly ash (cenospheres), resulting from incineration of fossil plant coals. In a preferred embodiment, the porous glass crystalline blocks are formed from perforated cenospheres of grain size -400+50, wherein the selected cenospheres are consolidated into the porous molded block with a binder, such as liquid silicate glass. The porous blocks are then subjected to repeated cycles of saturating with liquid waste, and drying, and after the last cycle the blocks are subjected to calcination to transform the dried salts to more stable oxides. Radioactive liquid waste can be further stabilized in the porous blocks by coating the internal surface of the block with metal oxides prior to adding the liquid waste, and by coating the outside of the block with a low-melting glass or a ceramic after the waste is loaded into the block.

  2. Radioactive Waste Management in A Hospital

    PubMed Central

    Khan, Shoukat; Syed, AT; Ahmad, Reyaz; Rather, Tanveer A.; Ajaz, M; Jan, FA

    2010-01-01

    Most of the tertiary care hospitals use radioisotopes for diagnostic and therapeutic applications. Safe disposal of the radioactive waste is a vital component of the overall management of the hospital waste. An important objective in radioactive waste management is to ensure that the radiation exposure to an individual (Public, Radiation worker, Patient) and the environment does not exceed the prescribed safe limits. Disposal of Radioactive waste in public domain is undertaken in accordance with the Atomic Energy (Safe disposal of radioactive waste) rules of 1987 promulgated by the Indian Central Government Atomic Energy Act 1962. Any prospective plan of a hospital that intends using radioisotopes for diagnostic and therapeutic procedures needs to have sufficient infrastructural and manpower resources to keep its ambient radiation levels within specified safe limits. Regular monitoring of hospital area and radiation workers is mandatory to assess the quality of radiation safety. Records should be maintained to identify the quality and quantity of radioactive waste generated and the mode of its disposal. Radiation Safety officer plays a key role in the waste disposal operations. PMID:21475524

  3. ASSESSMENT OF RADIOACTIVE AND NON-RADIOACTIVE CONTAMINANTS FOUND IN LOW LEVEL RADIOACTIVE WASTE STREAMS

    SciTech Connect

    R.H. Little, P.R. Maul, J.S.S. Penfoldag

    2003-02-27

    This paper describes and presents the findings from two studies undertaken for the European Commission to assess the long-term impact upon the environment and human health of non-radioactive contaminants found in various low level radioactive waste streams. The initial study investigated the application of safety assessment approaches developed for radioactive contaminants to the assessment of nonradioactive contaminants in low level radioactive waste. It demonstrated how disposal limits could be derived for a range of non-radioactive contaminants and generic disposal facilities. The follow-up study used the same approach but undertook more detailed, disposal system specific calculations, assessing the impacts of both the non-radioactive and radioactive contaminants. The calculations undertaken indicated that it is prudent to consider non-radioactive, as well as radioactive contaminants, when assessing the impacts of low level radioactive waste disposal. For some waste streams with relatively low concentrations of radionuclides, the potential post-closure disposal impacts from non-radioactive contaminants can be comparable with the potential radiological impacts. For such waste streams there is therefore an added incentive to explore options for recycling the materials involved wherever possible.

  4. Disposal of oil field wastes and NORM wastes into salt caverns.

    SciTech Connect

    Veil, J. A.

    1999-01-27

    Salt caverns can be formed through solution mining in the bedded or domal salt formations that are found in many states. Salt caverns have traditionally been used for hydrocarbon storage, but caverns have also been used to dispose of some types of wastes. This paper provides an overview of several years of research by Argonne National Laboratory on the feasibility and legality of using salt caverns for disposing of nonhazardous oil field wastes (NOW) and naturally occurring radioactive materials (NORM), the risk to human populations from this disposal method, and the cost of cavern disposal. Costs are compared between the four operating US disposal caverns and other commercial disposal options located in the same geographic area as the caverns. Argonne's research indicates that disposal of NOW into salt caverns is feasible and, in most cases, would not be prohibited by state agencies (although those agencies may need to revise their wastes management regulations). A risk analysis of several cavern leakage scenarios suggests that the risk from cavern disposal of NOW and NORM wastes is below accepted safe risk thresholds. Disposal caverns are economically competitive with other disposal options.

  5. Method for solidifying liquid radioactive wastes

    DOEpatents

    Berreth, Julius R.

    1976-01-01

    The quantity of nitrous oxides produced during the solidification of liquid radioactive wastes containing nitrates and nitrites can be substantially reduced by the addition to the wastes of a stoichiometric amount of urea which, upon heating, destroys the nitrates and nitrites, liberating nontoxic N.sub.2, CO.sub.2 and NH.sub.3.

  6. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM 2008

    SciTech Connect

    West, B.; Waltz, R.

    2009-06-11

    Aqueous radioactive wastes from Savannah River Site (SRS) separations and vitrification processes are contained in large underground carbon steel tanks. Inspections made during 2008 to evaluate these vessels and other waste handling facilities along with evaluations based on data from previous inspections are the subject of this report.

  7. Nondestructive assay of boxed radioactive waste

    SciTech Connect

    Gilles, W.P.; Roberts, R.J.; Jasen, W.G.

    1992-12-01

    This paper describes the problems related to the nondestructive assay (NDA) of boxed radioactive waste at the Hanford Site and how Westinghouse Hanford company (WHC) is solving the problems. The waste form and radionuclide content are described. The characteristics of the combined neutron and gamma-based measurement system are described.

  8. Apparatus and method for radioactive waste screening

    DOEpatents

    Akers, Douglas W.; Roybal, Lyle G.; Salomon, Hopi; Williams, Charles Leroy

    2012-09-04

    An apparatus and method relating to screening radioactive waste are disclosed for ensuring that at least one calculated parameter for the measurement data of a sample falls within a range between an upper limit and a lower limit prior to the sample being packaged for disposal. The apparatus includes a radiation detector configured for detecting radioactivity and radionuclide content of the of the sample of radioactive waste and generating measurement data in response thereto, and a collimator including at least one aperture to direct a field of view of the radiation detector. The method includes measuring a radioactive content of a sample, and calculating one or more parameters from the radioactive content of the sample.

  9. Reduction of INTEC Analytical Radioactive Liquid Waste

    SciTech Connect

    Johnson, Virgil James; Hu, Jian Sheng; Chambers, Andrea

    1999-06-01

    This report details the evaluation of the reduction in radioactive liquid waste from the analytical laboratories sent to the Process Effluent Waste system (deep tanks). The contributors are the Analytical Laboratories Department (ALD), the Waste Operations Department, the laboratories at CPP-637, and natural run off. Other labs were contacted to learn of methods used and if any new technologies had emerged. A waste generation database was made from the current methods in use in the ALD. From this database, methods were targeted to reduce waste. Individuals were contacted on ways to reduce waste. The results are: a new method generating much less waste, several methods being handled differently, some cleaning processes being changed to reduce waste, and changes to reduce chemicals to waste.

  10. Reduction of INTEC Analytical Radioactive Liquid Wastes

    SciTech Connect

    V. J. Johnson; J. S. Hu; A. G. Chambers

    1999-06-01

    This report details the evaluation of the reduction in radioactive liquid waste from the analytical laboratories sent to the Process Effluent Waste system (deep tanks). The contributors are the Analytical Laboratories Department (ALD), the Waste Operations Department, the laboratories at CPP-637, and natural run off. Other labs were contacted to learn the methods used and if any new technologies had emerged. A waste generation database was made from the current methods in used in the ALD. From this database, methods were targeted to reduce waste. Individuals were contacted on ways to reduce waste. The results are: a new method generating much less waste, several methods being handled differently, some cleaning processes being changed to reduce waste, and changes to reduce chemicals to waste.

  11. Molten salt destruction of energetic waste materials

    DOEpatents

    Brummond, W.A.; Upadhye, R.S.; Pruneda, C.O.

    1995-07-18

    A molten salt destruction process is used to treat and destroy energetic waste materials such as high explosives, propellants, and rocket fuels. The energetic material is pre-blended with a solid or fluid diluent in safe proportions to form a fluid fuel mixture. The fuel mixture is rapidly introduced into a high temperature molten salt bath. A stream of molten salt is removed from the vessel and may be recycled as diluent. Additionally, the molten salt stream may be pumped from the reactor, circulated outside the reactor for further processing, and delivered back into the reactor or cooled and circulated to the feed delivery system to further dilute the fuel mixture entering the reactor. 4 figs.

  12. Molten salt destruction of energetic waste materials

    DOEpatents

    Brummond, William A.; Upadhye, Ravindra S.; Pruneda, Cesar O.

    1995-01-01

    A molten salt destruction process is used to treat and destroy energetic waste materials such as high explosives, propellants, and rocket fuels. The energetic material is pre-blended with a solid or fluid diluent in safe proportions to form a fluid fuel mixture. The fuel mixture is rapidly introduced into a high temperature molten salt bath. A stream of molten salt is removed from the vessel and may be recycled as diluent. Additionally, the molten salt stream may be pumped from the reactor, circulated outside the reactor for further processing, and delivered back into the reactor or cooled and circulated to the feed delivery system to further dilute the fuel mixture entering the reactor.

  13. Radioactive tank waste remediation focus area

    SciTech Connect

    1996-08-01

    EM`s Office of Science and Technology has established the Tank Focus Area (TFA) to manage and carry out an integrated national program of technology development for tank waste remediation. The TFA is responsible for the development, testing, evaluation, and deployment of remediation technologies within a system architecture to characterize, retrieve, treat, concentrate, and dispose of radioactive waste stored in the underground stabilize and close the tanks. The goal is to provide safe and cost-effective solutions that are acceptable to both the public and regulators. Within the DOE complex, 335 underground storage tanks have been used to process and store radioactive and chemical mixed waste generated from weapon materials production and manufacturing. Collectively, thes tanks hold over 90 million gallons of high-level and low-level radioactive liquid waste in sludge, saltcake, and as supernate and vapor. Very little has been treated and/or disposed or in final form.

  14. Hazardous and radioactive waste incineration studies

    NASA Astrophysics Data System (ADS)

    Vavruska, J. S.; Stretz, L. A.; Borduin, L. C.

    Development and demonstration of a transuranic (TRU) waste volume-reduction process is described. A production-scale controlled air incinerator using commercially available equipment and technology was modified for solid radioactive waste service. This unit successfully demonstrated the volume reduction of transuranic (TRU) waste with an average TRU content of about 20 nCi/g. The same incinerator and offgas treatment system is being modified further to evaluate the destruction of hazardous liquid wastes such as polychlorinated biphenyls (PCBs) and hazardous solid wastes such as pentachlorophenol (PCP)-treated wood.

  15. 40 CFR 227.30 - High-level radioactive waste.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 40 Protection of Environment 25 2011-07-01 2011-07-01 false High-level radioactive waste. 227.30 Section 227.30 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from...

  16. 40 CFR 227.30 - High-level radioactive waste.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 25 2014-07-01 2014-07-01 false High-level radioactive waste. 227.30 Section 227.30 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from...

  17. 40 CFR 227.30 - High-level radioactive waste.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 26 2012-07-01 2011-07-01 true High-level radioactive waste. 227.30 Section 227.30 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from...

  18. 40 CFR 227.30 - High-level radioactive waste.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 40 Protection of Environment 26 2013-07-01 2013-07-01 false High-level radioactive waste. 227.30 Section 227.30 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from...

  19. Public involvement in radioactive waste management decisions

    SciTech Connect

    1994-04-01

    Current repository siting efforts focus on Yucca Mountain, Nevada, where DOE`s Office of Civilian Radioactive Waste Management (OCRWM) is conducting exploratory studies to determine if the site is suitable. The state of Nevada has resisted these efforts: it has denied permits, brought suit against DOE, and publicly denounced the federal government`s decision to study Yucca Mountain. The state`s opposition reflects public opinion in Nevada, and has considerably slowed DOE`s progress in studying the site. The Yucca Mountain controversy demonstrates the importance of understanding public attitudes and their potential influence as DOE develops a program to manage radioactive waste. The strength and nature of Nevada`s opposition -- its ability to thwart if not outright derail DOE`s activities -- indicate a need to develop alternative methods for making decisions that affect the public. This report analyzes public participation as a key component of this openness, one that provides a means of garnering acceptance of, or reducing public opposition to, DOE`s radioactive waste management activities, including facility siting and transportation. The first section, Public Perceptions: Attitudes, Trust, and Theory, reviews the risk-perception literature to identify how the public perceives the risks associated with radioactivity. DOE and the Public discusses DOE`s low level of credibility among the general public as the product, in part, of the department`s past actions. This section looks at the three components of the radioactive waste management program -- disposal, storage, and transportation -- and the different ways DOE has approached the problem of public confidence in each case. Midwestern Radioactive Waste Management Histories focuses on selected Midwestern facility-siting and transportation activities involving radioactive materials.

  20. Brine migration in salt and its implications in the geologic disposal of nuclear waste

    SciTech Connect

    Jenks, G.H.; Claiborne, H.C.

    1981-12-01

    This report respresents a comprehensive review and analysis of available information relating to brine migration in salt surrounding radioactive waste in a salt repository. The topics covered relate to (1) the characteristics of salt formations and waste packages pertinent to considerations of rates, amounts, and effects of brine migration, (2) experimental and theoretical information on brine migration, and (3) means of designing to minimize any adverse effects of brine migration. Flooding, brine pockets, and other topics were not considered, since these features will presumably be eliminated by appropriate site selection and repository design. 115 references.

  1. Radioactive waste management in the former USSR

    SciTech Connect

    Bradley, D.J.

    1992-06-01

    Radioactive waste materials--and the methods being used to treat, process, store, transport, and dispose of them--have come under increased scrutiny over last decade, both nationally and internationally. Nuclear waste practices in the former Soviet Union, arguably the world's largest nuclear waste management system, are of obvious interest and may affect practices in other countries. In addition, poor waste management practices are causing increasing technical, political, and economic problems for the Soviet Union, and this will undoubtedly influence future strategies. this report was prepared as part of a continuing effort to gain a better understanding of the radioactive waste management program in the former Soviet Union. the scope of this study covers all publicly known radioactive waste management activities in the former Soviet Union as of April 1992, and is based on a review of a wide variety of literature sources, including documents, meeting presentations, and data base searches of worldwide press releases. The study focuses primarily on nuclear waste management activities in the former Soviet Union, but relevant background information on nuclear reactors is also provided in appendixes.

  2. Pump station for radioactive waste water

    DOEpatents

    Whitton, John P.; Klos, Dean M.; Carrara, Danny T.; Minno, John J.

    2003-11-18

    A pump station for transferring radioactive particle containing waste water, includes: (a.) an enclosed sump having a vertically elongated right frusto conical wall surface and a bottom surface and (b.) a submersible volute centrifugal pump having a horizontally rotating impeller and a volute exterior surface. The sump interior surface, the bottom surface and the volute exterior surface are made of stainless steel having a 30 Ra or finer surface finish. A 15 Ra finish has been found to be most cost effective. The pump station is used for transferring waste water, without accumulation of radioactive fines.

  3. Collection and Segregation of Radioactive Waste. Principals for Characterization and Classification of Radioactive Waste

    SciTech Connect

    Dziewinska, K.M.

    1998-09-28

    Radioactive wastes are generated by all activities which utilize radioactive materials as part of their processes. Generally such activities include all steps in the nuclear fuel cycle (for power generation) and non-fuel cycle activities. The increasing production of radioisotopes in a Member State without nuclear power must be accompanied by a corresponding development of a waste management system. An overall waste management scheme consists of the following steps: segregation, minimization, treatment, conditioning, storage, transport, and disposal. To achieve a satisfactory overall management strategy, all steps have to be complementary and compatible. Waste segregation and minimization are of great importance mainly because they lead to cost reduction and reduction of dose commitments to the personnel that handle the waste. Waste characterization plays a significant part in the waste segregation and waste classification processes, it implicates required waste treatment process including the need for the safety assessment of treatment conditioning and storage facilities.

  4. Cementitious Stabilization of Mixed Wastes with High Salt Loadings

    SciTech Connect

    Spence, R.D.; Burgess, M.W.; Fedorov, V.V.; Downing, D.J.

    1999-04-01

    Salt loadings approaching 50 wt % were tolerated in cementitious waste forms that still met leach and strength criteria, addressing a Technology Deficiency of low salt loadings previously identified by the Mixed Waste Focus Area. A statistical design quantified the effect of different stabilizing ingredients and salt loading on performance at lower loadings, allowing selection of the more effective ingredients for studying the higher salt loadings. In general, the final waste form needed to consist of 25 wt % of the dry stabilizing ingredients to meet the criteria used and 25 wt % water to form a workable paste, leaving 50 wt % for waste solids. The salt loading depends on the salt content of the waste solids but could be as high as 50 wt % if all the waste solids are salt.

  5. Annual Radioactive Waste Tank Inspection Program 1994

    SciTech Connect

    McNatt, F.G. Sr.

    1995-04-01

    Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1994 to evaluate these vessels and evaluations based on data accrued by inspections made since the tanks were constructed are the subject of this report.

  6. High-level radioactive wastes. Supplement 1

    SciTech Connect

    McLaren, L.H.

    1984-09-01

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

  7. Annual Radioactive Waste Tank Inspection Program - 1997

    SciTech Connect

    McNatt, F.G.

    1998-05-01

    Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1997 to evaluate these vessels, and evaluations based on data accrued by inspections performed since the tanks were constructed are the subject of this report.

  8. Annual radioactive waste tank inspection program - 1999

    SciTech Connect

    Moore, C.J.

    2000-04-14

    Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1999 to evaluate these vessels and auxiliary appurtenances along with evaluations based on data accrued by inspections performed since the tanks were constructed are the subject of this report.

  9. Annual Radioactive Waste Tank Inspection Program - 1998

    SciTech Connect

    McNatt, F.G.

    1999-10-27

    Aqueous radioactive wastes from Savannah River Site separations processes are contained in large underground carbon steel tanks. Inspections made during 1998 to evaluate these vessels and auxiliary appurtenances, along with evaluations based on data accrued by inspections performed since the tanks were constructed, are the subject of this report.

  10. Annual radioactive waste tank inspection program - 1996

    SciTech Connect

    McNatt, F.G.

    1997-04-01

    Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1996 to evaluate these vessels, and evaluations based on data accrued by inspections performed since the tanks were constructed, are the subject of this report.

  11. Annual radioactive waste tank inspection program: 1995

    SciTech Connect

    McNatt, F.G. Sr.

    1996-04-01

    Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1995 to evaluate these vessels and evaluations based on data accrued by inspections performed since the tanks were constructed are the subject of this report

  12. Annual radioactive waste tank inspection program - 1992

    SciTech Connect

    McNatt, F.G.

    1992-12-31

    Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1992 to evaluate these vessels and evaluations based on data accrued by inspections made since the tanks were constructed are the subject of this report.

  13. Method of treating radioactively contaminated solvent waste

    SciTech Connect

    Jablonski, W.; Mallek, H.; Plum, W.

    1981-07-07

    A method of and apparatus for treating radioactively contaminated solvent waste are claimed. The solvent waste is supplied to material such as peat, vermiculite, diaton, etc. This material effects the distribution or dispersion of the solvent and absorbs the foreign substances found in the solvent waste. Air or an inert gas flows through the material in order to pick up the solvent portions which are volatile as a consequence of their vapor pressure. The thus formed gas mixture, which includes air or inert gas and solvent portions, is purified in a known manner by thermal, electrical, or catalytic combustion of the solvent portions.

  14. Radioactive Waste Management in Central Asia - 12034

    SciTech Connect

    Zhunussova, Tamara; Sneve, Malgorzata; Liland, Astrid

    2012-07-01

    After the collapse of the Soviet Union the newly independent states in Central Asia (CA) whose regulatory bodies were set up recently are facing problems with the proper management of radioactive waste and so called 'nuclear legacy' inherited from the past activities. During the former Soviet Union (SU) period, various aspects of nuclear energy use took place in CA republics of Kazakhstan, Kyrgyzstan, Tajikistan and Uzbekistan. Activities range from peaceful use of energy to nuclear testing for example at the former Semipalatinsk Nuclear Test Site (SNTS) in Kazakhstan, and uranium mining and milling industries in all four countries. Large amounts of radioactive waste (RW) have been accumulated in Central Asia and are waiting for its safe disposal. In 2008 the Norwegian Radiation Protection Authority (NRPA), with the support of the Norwegian Ministry of Foreign Affairs, has developed bilateral projects that aim to assist the regulatory bodies in Kazakhstan, Kyrgyzstan Tajikistan, and Uzbekistan (from 2010) to identify and draft relevant regulatory requirements to ensure the protection of the personnel, population and environment during the planning and execution of remedial actions for past practices and radioactive waste management in the CA countries. The participating regulatory authorities included: Kazakhstan Atomic Energy Agency, Kyrgyzstan State Agency on Environmental Protection and Forestry, Nuclear Safety Agency of Tajikistan, and State Inspectorate on Safety in Industry and Mining of Uzbekistan. The scope of the projects is to ensure that activities related to radioactive waste management in both planned and existing exposure situations in CA will be carried out in accordance with the international guidance and recommendations, taking into account the relevant regulatory practice from other countries in this area. In order to understand the problems in the field of radioactive waste management we have analysed the existing regulations through the so

  15. Radioactive Waste Burial Grounds. Environmental Information Document

    SciTech Connect

    Jaegge, W.J.; Kolb, N.L.; Looney, B.B.; Marine, I.W.; Towler, O.A.; Cook, J.R.

    1987-03-01

    This document provides environmental information on postulated closure options for the Radioactive Waste Burial Grounds at the Savannah River Plant and was developed as background technical documentation for the Department of Energy`s proposed Environmental Impact Statement (EIS) on waste management activities for groundwater protection at the plant. The results of groundwater and atmospheric pathway analyses, accident analysis, and other environmental assessments discussed in this document are based upon a conservative analysis of all foreseeable scenarios as defined by the National Environmental Policy Act (CFR, 1986). The scenarios do not necessarily represent actual environmental conditions. This document is not meant to be used as a closure plan or other regulatory document to comply with required federal or state environmental regulations. The closure options considered for the Radioactive Waste Burial Grounds are waste removal and closure, no waste removal and closure, and no action. The predominant pathways for human exposure to chemical and/or radioactive constituents are through surface, subsurface, and atmospheric transport. Modeling calculations were made to determine the risks to human population via these general pathways for the three postulated closure options. An ecological assessment was conducted to predict the environmental impacts on aquatic and terrestrial biota. The relative costs for each of the closure options were estimated.

  16. Radioactive waste isolation in salt: peer review of the Texas Bureau of Economic Geology's report on the Petrographic, Stratigraphic, and Structural Evidence for Dissolution of Upper Permian Bedded Salt, Texas Panhandle

    SciTech Connect

    Fenster, D.F.; Anderson, R.Y.; Gonzales, S.; Baker, V.R.; Edgar, D.E.; Harrison, W.

    1984-08-01

    The following recommendations for improving the Texas Bureau of Economic Geology (TBEG) report entitled Petrographic, Stratigraphic, and Structural Evidence for Dissolution of Upper Permian Bedded Salt, Texas Panhandle have been abstracted from the body of this review report. The TBEG report should be resided to conform to one of the following alternatives: (1) If the report is intended to be a review or summary of previous work, it should contain more raw data, be edited to give equal treatment to all types of data, and include summary tables and additional figures. (2) If the report is intended to be a description and interpretation of petrographic evidence for salt dissolution, supported by collateral stratigraphic and structural evidence, the relevant indirect and direct data should become the focal point of the report. The following recommendations apply to one or both of the options listed above. (1) The text should differentiate more carefully between the data and inferences based on those data. (2) The authors should retain the qualifiers present in cited works. Statements in the report that are based on earlier papers are sometimes stronger than those in the papers themselves. (3) The next revision should present more complete data. (4) The authors should achieve a more balanced presentation of alternative hypotheses and interpretations. They could then discuss the relative merits of the alternative interpretations. (5) More attention should be given to clear exposition.

  17. Soluble pig for radioactive waste transfer lines

    SciTech Connect

    Ohl, P.C., Westinghouse Hanford

    1996-12-02

    Flushing transfer pipe after radioactive waste transfers generates thousands of gallons of additional radioactive waste each year at the Hanford site. The use of pneumatic pigging with waste soluble pigs as a means to clear transfer piping may be an effective alternative to raw water flushes. A feasibility study was performed by a group of senior mechanical engineering students for their senior design project as part of their curriculum at Washington State University. The students divided the feasibility study into three sub-projects involving: (1) materials research, (2) delivery system design, and (3) mockup fabrication and testing. The students screened through twenty-three candidate materials and selected a thermoplastic polymer combined 50:50 wt% with sucrose to meet the established material performance criteria. The students also prepared a conceptual design of a remote pneumatic delivery system and constructed a mockup section of transfer pipe for testing the prototype pigs.

  18. Handbook of high-level radioactive waste transportation

    SciTech Connect

    Sattler, L.R.

    1992-10-01

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

  19. Greater confinement disposal of radioactive wastes

    SciTech Connect

    Trevorrow, L.E.; Gilbert, T.L.; Luner, C.; Merry-Libby, P.A.; Meshkov, N.K.; Yu, C.

    1985-01-01

    Low-level radioactive waste (LLW) includes a broad spectrum of different radionuclide concentrations, half-lives, and hazards. Standard shallow-land burial practice can provide adequate protection of public health and safety for most LLW. A small volume fraction (approx. 1%) containing most of the activity inventory (approx. 90%) requires specific measures known as greater-confinement disposal (GCD). Different site characteristics and different waste characteristics - such as high radionuclide concentrations, long radionuclide half-lives, high radionuclide mobility, and physical or chemical characteristics that present exceptional hazards - lead to different GCD facility design requirements. Facility design alternatives considered for GCD include the augered shaft, deep trench, engineered structure, hydrofracture, improved waste form, and high-integrity container. Selection of an appropriate design must also consider the interplay between basic risk limits for protection of public health and safety, performance characteristics and objectives, costs, waste-acceptance criteria, waste characteristics, and site characteristics.

  20. Polyethylene encapsulatin of nitrate salt wastes: Waste form stability, process scale-up, and economics

    SciTech Connect

    Kalb, P.D.; Heiser, J.H. III; Colombo, P.

    1991-07-01

    A polyethylene encapsulation system for treatment of low-level radioactive, hazardous, and mixed wastes has been developed at Brookhaven National Laboratory. Polyethylene has several advantages compared with conventional solidification/stabilization materials such as hydraulic cements. Waste can be encapsulated with greater efficiency and with better waste form performance than is possible with hydraulic cement. The properties of polyethylene relevant to its long-term durability in storage and disposal environments are reviewed. Response to specific potential failure mechanisms including biodegradation, radiation, chemical attack, flammability, environmental stress cracking, and photodegradation are examined. These data are supported by results from extensive waste form performance testing including compressive yield strength, water immersion, thermal cycling, leachability of radioactive and hazardous species, irradiation, biodegradation, and flammability. The bench-scale process has been successfully tested for application with a number of specific problem'' waste streams. Quality assurance and performance testing of the resulting waste form confirmed scale-up feasibility. Use of this system at Rocky Flats Plant can result in over 70% fewer drums processed and shipped for disposal, compared with optimal cement formulations. Based on the current Rocky Flats production of nitrate salt per year, polyethylene encapsulation can yield an estimated annual savings between $1.5 million and $2.7 million, compared with conventional hydraulic cement systems. 72 refs., 23 figs., 16 tabs.

  1. Control of radioactive waste-glass melters

    SciTech Connect

    Bickford, D.F. ); Hrma, P. ); Bowan, B.W. II )

    1990-01-01

    Slurries of simulated high level radioactive waste and glass formers have been isothermally reacted and analyzed to identify the sequence of the major chemical reactions in waste vitrification, their effect on glass production rate, and the development of leach resistance. Melting rates of waste batches have been increased by the addition of reducing agents (formic acid, sucrose) and nitrates. The rate increases are attributable in part to exothermic reactions which occur at critical stages in the vitrification process. Nitrates must be balanced by adequate reducing agents to avoid the formation of persistent foam, which would destabilize the melting process. The effect of foaming on waste glass production rates is analyzed, and melt rate limitations defined for waste-glass melters, based upon measurable thermophysical properties. Minimum melter residence times required to homogenize glass and assure glass quality are much smaller than those used in current practice. Thus, melter size can be reduced without adversely affecting glass quality. Physical chemistry and localized heat transfer of the waste-glass melting process are examined, to refine the available models for predicting and assuring glass production rate. It is concluded that the size of replacement melters and future waste processing facilities can be significantly decreased if minimum heat transfer requirements for effective melting are met by mechanical agitation. A new class of waste glass melters has been designed, and proof of concept tests completed on simulated High Level Radioactive Waste slurry. Melt rates have exceeded 155 kg m{sup {minus}2} h{sup {minus}1} with slurry feeds (32 lb ft{sup {minus}2} h{sup {minus}1}), and 229 kg kg m{sup {minus}2} h{sup {minus}1} with dry feed (47 lb ft{sup {minus}2} h{sup {minus}1}). This is about 8 times the melt rate possible in conventional waste- glass melters of the same size. 39 refs., 5 figs., 9 tabs.

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

    SciTech Connect

    Fischer, D.K.; Gitt, M.; Williams, G.A.; Branch, S.; Otis, M.D.; McKenzie-Carter, M.A.; Schurman, D.L.

    1991-07-01

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

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

    SciTech Connect

    Fischer, D.K.; Gitt, M.; Williams, G.A.; Branch, S. ); Otis, M.D.; McKenzie-Carter, M.A.; Schurman, D.L. )

    1991-07-01

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

  4. Disposal of nonhazardous oil field wastes into salt caverns

    SciTech Connect

    Veil, J.; Elcock, D.; Raivel, M.; Caudle, D.

    1996-12-31

    Bedded and domal salt deposits occur in many states. If salt deposits are thick enough, salt caverns can be formed through solution mining. These caverns are created either incidentally as a result of salt recovery or intentionally to create an underground chamber that can be used for storing hydrocarbon products or disposing of wastes. This paper evaluates the legality, feasibility, and suitability of disposing of nonhazardous oil and gas exploration, development, and production wastes (hereafter referred to as oil field wastes, unless otherwise noted) in salt caverns.

  5. Delivery system for molten salt oxidation of solid waste

    DOEpatents

    Brummond, William A.; Squire, Dwight V.; Robinson, Jeffrey A.; House, Palmer A.

    2002-01-01

    The present invention is a delivery system for safety injecting solid waste particles, including mixed wastes, into a molten salt bath for destruction by the process of molten salt oxidation. The delivery system includes a feeder system and an injector that allow the solid waste stream to be accurately metered, evenly dispersed in the oxidant gas, and maintained at a temperature below incineration temperature while entering the molten salt reactor.

  6. Geological problems in radioactive waste isolation

    SciTech Connect

    Witherspoon, P.A.

    1991-01-01

    The problem of isolating radioactive wastes from the biosphere presents specialists in the fields of earth sciences with some of the most complicated problems they have ever encountered. This is especially true for high level waste (HLW) which must be isolated in the underground and away from the biosphere for thousands of years. Essentially every country that is generating electricity in nuclear power plants is faced with the problem of isolating the radioactive wastes that are produced. The general consensus is that this can be accomplished by selecting an appropriate geologic setting and carefully designing the rock repository. Much new technology is being developed to solve the problems that have been raised and there is a continuing need to publish the results of new developments for the benefit of all concerned. The 28th International Geologic Congress that was held July 9--19, 1989 in Washington, DC provided an opportunity for earth scientists to gather for detailed discussions on these problems. Workshop W3B on the subject, Geological Problems in Radioactive Waste Isolation -- A World Wide Review'' was organized by Paul A Witherspoon and Ghislain de Marsily and convened July 15--16, 1989 Reports from 19 countries have been gathered for this publication. Individual papers have been cataloged separately.

  7. Combustible radioactive waste treatment by incineration and chemical digestion

    SciTech Connect

    Stretz, L.A.; Crippen, M.D.; Allen, C.R.

    1980-05-28

    A review is given of present and planned combustible radioactive waste treatment systems in the US. Advantages and disadvantages of various systems are considered. Design waste streams are discussed in relation to waste composition, radioactive contaminants by amount and type, and special operating problems caused by the waste.

  8. CHARACTERIZATION OF HIGH PHOSPHATE RADIOACTIVE TANK WASTE AND SIMULANT DEVELOPMENT

    SciTech Connect

    Lumetta, Gregg J.; McNamara, Bruce K.; Buck, Edgar C.; Fiskum, Sandra K.; Snow, Lanee A.

    2009-10-15

    A sample of high-level radioactive tank waste was characterized to provide a basis for developing a waste simulant. The simulant is required for engineered-scaled testing of pretreatment processes in a non-radiological facility. The waste material examined was derived from the bismuth phosphate process, which was the first industrial process implemented to separate plutonium from irradiated nuclear fuel. The bismuth phosphate sludge is a complex mixture rich in bismuth, iron, sodium, phosphorus, silicon, and uranium. The form of phosphorus in this particular tank waste material is of specific importance because that is the primary component (other than water-soluble sodium salts) that must be removed from the high-level waste solids by pretreatment. This work shows unequivocally that the phosphorus present in this waste material is not present as bismuth phosphate. Rather, the phosphorus appears to be incorporated mostly into an amorphous iron(III) phosphate species. The bismuth in the sludge solids is best described as bismuth ferrite, BiFeO3. Infrared spectral data, microscopy, and thermal analysis data are presented to support these conclusions. The behavior of phosphorus during caustic leaching of the bismuth phosphate sludge solids is also discussed.

  9. Risk methodology for geologic disposal of radioactive waste

    SciTech Connect

    Cranwell, R.M.; Campbell, J.E.; Ortiz, N.R. ); Guzowski, R.V. )

    1990-04-01

    This report contains the description of a procedure for selecting scenarios that are potentially important to the isolation of high- level radioactive wastes in deep geologic formations. In this report, the term scenario is used to represent a set of naturally occurring and/or human-induced conditions that represent realistic future states of the repository, geologic systems, and ground-water flow systems that might affect the release and transport of radionuclides from the repository to humans. The scenario selection procedure discussed in this report is demonstrated by applying it to the analysis of a hypothetical waste disposal site containing a bedded-salt formation as the host medium for the repository. A final set of 12 scenarios is selected for this site. 52 refs., 48 figs., 5 tabs.

  10. Caustic Recycle from Hanford Tank Waste Using NaSICON Ceramic Membrane Salt Splitting Process

    SciTech Connect

    Fountain, Matthew S.; Kurath, Dean E.; Sevigny, Gary J.; Poloski, Adam P.; Pendleton, J.; Balagopal, S.; Quist, M.; Clay, D.

    2009-02-20

    A family of inorganic ceramic materials, called sodium (Na) Super Ion Conductors (NaSICON), has been studied at Pacific Northwest National Laboratory (PNNL) to investigate their ability to separate sodium from radioactively contaminated sodium salt solutions for treating U.S. Department of Energy (DOE) tank wastes. Ceramatec Inc. developed and fabricated a membrane containing a proprietary NAS-GY material formulation that was electrochemically tested in a bench-scale apparatus with both a simulant and a radioactive tank-waste solution to determine the membrane performance when removing sodium from DOE tank wastes. Implementing this sodium separation process can result in significant cost savings by reducing the disposal volume of low-activity wastes and by producing a NaOH feedstock product for recycle into waste treatment processes such as sludge leaching, regenerating ion exchange resins, inhibiting corrosion in carbon-steel tanks, or retrieving tank wastes.

  11. Treatment of Difficult Wastes with Molten Salt Oxidation

    SciTech Connect

    Hsu, P C; Kwak, S

    2003-02-21

    Molten salt oxidation (MSO) is a good alternative to incineration for the treatment of a variety of organic wastes such as explosives, low-level mixed waste streams, PCB contaminated oils, spent resins and carbon. Since mid-1990s, the U.S. Army Defense Ammunition Center (DAC) and the Department of Energy (DOE) have jointly invested in MSO development at the Lawrence Livermore National Laboratory (LLNL). LLNL first demonstrated the MSO process for the effective destruction of explosives, explosives-contaminated materials, and other wastes on a 1.5-kg/hr bench-scale unit, and then in an integrated MSO facility capable of treating 8 kg/hr of low-level radioactive mixed wastes. Several MSO systems have been built with sizes up to 10 ft in height and 16 inches in diameter. LLNL in 2001 completed a MSO plant for DAC for the destruction of explosives-contaminated sludge and explosives-contaminated carbon. We will present in this paper our latest demonstration data and our operational experience with MSO.

  12. System for handling and storing radioactive waste

    DOEpatents

    Anderson, J.K.; Lindemann, P.E.

    1982-07-19

    A system and method are claimed for handling and storing spent reactor fuel and other solid radioactive waste, including canisters to contain the elements of solid waste, storage racks to hold a plurality of such canisters, storage bays to store these racks in isolation by means of shielded doors in the bays. This system also includes means for remotely positioning the racks in the bays and an access tunnel within which the remotely operated means is located to position a rack in a selected bay. The modular type of these bays will facilitate the construction of additional bays and access tunnel extension.

  13. System for handling and storing radioactive waste

    DOEpatents

    Anderson, John K.; Lindemann, Paul E.

    1984-01-01

    A system and method for handling and storing spent reactor fuel and other solid radioactive waste, including canisters to contain the elements of solid waste, storage racks to hold a plurality of such canisters, storage bays to store these racks in isolation by means of shielded doors in the bays. This system also includes means for remotely positioning the racks in the bays and an access tunnel within which the remotely operated means is located to position a rack in a selected bay. The modular type of these bays will facilitate the construction of additional bays and access tunnel extension.

  14. Waste Isolation Pilot Plant Salt Decontamination Testing

    SciTech Connect

    Rick Demmer; Stephen Reese

    2014-09-01

    On February 14, 2014, americium and plutonium contamination was released in the Waste Isolation Pilot Plant (WIPP) salt caverns. At the request of WIPP’s operations contractor, Idaho National Laboratory (INL) personnel developed several methods of decontaminating WIPP salt, using surrogate contaminants and also americium (241Am). The effectiveness of the methods is evaluated qualitatively, and to the extent possible, quantitatively. One of the requirements of this effort was delivering initial results and recommendations within a few weeks. That requirement, in combination with the limited scope of the project, made in-depth analysis impractical in some instances. Of the methods tested (dry brushing, vacuum cleaning, water washing, strippable coatings, and mechanical grinding), the most practical seems to be water washing. Effectiveness is very high, and it is very easy and rapid to deploy. The amount of wastewater produced (2 L/m2) would be substantial and may not be easy to manage, but the method is the clear winner from a usability perspective. Removable surface contamination levels (smear results) from the strippable coating and water washing coupons found no residual removable contamination. Thus, whatever is left is likely adhered to (or trapped within) the salt. The other option that shows promise is the use of a fixative barrier. Bartlett Nuclear, Inc.’s Polymeric Barrier System (PBS) proved the most durable of the coatings tested. The coatings were not tested for contaminant entrapment, only for coating integrity and durability.

  15. Borehole Miner - Extendible Nozzle Development for Radioactive Waste Dislodging and Retrieval from Underground Storage Tanks

    SciTech Connect

    CW Enderlin; DG Alberts; JA Bamberger; M White

    1998-09-25

    This report summarizes development of borehole-miner extendible-nozzle water-jetting technology for dislodging and retrieving salt cake, sludge} and supernate to remediate underground storage tanks full of radioactive waste. The extendible-nozzle development was based on commercial borehole-miner technology.

  16. CHAPTER 5-RADIOACTIVE WASTE MANAGEMENT

    SciTech Connect

    Marra, J.

    2010-05-05

    The ore pitchblende was discovered in the 1750's near Joachimstal in what is now the Czech Republic. Used as a colorant in glazes, uranium was identified in 1789 as the active ingredient by chemist Martin Klaproth. In 1896, French physicist Henri Becquerel studied uranium minerals as part of his investigations into the phenomenon of fluorescence. He discovered a strange energy emanating from the material which he dubbed 'rayons uranique.' Unable to explain the origins of this energy, he set the problem aside. About two years later, a young Polish graduate student was looking for a project for her dissertation. Marie Sklodowska Curie, working with her husband Pierre, picked up on Becquerel's work and, in the course of seeking out more information on uranium, discovered two new elements (polonium and radium) which exhibited the same phenomenon, but were even more powerful. The Curies recognized the energy, which they now called 'radioactivity,' as something very new, requiring a new interpretation, new science. This discovery led to what some view as the 'golden age of nuclear science' (1895-1945) when countries throughout Europe devoted large resources to understand the properties and potential of this material. By World War II, the potential to harness this energy for a destructive device had been recognized and by 1939, Otto Hahn and Fritz Strassman showed that fission not only released a lot of energy but that it also released additional neutrons which could cause fission in other uranium nuclei leading to a self-sustaining chain reaction and an enormous release of energy. This suggestion was soon confirmed experimentally by other scientists and the race to develop an atomic bomb was on. The rest of the development history which lead to the bombing of Hiroshima and Nagasaki in 1945 is well chronicled. After World War II, development of more powerful weapons systems by the United States and the Soviet Union continued to advance nuclear science. It was this defense

  17. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM - 2011

    SciTech Connect

    West, B.; Waltz, R.

    2012-06-21

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

  18. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM 2009

    SciTech Connect

    West, B.; Waltz, R.

    2010-06-21

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

  19. Introduction to Special Section on Geophysical Investigations of Proposed Radioactive Waste Disposal Sites

    NASA Astrophysics Data System (ADS)

    Oliver, H. W.

    1987-07-01

    A symposium on "Geophysical Investigations of Proposed Radioactive Waste Disposal Sites" was held at the Fall Meeting of the American Geophysical Union, December 13, 1982. Since then, five of the papers presented at the symposium have been published in the Journal of Geophysical Research and an additional six papers are included in this issue. Three of the current papers involve geophysical research at Yucca Mountain, Nevada; two papers are on subsurface structure and fracturing of the Strath-Halladale granite in northern Scotland, a prime candidate for rad waste storage in the United Kingdom; and a general paper is included on the application of various geophysical methods for characterizing all the potential storage sites in the United States under consideration by the U.S. Department of Energy. In 1982, the following nine sites in the United States (Figure 1) were under consideration by the U.S. Department of Energy for the first U.S. repository of high-level radioactive waste (HLW). The host rock at each site is noted in parentheses (from NW to SE): Hanford, Washington (Miocene basalt flows); Yucca Mountain, Nevada (Tertiary tuff); Davis Canyon, Utah, (bedded salt); Lavender Canyon, Utah (bedded salt); Deaf Smith, Texas (Permian bedded salt); Swisher County, Texas (Permian bedded salt); Vacherie dome, Louisiana (domal salt); Richton dome, Mississippi (domal salt); and Cypress Creek dome, Mississippi (domal salt)

  20. Annual radioactive waste tank inspection program -- 1993

    SciTech Connect

    McNatt, F.G. Sr.

    1994-05-01

    Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1993 to evaluate these vessels, and evaluations based on data accrued by inspections made since the tanks were constructed, are the subject of this report. The 1993 inspection program revealed that the condition of the Savannah River Site waste tanks had not changed significantly from that reported in the previous annual report. No new leaksites were observed. No evidence of corrosion or materials degradation was observed in the waste tanks. However, degradation was observed on covers of the concrete encasements for the out-of-service transfer lines to Tanks 1 through 8.

  1. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM 2010

    SciTech Connect

    West, B.; Waltz, R.

    2011-06-23

    Aqueous radioactive wastes from Savannah River Site (SRS) separations and vitrification processes are contained in large underground carbon steel tanks. Inspections made during 2010 to evaluate these vessels and other waste handling facilities along with evaluations based on data from previous inspections are the subject of this report. The 2010 inspection program revealed that the structural integrity and waste confinement capability of the Savannah River Site waste tanks were maintained. All inspections scheduled per SRR-LWE-2009-00138, HLW Tank Farm Inspection Plan for 2010, were completed. Ultrasonic measurements (UT) performed in 2010 met the requirements of C-ESG-00006, In-Service Inspection Program for High Level Waste Tanks, Rev. 3, and WSRC-TR-2002-00061, Rev.6. UT inspections were performed on Tanks 30, 31 and 32 and the findings are documented in SRNL-STI-2010-00533, Tank Inspection NDE Results for Fiscal Year 2010, Waste Tanks 30, 31 and 32. A total of 5824 photographs were made and 1087 visual and video inspections were performed during 2010. Ten new leaksites at Tank 5 were identified in 2010. The locations of these leaksites are documented in C-ESR-G-00003, SRS High Level Waste Tank Leaksite Information, Rev.5. Ten leaksites at Tank 5 were documented during tank wall/annulus cleaning activities. None of these new leaksites resulted in a release to the environment. The leaksites were documented during wall cleaning activities and the waste nodules associated with the leaksites were washed away. Previously documented leaksites were reactivated at Tank 12 during waste removal activities.

  2. RADIOACTIVE WASTE STREAMS FROM VARIOUS POTENTIAL NUCLEAR FUEL CYCLE OPTIONS

    SciTech Connect

    Nick Soelberg; Steve Piet

    2010-11-01

    Five fuel cycle options, about which little is known compared to more commonly known options, have been studied in the past year for the United States Department of Energy. These fuel cycle options, and their features relative to uranium-fueled light water reactor (LWR)-based fuel cycles, include: • Advanced once-through reactor concepts (Advanced Once-Through, or AOT) – intended for high uranium utilization and long reactor operating life, use depleted uranium in some cases, and avoid or minimize used fuel reprocessing • Fission-fusion hybrid (FFH) reactor concepts – potential variations are intended for high uranium or thorium utilization, produce fissile material for use in power generating reactors, or transmute transuranic (TRU) and some radioactive fission product (FP) isotopes • High temperature gas reactor (HTGR) concepts - intended for high uranium utilization, high reactor thermal efficiencies; they have unique fuel designs • Molten salt reactor (MSR) concepts – can breed fissile U-233 from Th fuel and avoid or minimize U fuel enrichment, use on-line reprocessing of the used fuel, produce lesser amounts of long-lived, highly radiotoxic TRU elements, and avoid fuel assembly fabrication • Thorium/U-233 fueled LWR (Th/U-233) concepts – can breed fissile U-233 from Th fuel and avoid or minimize U fuel enrichment, and produce lesser amounts of long-lived, highly radiotoxic TRU elements. These fuel cycle options could result in widely different types and amounts of used or spent fuels, spent reactor core materials, and waste streams from used fuel reprocessing, such as: • Highly radioactive, high-burnup used metal, oxide, or inert matrix U and/or Th fuels, clad in Zr, steel, or composite non-metal cladding or coatings • Spent radioactive-contaminated graphite, SiC, carbon-carbon-composite, metal, and Be reactor core materials • Li-Be-F salts containing U, TRU, Th, and fission products • Ranges of separated or un-separated activation

  3. Salt disposal of heat-generating nuclear waste.

    SciTech Connect

    Leigh, Christi D.; Hansen, Francis D.

    2011-01-01

    This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from United

  4. 40 CFR 147.3005 - Radioactive waste injection wells.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... dispose of radioactive waste (as defined in 10 CFR part 20, appendix B, table II, but not including high level and transuranic waste and spent nuclear fuel covered by 40 CFR part 191) shall comply with the... 40 Protection of Environment 23 2014-07-01 2014-07-01 false Radioactive waste injection wells....

  5. 40 CFR 147.3005 - Radioactive waste injection wells.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... dispose of radioactive waste (as defined in 10 CFR part 20, appendix B, table II, but not including high level and transuranic waste and spent nuclear fuel covered by 40 CFR part 191) shall comply with the... 40 Protection of Environment 24 2012-07-01 2012-07-01 false Radioactive waste injection wells....

  6. 40 CFR 147.3005 - Radioactive waste injection wells.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... dispose of radioactive waste (as defined in 10 CFR part 20, appendix B, table II, but not including high level and transuranic waste and spent nuclear fuel covered by 40 CFR part 191) shall comply with the... 40 Protection of Environment 24 2013-07-01 2013-07-01 false Radioactive waste injection wells....

  7. 40 CFR 147.3005 - Radioactive waste injection wells.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... dispose of radioactive waste (as defined in 10 CFR part 20, appendix B, table II, but not including high level and transuranic waste and spent nuclear fuel covered by 40 CFR part 191) shall comply with the... 40 Protection of Environment 22 2010-07-01 2010-07-01 false Radioactive waste injection wells....

  8. 40 CFR 147.3005 - Radioactive waste injection wells.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... dispose of radioactive waste (as defined in 10 CFR part 20, appendix B, table II, but not including high level and transuranic waste and spent nuclear fuel covered by 40 CFR part 191) shall comply with the... 40 Protection of Environment 23 2011-07-01 2011-07-01 false Radioactive waste injection wells....

  9. Transporting Radioactive Waste: An Engineering Activity. Grades 5-12.

    ERIC Educational Resources Information Center

    HAZWRAP, The Hazardous Waste Remedial Actions Program.

    This brochure contains an engineering activity for upper elementary, middle school, and high school students that examines the transportation of radioactive waste. The activity is designed to inform students about the existence of radioactive waste and its transportation to disposal sites. Students experiment with methods to contain the waste and…

  10. Hyponatremia—What Is Cerebral Salt Wasting?

    PubMed Central

    Momi, Jasminder; Tang, Christopher M; Abcar, Antoine C; Kujubu, Dean A; Sim, John J

    2010-01-01

    Background: Hyponatremia is a common electrolyte imbalance in hospitalized patients. It is associated with significant morbidity and mortality, especially if the underlying cause is incorrectly diagnosed and not treated appropriately. Often, the hospitalist is faced with a clinical dilemma when a patient presents with hyponatremia of an unclear etiology and with uncertain volume status. Syndrome of inappropriate antidiuretic hormone (SIADH) is frequently diagnosed in this clinical setting, but cerebral salt wasting (CSW) is an important diagnosis to consider. Objective: We wanted to describe the diagnosis, treatment, and history of CSW to provide clinicians with a better understanding of the differential diagnosis for hyponatremia. Conclusion: CSW is a process of extracellular volume depletion due to a tubular defect in sodium transport. Two postulated mechanisms for CSW are the excess secretion of natriuretic peptides and the loss of sympathetic stimulation to the kidney. Making the distinction between CSW and SIADH is important because the treatment for the two conditions is very different. PMID:20740122

  11. Future radioactive liquid waste streams study

    SciTech Connect

    Rey, A.S.

    1993-11-01

    This study provides design planning information for the Radioactive Liquid Waste Treatment Facility (RLWTF). Predictions of estimated quantities of Radioactive Liquid Waste (RLW) and radioactivity levels of RLW to be generated are provided. This information will help assure that the new treatment facility is designed with the capacity to treat generated RLW during the years of operation. The proposed startup date for the RLWTF is estimated to be between 2002 and 2005, and the life span of the facility is estimated to be 40 years. The policies and requirements driving the replacement of the current RLW treatment facility are reviewed. Historical and current status of RLW generation at Los Alamos National Laboratory are provided. Laboratory Managers were interviewed to obtain their insights into future RLW activities at Los Alamos that might affect the amount of RLW generated at the Lab. Interviews, trends, and investigation data are analyzed and used to create scenarios. These scenarios form the basis for the predictions of future RLW generation and the level of RLW treatment capacity which will be needed at LANL.

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

    SciTech Connect

    Not Available

    1994-08-01

    This report presents a history of commercial low-level radioactive waste disposal in the United States, with emphasis on the history of six commercially operated low-level radioactive waste disposal facilities. The report includes a brief description of important steps that have been taken during the last decade to ensure the safe disposal of low-level radioactive waste in the 1990s and beyond. These steps include the issuance of comprehensive State and Federal regulations governing the disposal of low-level radioactive waste, and the enactment of Federal laws making States responsible for the disposal of such waste generated within their borders.

  13. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM- 2007

    SciTech Connect

    West, B; Ruel Waltz, R

    2008-06-05

    Aqueous radioactive wastes from Savannah River Site (SRS) separations and vitrification processes are contained in large underground carbon steel tanks. The 2007 inspection program revealed that the structural integrity and waste confinement capability of the Savannah River Site waste tanks were maintained. A very small amount of material had seeped from Tank 12 from a previously identified leaksite. The material observed had dried on the tank wall and did not reach the annulus floor. A total of 5945 photographs were made and 1221 visual and video inspections were performed during 2007. Additionally, ultrasonic testing was performed on four Waste Tanks (15, 36, 37 and 38) in accordance with approved inspection plans that met the requirements of WSRC-TR-2002- 00061, Revision 2 'In-Service Inspection Program for High Level Waste Tanks'. The Ultrasonic Testing (UT) In-Service Inspections (ISI) are documented in a separate report that is prepared by the ISI programmatic Level III UT Analyst. Tanks 15, 36, 37 and 38 are documented in 'Tank Inspection NDE Results for Fiscal Year 2007'; WSRC-TR-2007-00064.

  14. Decontamination of a radioactive waste liquid by electrodialysis

    SciTech Connect

    Lundstrom, J.E.

    1987-02-24

    An apparatus is described for the removal and recovery of acidic and radioactive components of a liquid waste stream comprising in combination a series of at least two electrodialysis units or stacks. Each stack is comprised of a cathode chamber at one terminal end, and an anode chamber at the opposite terminal end. The chambers contain respectively a cathode and anode electrode. A deacidification stack is the first stack in the series having all of its chambers being separated one from the other by separate anion selective membranes defining between the electrode chambers at least one neutral liquid chamber positioned adjacent to the cathode chamber. A primary desalting stack is the second stack in the series comprising a multi-chamber unit having alternating salt diluting and salt concentrating chambers defined by alternating cation and anion selective membranes, means for introducing a liquid to be treated into the cathode chamber of the deacidification stack with exit means for withdrawal of the liquid, and means for passing the withdrawn liquid into and out of the salt diluting chambers of the primary desalting stack. A means is included for introducing a liquid into and out of the concentrating and electrode chambers of the primary desalting stack and for passing a direct electric current transversely across the membranes and chambers of each stack in the series.

  15. Significance of radiation effects in solid radioactive waste

    SciTech Connect

    Permar, P H; McDonell, W R

    1980-01-01

    Proposed NRC criteria for disposal of high-level nuclear waste require development of waste packages to contain radionuclide for at least 1000 years, and design of repositories to prevent radionuclide release at an annual rate greater than 1 part in 100,000 of the total activity. The high-level wastes that are now temporarily stored as aqueous salts, sludges, and calcines must be converted to high-integrity solid forms that resist deterioration from radiation and other effects of long-term storage. Spent fuel may be encapsulated for similar long-term storage. Candidate waste forms beside the spent fuel elements themselves, include borosilicate and related glasses, mineral-like crystalline ceramics, concrete formulations, and metal-matrix glass or ceramic composites. these waste forms will sustain damage produced by beta-gamma radiation up to 10/sup 12/ rads, by alpha radiation up to 10/sup 19/ particles/g, by internal helium generation greater than about 0.1 atom percent, and by the atom transmutations accompanying radioactive decay. Current data indicate that under these conditions the glass forms suffer only minor volume changes, stored energy deposition, and leachability effects. The crystalline ceramics appear susceptible to the potentially more severe alterations accompanying metamictization and natural analogs of candidate materials are being examined to establish their suitability as waste forms. Helium concentrations in the waste forms are generally below thresholds for severe damage in either glass or crystalline ceramics at low temperatures, but microstructural effects are not well characterized. Transmutation effects remain to be established.

  16. Radioactive waste disposal via electric propulsion

    NASA Technical Reports Server (NTRS)

    Burns, R. E.

    1975-01-01

    It is shown that space transportation is a feasible method of removal of radioactive wastes from the biosphere. The high decay heat of the isotopes powers a thermionic generator which provides electrical power for ion thrust engines. The massive shields (used to protect ground and flight personnel) are removed in orbit for subsequent reuse; the metallic fuel provides a shield for the avionics that guides the orbital stage to solar system escape. Performance calculations indicate that 4000 kg. of actinides may be removed per Shuttle flight. Subsidiary problems - such as cooling during ascent - are discussed.

  17. Radioactive Waste Management Complex performance assessment: Draft

    SciTech Connect

    Case, M.J.; Maheras, S.J.; McKenzie-Carter, M.A.; Sussman, M.E.; Voilleque, P.

    1990-06-01

    A radiological performance assessment of the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory was conducted to demonstrate compliance with appropriate radiological criteria of the US Department of Energy and the US Environmental Protection Agency for protection of the general public. The calculations involved modeling the transport of radionuclides from buried waste, to surface soil and subsurface media, and eventually to members of the general public via air, ground water, and food chain pathways. Projections of doses were made for both offsite receptors and individuals intruding onto the site after closure. In addition, uncertainty analyses were performed. Results of calculations made using nominal data indicate that the radiological doses will be below appropriate radiological criteria throughout operations and after closure of the facility. Recommendations were made for future performance assessment calculations.

  18. Membrane purification in radioactive waste management: a short review.

    PubMed

    Ambashta, Ritu D; Sillanpää, Mika E T

    2012-02-01

    Radiation hazards of radionuclides arising from nuclear plant facilities are well known. Separation technologies are used to concentrate the radionuclides and prevent the spread of this hazard to the environment. The present review describes the recent advances made in radioactive waste treatment using membrane separation technology. The first part discusses the membrane methods for collective separation of radionuclides and the second part discusses the membrane methods for selective separation of individual radionuclides. For the collection separation of radionulides, methods include reverse osmosis, precipitation followed by ultrafiltration or microfiltration and membrane distillation. Individual elements have been separated using liquid supported membranes, polymer inclusion membranes, solid polymer based electrolysis, nanofiltration, electrochemical salt-splitting process and other advanced separation methods. PMID:22204752

  19. Expected brine movement at potential nuclear waste repository salt sites

    SciTech Connect

    McCauley, V.S.; Raines, G.E.

    1987-08-01

    The BRINEMIG brine migration code predicts rates and quantities of brine migration to a waste package emplaced in a high-level nuclear waste repository in salt. The BRINEMIG code is an explicit time-marching finite-difference code that solves a mass balance equation and uses the Jenks equation to predict velocities of brine migration. Predictions were made for the seven potentially acceptable salt sites under consideration as locations for the first US high-level nuclear waste repository. Predicted total quantities of accumulated brine were on the order of 1 m/sup 3/ brine per waste package or less. Less brine accumulation is expected at domal salt sites because of the lower initial moisture contents relative to bedded salt sites. Less total accumulation of brine is predicted for spent fuel than for commercial high-level waste because of the lower temperatures generated by spent fuel. 11 refs., 36 figs., 29 tabs.

  20. Radioactive Waste Packaging of Conditioned Waste at Kozloduy NPP Site

    SciTech Connect

    Genchev, G.; Dimov, D.; Russev, K.

    2006-07-01

    An important part of Safety Management of conditioned low and intermediate level Radioactive Waste (RAW) is their packaging and containers for transport, storage and final disposal. A reinforced concrete container (RCC) has been developed to take cemented super compacted dry waste and cement solidified liquid waste at Kozloduy Nuclear Power Plant (KNPP). The container is to be used as a packaging of transportation, storage and final disposal of RAW conditioned by cementation KNPP specialists constructed and performed tests on the container. These tests were possible thanks to a review of European Community States experience, USA experience and IAEA documents. The container was tested by a team of specialists from KNPP, project specialists, fabricator of the containers and from Bulgarian Regulatory Body under IAEA Safety Standards, Safety Series, TECDOC, TRS and Bulgarian Standards. An expert from IAEA was a member of the testing group for RCC examinations. (authors)

  1. HANDBOOK: VITRIFICATION TECHNOLOGIES FOR TREATMENT OF HAZARDOUS AND RADIOACTIVE WASTE

    EPA Science Inventory

    The applications and limitations of vitrification technologies for treating hazardous and radioactive waste are presented. everal subgroups of vitrifications technologies exist. iscussions of glass structure, applicable waste types, off gas treatment, testing and evaluation proce...

  2. Issue briefs on low-level radioactive wastes

    SciTech Connect

    Not Available

    1981-01-01

    This report contains 4 Issue Briefs on low-level radioactive wastes. They are entitled: Handling, Packaging, and Transportation, Economics of LLW Management, Public Participation and Siting, and Low Level Waste Management.

  3. Closing Radioactive Waste Tanks in South Carolina

    SciTech Connect

    Newman, J.L.

    2000-08-29

    The Savannah River Site (SRS) is owned by the US Department of Energy (DOE) and is operated by the Westinghouse Savannah River Company (WSRC). Since the early 1950s, the primary mission of the site has been to produce nuclear materials for national defense. The chemical separations processes used to recover uranium and plutonium from production reactor fuel and target assemblies in the chemical separations area at SRS generated liquid high-level radioactive waste. This waste, which now amounts to approximately 34 million gallons, is stored in underground tanks in the F- and H-Areas near the center of the site. DOE is closing the High Level Waste (HLW) tank systems, which are permitted by SCDHEC under authority of the South Carolina Pollution Control Act (SCPCA) as wastewater treatment facilities, in accordance with South Carolina Regulation R.61-82, ''Proper Closeout of Wastewater Treatment Facilities''. To date, two HLW tank systems have been closed in place. Closure of these tanks is the first of its kind in the US. This paper describes the waste tank closure methodologies, standards and regulatory background.

  4. High-temperature vacuum distillation separation of plutonium waste salts

    SciTech Connect

    Garcia, E.

    1996-10-01

    In this task, high-temperature vacuum distillation separation is being developed for residue sodium chloride-potassium chloride salts resulting from past pyrochemical processing of plutonium. This process has the potential of providing clean separation of the salt and the actinides with minimal amounts of secondary waste generation. The process could produce chloride salt that could be discarded as low-level waste (LLW) or low actinide content transuranic (TRU) waste, and a concentrated actinide oxide powder that would meet long-term storage standards (DOE-DTD-3013-94) until a final disposition option for all surplus plutonium is chosen.

  5. The political science of radioactive waste disposal

    SciTech Connect

    Jacobi, L.R. Jr.

    1996-06-01

    This paper was first presented at the annual meeting of the HPS in New Orleans in 1984. Twelve years later, the basic lessons learned are still found to be valid. In 1984, the following things were found to be true: A government agency is preferred by the public over a private company to manage radioactive waste. Semantics are important--How you say it is important, but how it is heard is more important. Public information and public relations are very important, but they are the last thing of concern to a scientist. Political constituency is important. Don`t overlook the need for someone to be on your side. Don`t forget that the media is part of the political process-they can make you or break you. Peer technical review is important, but so is citizen review. Sociology is an important issue that scientists and technical people often overlook. In summary, despite the political nature of radioactive waste disposal, it is as true today as it was in 1984 that technical facts must be used to reach sound technical conclusions. Only then, separately and openly, should political factors be considered. So, what can be said today that wasn`t said in 1984? Nothing. {open_quotes}It`s deja vu all over again.{close_quotes}

  6. Area 5 Radioactive Waste Management Site Safety Assessment Document

    SciTech Connect

    Horton, K.K.; Kendall, E.W.; Brown, J.J.

    1980-02-01

    The Area 5 Radioactive Waste Management Safety Assessment Document evaluates site characteristics, facilities and operating practices which contribute to the safe handling and storage/disposal of radioactive wastes at the Nevada Test Site. Physical geography, cultural factors, climate and meteorology, geology, hydrology (with emphasis on radionuclide migration), ecology, natural phenomena, and natural resources are discussed and determined to be suitable for effective containment of radionuclides. Also considered, as a separate section, are facilities and operating practices such as monitoring; storage/disposal criteria; site maintenance, equipment, and support; transportation and waste handling; and others which are adequate for the safe handling and storage/disposal of radioactive wastes. In conclusion, the Area 5 Radioactive Waste Management Site is suitable for radioactive waste handling and storage/disposal for a maximum of twenty more years at the present rate of utilization.

  7. Legislative and regulatory aspects of radioactive waste management in France

    SciTech Connect

    Niel, J.C.

    1996-08-01

    The French legislative and regulatory framework for safe management of radioactive waste is presented. Emphasis is put on legislative aspects for the management of high-level waste and on the operation of surface disposal for low-level waste. Other topics such as policy and issues for very low-level waste or dismantling are also briefly developed.

  8. Karlsruhe Database for Radioactive Wastes (KADABRA) - Accounting and Management System for Radioactive Waste Treatment - 12275

    SciTech Connect

    Himmerkus, Felix; Rittmeyer, Cornelia

    2012-07-01

    The data management system KADABRA was designed according to the purposes of the Cen-tral Decontamination Department (HDB) of the Wiederaufarbeitungsanlage Karlsruhe Rueckbau- und Entsorgungs-GmbH (WAK GmbH), which is specialized in the treatment and conditioning of radioactive waste. The layout considers the major treatment processes of the HDB as well as regulatory and legal requirements. KADABRA is designed as an SAG ADABAS application on IBM system Z mainframe. The main function of the system is the data management of all processes related to treatment, transfer and storage of radioactive material within HDB. KADABRA records the relevant data concerning radioactive residues, interim products and waste products as well as the production parameters relevant for final disposal. Analytical data from the laboratory and non destructive assay systems, that describe the chemical and radiological properties of residues, production batches, interim products as well as final waste products, can be linked to the respective dataset for documentation and declaration. The system enables the operator to trace the radioactive material through processing and storage. Information on the actual sta-tus of the material as well as radiological data and storage position can be gained immediately on request. A variety of programs accessed to the database allow the generation of individual reports on periodic or special request. KADABRA offers a high security standard and is constantly adapted to the recent requirements of the organization. (authors)

  9. Controlled Containment, Radioactive Waste Management in the Netherlands

    SciTech Connect

    Codee, H.

    2002-02-26

    All radioactive waste produced in The Netherlands is managed by COVRA, the central organization for radioactive waste. The Netherlands forms a good example of a country with a small nuclear power program which will end in the near future. However, radioisotope production, nuclear research and other industrial activities will continue to produce radioactive waste. For the small volume, but broad spectrum of radioactive waste, including TENORM, The Netherlands has developed a management system based on the principles to isolate, to control and to monitor the waste. Long term storage is an essential element of the management system and forms a necessary step in the strategy of controlled containment that will ultimately result in final removal of the waste. Since the waste will remain retrievable for long time new technologies and new disposal options can be applied when available and feasible.

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

    SciTech Connect

    Not Available

    1990-10-01

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

  11. Salt caverns show promise for nonhazardous oil field waste disposal

    SciTech Connect

    Veil, J.A.

    1996-11-18

    Salt caverns show promise for the disposal of non-hazardous oil field wastes, and there are no apparent regulatory barriers to this application. Solution-mined salt caverns have been used for many years for storing hydrocarbon products. Argonne National laboratory has reviewed the legality, technical suitability, and feasibility of disposing of nonhazardous oil and gas exploration and production wastes in salt caverns. An analysis of regulations indicates that there are no outright regulatory prohibitions on cavern disposal of oil field wastes at either the federal level or in the 11 oil-producing states that were studied (Kansas, Louisiana, Michigan, Mississippi, New Mexico, New York, North Dakota, Ohio, Oklahoma, Pennsylvania, and Texas). The paper discusses the two types of salt deposits in the US, regulatory concerns, wastes, cavern design, disposal operations, closure and remediation, and results of the feasibility study.

  12. High-Level Radioactive Waste: Safe Storage and Ultimate Disposal.

    ERIC Educational Resources Information Center

    Dukert, Joseph M.

    Described are problems and techniques for safe disposal of radioactive waste. Degrees of radioactivity, temporary storage, and long-term permanent storage are discussed. Included are diagrams of estimated waste volumes to the year 2000 and of an artist's conception of a permanent underground disposal facility. (SL)

  13. Radioactive Waste Information for 1998 and Record-To-Date

    SciTech Connect

    D. L. French; R. E. Tallman; K. A. Taylor

    1999-07-01

    This document presents detailed data, bar graphs, and pie charts on volume, radioactivity; isotopic identity, origin, and status of radioactive waste for calendar year 1998 at the Idaho National Engineering and Environmental Laboratory (INEEL). The data presented are from the INEEL Integrated Waste Information System.

  14. An analysis of the technical status of high level radioactive waste and spent fuel management systems

    NASA Technical Reports Server (NTRS)

    English, T.; Miller, C.; Bullard, E.; Campbell, R.; Chockie, A.; Divita, E.; Douthitt, C.; Edelson, E.; Lees, L.

    1977-01-01

    The technical status of the old U.S. mailine program for high level radioactive nuclear waste management, and the newly-developing program for disposal of unreprocessed spent fuel was assessed. The method of long term containment for both of these waste forms is considered to be deep geologic isolation in bedded salt. Each major component of both waste management systems is analyzed in terms of its scientific feasibility, technical achievability and engineering achievability. The resulting matrix leads to a systematic identification of major unresolved technical or scientific questions and/or gaps in these programs.

  15. Process for converting sodium nitrate-containing, caustic liquid radioactive wastes to solid insoluble products

    DOEpatents

    Barney, Gary S.; Brownell, Lloyd E.

    1977-01-01

    A method for converting sodium nitrate-containing, caustic, radioactive wastes to a solid, relatively insoluble, thermally stable form is provided and comprises the steps of reacting powdered aluminum silicate clay, e.g., kaolin, bentonite, dickite, halloysite, pyrophyllite, etc., with the sodium nitrate-containing radioactive wastes which have a caustic concentration of about 3 to 7 M at a temperature of 30.degree. C to 100.degree. C to thereby entrap the dissolved radioactive salts in the aluminosilicate matrix. In one embodiment the sodium nitrate-containing, caustic, radioactive liquid waste, such as neutralized Purex-type waste, or salts or oxide produced by evaporation or calcination of these liquid wastes (e.g., anhydrous salt cake) is converted at a temperature within the range of 30.degree. C to 100.degree. C to the solid mineral form-cancrinite having an approximate chemical formula 2(NaAlSiO.sub.4) .sup.. xSalt.sup.. y H.sub.2 O with x = 0.52 and y = 0.68 when the entrapped salt is NaNO.sub.3. In another embodiment the sodium nitrate-containing, caustic, radioactive liquid is reacted with the powdered aluminum silicate clay at a temperature within the range of 30.degree. C to 100.degree. C, the resulting reaction product is air dried eitheras loose powder or molded shapes (e.g., bricks) and then fired at a temperature of at least 600.degree. C to form the solid mineral form-nepheline which has the approximate chemical formula of NaAlSiO.sub.4. The leach rate of the entrapped radioactive salts with distilled water is reduced essentially to that of the aluminosilicate lattice which is very low, e.g., in the range of 10.sup.-.sup.2 to 10.sup.-.sup.4 g/cm.sup.2 -- day for cancrinite and 10.sup.-.sup.3 to 10.sup.-.sup.5 g/cm.sup.2 -- day for nepheline.

  16. Taipower`s radioactive waste management program

    SciTech Connect

    Lee, B.C.C.

    1996-09-01

    Nuclear safety and radioactive waste management are the two major concerns of nuclear power in Taiwan. Recognizing that it is an issue imbued with political and social-economic concerns, Taipower has established an integrated nuclear backend management system and its associated financial and mechanism. For LLW, the Orchid Island storage facility will play an important role in bridging the gap between on-site storage and final disposal of LLW. Also, on-site interim storage of spent fuel for 40 years or longer will provide Taipower with ample time and flexibility to adopt the suitable alternative of direct disposal or reprocessing. In other words, by so exercising interim storage option, Taipower will be in a comfortable position to safely and permanently dispose of radwaste without unduly forgoing the opportunities of adopting better technologies or alternatives. Furthermore, Taipower will spare no efforts to communicate with the general public and make her nuclear backend management activities accountable to them.

  17. Hydrothermal processing of radioactive combustible waste

    SciTech Connect

    Worl, L.A.; Buelow, S.J.; Harradine, D.; Le, L.; Padilla, D.D.; Roberts, J.H.

    1998-09-01

    Hydrothermal processing has been demonstrated for the treatment of radioactive combustible materials for the US Department of Energy. A hydrothermal processing system was designed, built and tested for operation in a plutonium glovebox. Presented here are results from the study of the hydrothermal oxidation of plutonium and americium contaminated organic wastes. Experiments show the destruction of the organic component to CO{sub 2} and H{sub 2}O, with 30 wt.% H{sub 2}O{sub 2} as an oxidant, at 540 C and 46.2 MPa. The majority of the actinide component forms insoluble products that are easily separated by filtration. A titanium liner in the reactor and heat exchanger provide corrosion resistance for the oxidation of chlorinated organics. The treatment of solid material is accomplished by particle size reduction and the addition of a viscosity enhancing agent to generate a homogeneous pumpable mixture.

  18. Handling of liquid radioactive wastes produced during the decommissioning of nuclear-powered submarines

    SciTech Connect

    Martynov, B.V.

    1995-10-01

    Liquid radioactive wastes are produced during the standard decontamination of the reactor loop and liquidation of the consequences of accidents. In performing the disassembly work on decommissioned nuclear-powered submarines, the equipment must first be decontaminated. All this leads to the formation of a large quantity of liquid wastes with a total salt content of more then 3l-5 g/liter and total {beta}-activity of up to 1 {center_dot}10{sup {minus}4} Ci/liter. One of the most effective methods for reprocessing these wastes - evaporation - has limitations: The operating expenses are high and the apparatus requires expensive alloyed steel. The methods of selective sorption of radionuclides on inorganic sorbents are used for reprocessing liquid wastes form the nuclear-powered fleet. A significant limitation of the method is the large decrease in sorption efficiency with increasing total salt-content of the wastes. In some works, in which electrodialysis is used for purification of the salt wastes, the total salt content can be decreased by a factor of 10-100 and the same quantity of radionuclides can be removed. We have developed an electrodialysis-sorption scheme for purifying salt wastes that makes it possible to remove radionuclides to the radiation safety standard and chemically harmful substances to the health standards. The scheme includes electrodialysis desalinization (by 90% per pass on the EDMS apparatus), followed by additional purification of the diluent on synthetic zeolites and electro-osmotic concentration (to 200-250 g/liter on the EDK apparatus). The secondard wastes---salt concentrates and spent sorbents---are solidified. (This is the entire text of the article.)

  19. Management and disposal of waste from sites contaminated by radioactivity

    NASA Astrophysics Data System (ADS)

    Roberts, Carlyle J.

    1998-06-01

    Various methods of managing and disposing of wastes generated by decontamination and decommissioning (D & D) activities are described. This review of current waste management practices includes a description of waste minimization and volume reduction techniques and their applicability to various categories of radwaste. The importance of the physical properties of the radiation and radioactivity in determining the methodology of choice throughout the D & D process is stressed. The subject is introduced by a survey of the common types of radioactive contamination that must be managed and the more important hazards associated with each type. Comparisons are made among high level, transuranic, low level, and radioactive mixed waste, and technologically-enhanced, naturally-occurring radioactive material (TENORM). The development of appropriate clean-up criteria for each category of contaminated waste is described with the aid of examples drawn from actual practice. This includes a discussion of the application of pathway analysis to the derivation of residual radioactive material guidelines. The choice between interim storage and permanent disposal of radioactive wastes is addressed. Approaches to permanent disposal of each category of radioactive waste are described and illustrated with examples of facilities that have been constructed or are planned for implementation in the near future. Actual experience at older, existing, low-level waste disposal facilities is discussed briefly.

  20. Waste salt disposal at the Savannah River Plant. [Saltstone

    SciTech Connect

    Langton, C.A.; Oblath, S.B.; Pepper, D.W.; Wilhite, E.L.

    1986-01-01

    Waste salt solution, produced during processing of high-level nuclear waste, will be incorporated in a cement matrix for emplacement in an engineered disposal facility. Wasteform characteristics and disposal facility details will be presented along with results of a field test of wasteform contaminant release and of modeling studies to predict releases. 5 refs., 11 figs., 5 tabs.

  1. Leveraging Radioactive Waste Disposal at WIPP for Science

    NASA Astrophysics Data System (ADS)

    Rempe, N. T.

    2008-12-01

    Salt mines are radiologically much quieter than other underground environments because of ultra-low concentrations of natural radionuclides (U, Th, and K) in the host rock; therefore, the Waste Isolation Pilot Plant (WIPP), a government-owned, 655m deep geologic repository that disposes of radioactive waste in thick salt near Carlsbad, New Mexico, has for the last 15 years hosted highly radiation-sensitive experiments. Incidentally, Nature started her own low background experiment 250ma ago, preserving viable bacteria, cellulose, and DNA in WIPP salt. The Department of Energy continues to make areas of the WIPP underground available for experiments, freely offering its infrastructure and access to this unique environment. Even before WIPP started disposing of waste in 1999, the Room-Q alcove (25m x 10m x 4m) housed a succession of small experiments. They included development and calibration of neutral-current detectors by Los Alamos National Laboratory (LANL) for the Sudbury Neutrino Observatory, a proof-of-concept by Ohio State University of a flavor-sensitive neutrino detector for supernovae, and research by LANL on small solid- state dark matter detectors. Two currently active experiments support the search for neutrino-less double beta decay as a tool to better define the nature and mass of the neutrino. That these delicate experiments are conducted in close vicinity to, but not at all affected by, megacuries of radioactive waste reinforces the safety argument for the repository. Since 2003, the Majorana collaboration is developing and testing various detector designs inside a custom- built clean room in the Room-Q alcove. Already low natural background readings are reduced further by segmenting the germanium detectors, which spatially and temporally discriminates background radiation. The collaboration also demonstrated safe copper electro-forming underground, which minimizes cosmogenic background in detector assemblies. The largest currently used experimental

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

    SciTech Connect

    Burns, M.E.

    1988-01-01

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

  3. Technical evaluation of proposed Ukrainian Central Radioactive Waste Processing Facility

    SciTech Connect

    Gates, R.; Glukhov, A.; Markowski, F.

    1996-06-01

    This technical report is a comprehensive evaluation of the proposal by the Ukrainian State Committee on Nuclear Power Utilization to create a central facility for radioactive waste (not spent fuel) processing. The central facility is intended to process liquid and solid radioactive wastes generated from all of the Ukrainian nuclear power plants and the waste generated as a result of Chernobyl 1, 2 and 3 decommissioning efforts. In addition, this report provides general information on the quantity and total activity of radioactive waste in the 30-km Zone and the Sarcophagus from the Chernobyl accident. Processing options are described that may ultimately be used in the long-term disposal of selected 30-km Zone and Sarcophagus wastes. A detailed report on the issues concerning the construction of a Ukrainian Central Radioactive Waste Processing Facility (CRWPF) from the Ukrainian Scientific Research and Design institute for Industrial Technology was obtained and incorporated into this report. This report outlines various processing options, their associated costs and construction schedules, which can be applied to solving the operating and decommissioning radioactive waste management problems in Ukraine. The costs and schedules are best estimates based upon the most current US industry practice and vendor information. This report focuses primarily on the handling and processing of what is defined in the US as low-level radioactive wastes.

  4. Assessment of public perception of radioactive waste management in Korea.

    SciTech Connect

    Trone, Janis R.; Cho, SeongKyung; Whang, Jooho; Lee, Moo Yul

    2011-11-01

    The essential characteristics of the issue of radioactive waste management can be conceptualized as complex, with a variety of facets and uncertainty. These characteristics tend to cause people to perceive the issue of radioactive waste management as a 'risk'. This study was initiated in response to a desire to understand the perceptions of risk that the Korean public holds towards radioactive waste and the relevant policies and policy-making processes. The study further attempts to identify the factors influencing risk perceptions and the relationships between risk perception and social acceptance.

  5. Disposition of salt-waste from pyrochemical nuclear fuel processing

    SciTech Connect

    Vance, E.R.

    2007-07-01

    Waste salts from pyrochemical processing of nuclear fuel can be immobilised in sodalite if consolidated by hot isostatic pressing (HIP) at {approx}750 deg. C/100 MPa in thick stainless steel 316 cans. Other canning materials for this purpose also look possible. Spodiosite-based waste forms do not look promising in terms of leach resistance and their incorporation of alkali ions and compatibility with other phases which could potentially accommodate fission products, such as NaZr{sub 2}(PO{sub 4}){sub 3} or alumino-phosphate glass. Chloro- or fluor-apatite-based waste forms however have been reported to successfully accommodate fission products and alkalis which would be derived from either chloride- or fluoride-based waste pyro-processing salts. The presence of 10 or 20 wt% of additional Whitlockite, Ca{sub 3}(PO{sub 4}){sub 2}, should allow chemical flexibility to maintain the same qualitative phase assemblage when there are variations in the waste feed and in the waste/precursor ratios. Experimental verification of incorporation of the full complement of waste salts and fission products is not yet complete however. Apatite-rich samples could likely be HIPed in Inconel 600 cans. Other candidate HIP canning materials such as Alloy 22 or Inconel 625 are under study by encapsulating them in the candidate waste form and studying their interaction or otherwise with the waste form. (author)

  6. Solidification of radioactive waste in a cement/lime mixture

    SciTech Connect

    Zhou, H.; Colombo, P.

    1984-01-01

    The suitability of a cement/lime mixture for use as a solidification agent for different types of wastes was investigated. This work includes studies directed towards determining the wasted/binder compositional field over which successful solidification occurs with various wastes and the measurement of some of the waste from properties relevant to evaluating the potential for the release of radionuclides to the environment. In this study, four types of low-level radioactive wastes were simulated for incorporation into a cement/lime mixture. These were boric acid waste, sodium sulfate wastes, aion exchange resins and incinerator ash. 7 references, 3 figures, 2 tables.

  7. Health physics fundamentals, radiation protection, and radioactive waste treatment. Volume ten

    SciTech Connect

    Not Available

    1986-01-01

    Topics include health physics fundamentals (is radiation dangerous, what is health physics, federal regulations, presence of radiation, sources of radiation, types of radiation), radiation protection (amounts of radiation, radiation measurement, individual radiation exposure measurements, reducing the effects of radiation), and radioactive waste treatment (what are radioactive wastes, gaseous radioactive waste, liquid radioactive waste, solid radioactive waste, methods of rad-waste treatment, PWR and BWR radwaste treatment.

  8. Encapsulation of mixed radioactive and hazardous waste contaminated incinerator ash in modified sulfur cement

    SciTech Connect

    Kalb, P.D.; Heiser, J.H. III; Colombo, P.

    1990-01-01

    Some of the process waste streams incinerated at various Department of Energy (DOE) facilities contain traces of both low-level radioactive (LLW) and hazardous constituents, thus yielding ash residues that are classified as mixed waste. Work is currently being performed at Brookhaven National Laboratory (BNL) to develop new and innovative materials for encapsulation of DOE mixed wastes including incinerator ash. One such material under investigation is modified sulfur cement, a thermoplastic developed by the US Bureau of Mines. Monolithic waste forms containing as much as 55 wt % incinerator fly ash from Idaho national Engineering Laboratory (INEL) have been formulated with modified sulfur cement, whereas maximum waste loading for this waste in hydraulic cement is 16 wt %. Compressive strength of these waste forms exceeded 27.6 MPa. Wet chemical and solid phase waste characterization analyses performed on this fly ash revealed high concentrations of soluble metal salts including Pb and Cd, identified by the Environmental Protection Agency (EPA) as toxic metals. Leach testing of the ash according to the EPA Toxicity Characteristic Leaching Procedure (TCLP) resulted in concentrations of Pb and Cd above allowable limits. Encapsulation of INEL fly ash in modified sulfur cement with a small quantity of sodium sulfide added to enhance retention of soluble metal salts reduced TCLP leachate concentrations of Pb and Cd well below EPA concentration criteria for delisting as a toxic hazardous waste. 12 refs., 4 figs., 2 tabs.

  9. Innovative Process for Comprehensive Treatment of Liquid Radioactive Waste - 12551

    SciTech Connect

    Penzin, R.A.; Sarychev, G.A.

    2012-07-01

    the necessity to take emergency measures and to use marine water for cooling of reactor zone in contravention of the technological regulations. In these cases significant amount of liquid radioactive wastes of complex physicochemical composition is being generated, the purification of which by traditional methods is close to impossible. According to the practice of elimination of the accident after-effects at NPP 'Fukushima' there are still no technical means for the efficient purification of liquid radioactive wastes of complex composition like marine water from radionuclides. Therefore development of state-of-the-art highly efficient facilities capable of fast and safe purification of big amounts of liquid radioactive wastes of complex physicochemical composition from radionuclides turns to be utterly topical problem. Cesium radionuclides, being extremely dangerous for the environment, present over 90% of total radioactivity contained in liquid radioactive wastes left as a result of accidents at nuclear power objects. For the purpose of radiation accidents aftereffects liquidation VNIIHT proposes to create a plant for LRW reprocessing, consisting of 4 major technological modules: Module of LRW pretreatment to remove mechanical and organic impurities including oil products; Module of sorption purification of LWR by means of selective inorganic sorbents; Module of reverse osmotic purification and desalination; Module of deep evaporation of LRW concentrates. The first free modules are based on completed technological and designing concepts implemented by VNIIHT in the framework of LLRW Project in the period of 2000-2001 in Russia for comprehensive treatment of LWR of atomic fleet. These industrial plants proved to be highly efficient and secure during their long operation life. Module of deep evaporation is a new technological development. It will ensure conduction of evaporation and purification of LRW of different physicochemical composition, including those

  10. Disposal of NORM-contaminated oil field wastes in salt caverns -- Legality, technical feasibility, economics, and risk

    SciTech Connect

    Veil, J.A.; Smith, K.P.; Tomasko, D.; Elcock, D.; Blunt, D.; Williams, G.P.

    1998-07-01

    Some types of oil and gas production and processing wastes contain naturally occurring radioactive materials (NORM). If NORM is present at concentrations above regulatory levels in oil field waste, the waste requires special disposal practices. The existing disposal options for wastes containing NORM are limited and costly. This paper evaluates the legality, technical feasibility, economics, and human health risk of disposing of NORM-contaminated oil field wastes in salt caverns. Cavern disposal of NORM waste is technically feasible and poses a very low human health risk. From a legal perspective, there are no fatal flaws that would prevent a state regulatory agency from approaching cavern disposal of NORM. On the basis of the costs charged by caverns currently used for disposal of nonhazardous oil field waste (NOW), NORM waste disposal caverns could be cost competitive with existing NORM waste disposal methods when regulatory agencies approve the practice.

  11. Elimination of liquid discharge to the environment from the TA-50 Radioactive Liquid Waste Treatment Facility

    SciTech Connect

    Moss, D.; Williams, N.; Hall, D.; Hargis, K.; Saladen, M.; Sanders, M.; Voit, S.; Worland, P.; Yarbro, S.

    1998-06-01

    Alternatives were evaluated for management of treated radioactive liquid waste from the radioactive liquid waste treatment facility (RLWTF) at Los Alamos National Laboratory. The alternatives included continued discharge into Mortandad Canyon, diversion to the sanitary wastewater treatment facility and discharge of its effluent to Sandia Canyon or Canada del Buey, and zero liquid discharge. Implementation of a zero liquid discharge system is recommended in addition to two phases of upgrades currently under way. Three additional phases of upgrades to the present radioactive liquid waste system are proposed to accomplish zero liquid discharge. The first phase involves minimization of liquid waste generation, along with improved characterization and monitoring of the remaining liquid waste. The second phase removes dissolved salts from the reverse osmosis concentrate stream to yield a higher effluent quality. In the final phase, the high-quality effluent is reused for industrial purposes within the Laboratory or evaporated. Completion of these three phases will result in zero discharge of treated radioactive liquid wastewater from the RLWTF.

  12. Sulfur polymer cement, a final waste form for radioactive and hazardous wastes

    SciTech Connect

    Darnell, G.R.

    1996-12-31

    Because of its unusual properties, sulfur polymer cement (SPC) is a promising solidification and stabilization agent for radioactive and hazardous wastes. SPC accepts no water and requires no activation agents. It always melts at 115 C and pours at 135 C; therefore, economical remediation is offered through remelt and addition of additives or more SPC to meet specifications. Compressive strength upon cooling is approximately 27.6 MPa (4,000 psi). SPC has survived for years in acids and salts that destroy or severely damage hydraulic concretes in months or even weeks. In tests with 5 wt% loading of pure toxic metal oxides in powder form, the US Environmental Protection Agency`s the toxicity characteristic leaching procedure shows the maximum leachate concentration of mercury, lead, silver, arsenic, barium, and chromium to be less than their established threshold limits. Tests to determine SPC`s expected longevity are being conducted and are encouraging.

  13. Development of characterization protocol for mixed liquid radioactive waste classification

    SciTech Connect

    Zakaria, Norasalwa; Wafa, Syed Asraf; Wo, Yii Mei; Mahat, Sarimah

    2015-04-29

    Mixed liquid organic waste generated from health-care and research activities containing tritium, carbon-14, and other radionuclides posed specific challenges in its management. Often, these wastes become legacy waste in many nuclear facilities and being considered as ‘problematic’ waste. One of the most important recommendations made by IAEA is to perform multistage processes aiming at declassification of the waste. At this moment, approximately 3000 bottles of mixed liquid waste, with estimated volume of 6000 litres are currently stored at the National Radioactive Waste Management Centre, Malaysia and some have been stored for more than 25 years. The aim of this study is to develop a characterization protocol towards reclassification of these wastes. The characterization protocol entails waste identification, waste screening and segregation, and analytical radionuclides profiling using various analytical procedures including gross alpha/ gross beta, gamma spectrometry, and LSC method. The results obtained from the characterization protocol are used to establish criteria for speedy classification of the waste.

  14. Radioactive-waste container with leak monitor

    SciTech Connect

    Janberg, K.G.; Methling, D.

    1985-01-22

    A container has a massive metallic vessel whose interior is adapted to receive radioactive waste and whose mouth is formed with inner and outer spaced generally planar and annular vessel shoulders and formed there-between with a nonplanar intermediate annular vessel surface. A massive metallic cover formed with a plug fits in the mouth and has respective inner and outer plug shoulders closely juxtaposed with the vessel shoulders and a nonplanar intermediate annular plug surface complementary to the intermediate vessel surface. An inner ring seal engages snugly between the inner shoulders. A pair of generally concentric and spaced outer ring seals engage snugly between the outer shoulders and forming an annular outer chamber therebetween. An intermediate ring seal engages snugly between the intermediate surfaces and forms therebetween and with the inner ring seal an annular inner chamber and therebetween and with the outer ring seals an intermediate chamber. The cover is formed with respective inner, intermediate, and outer passages each having one end opening into the respective chamber and another end. Valves are provided on the cover at the other ends of the passages for sampling gases therein and in the respective chambers.

  15. Disposal of NORM-Contaminated Oil Field Wastes in Salt Caverns

    SciTech Connect

    Blunt, D.L.; Elcock, D.; Smith, K.P.; Tomasko, D.; Viel, J.A.; and Williams, G.P.

    1999-01-21

    In 1995, the U.S. Department of Energy (DOE), Office of Fossil Energy, asked Argonne National Laboratory (Argonne) to conduct a preliminary technical and legal evaluation of disposing of nonhazardous oil field waste (NOW) into salt caverns. That study concluded that disposal of NOW into salt caverns is feasible and legal. If caverns are sited and designed well, operated carefully, closed properly, and monitored routinely, they can be a suitable means of disposing of NOW (Veil et al. 1996). Considering these findings and the increased U.S. interest in using salt caverns for NOW disposal, the Office of Fossil Energy asked Argonne to conduct further research on the cost of cavern disposal compared with the cost of more traditional NOW disposal methods and on preliminary identification and investigation of the risks associated with such disposal. The cost study (Veil 1997) found that disposal costs at the four permitted disposal caverns in the United States were comparable to or lower than the costs of other disposal facilities in the same geographic area. The risk study (Tomasko et al. 1997) estimated that both cancer and noncancer human health risks from drinking water that had been contaminated by releases of cavern contents were significantly lower than the accepted risk thresholds. Since 1992, DOE has funded Argonne to conduct a series of studies evaluating issues related to management and disposal of oil field wastes contaminated with naturally occurring radioactive material (NORM). Included among these studies were radiological dose assessments of several different NORM disposal options (Smith et al. 1996). In 1997, DOE asked Argonne to conduct additional analyses on waste disposal in salt caverns, except that this time the wastes to be evaluated would be those types of oil field wastes that are contaminated by NORM. This report describes these analyses. Throughout the remainder of this report, the term ''NORM waste'' is used to mean ''oil field waste

  16. Disposal of NORM-contaminated oil field wastes in Salt Caverns.

    SciTech Connect

    Veil, J. A.; Smith, K. P.; Tomasko, D.; Elcock, D.; Blunt, D.; Williams, G. P.

    1998-08-28

    In 1995, the U.S. Department of Energy (DOE), Office of Fossil Energy, asked Argonne National Laboratory (Argonne) to conduct a preliminary technical and legal evaluation of disposing of nonhazardous oil field waste (NOW) into salt caverns. That study concluded that disposal of NOW into salt caverns is feasible and legal. If caverns are sited and designed well, operated carefully, closed properly, and monitored routinely, they can be a suitable means of disposing of NOW (Veil et al. 1996). Considering these findings and the increased U.S. interest in using salt caverns for NOW disposal, the Office of Fossil Energy asked Argonne to conduct further research on the cost of cavern disposal compared with the cost of more traditional NOW disposal methods and on preliminary identification and investigation of the risks associated with such disposal. The cost study (Veil 1997) found that disposal costs at the four permitted disposal caverns in the United States were comparable to or lower than the costs of other disposal facilities in the same geographic area. The risk study (Tomasko et al. 1997) estimated that both cancer and noncancer human health risks from drinking water that had been contaminated by releases of cavern contents were significantly lower than the accepted risk thresholds. Since 1992, DOE has funded Argonne to conduct a series of studies evaluating issues related to management and disposal of oil field wastes contaminated with naturally occurring radioactive material (NORM). Included among these studies were radiological dose assessments of several different NORM disposal options (Smith et al. 1996). In 1997, DOE asked Argonne to conduct additional analyses on waste disposal in salt caverns, except that this time the wastes to be evaluated would be those types of oil field wastes that are contaminated by NORM. This report describes these analyses. Throughout the remainder of this report, the term ''NORM waste'' is used to mean ''oil field waste

  17. Journey to the Nevada Test Site Radioactive Waste Management Complex

    SciTech Connect

    2007-08-01

    Journey to the Nevada Test Site Radioactive Waste Management Complex begins with a global to regional perspective regarding the location of low-level and mixed low-level waste disposal at the Nevada Test Site. For decades, the Nevada National Security Site (NNSS) has served as a vital disposal resource in the nation-wide cleanup of former nuclear research and testing facilities. State-of-the-art waste management sites at the NNSS offer a safe, permanent disposal option for U.S. Department of Energy/U.S. Department of Defense facilities generating cleanup-related radioactive waste.

  18. Journey to the Nevada Test Site Radioactive Waste Management Complex

    ScienceCinema

    None

    2014-10-28

    Journey to the Nevada Test Site Radioactive Waste Management Complex begins with a global to regional perspective regarding the location of low-level and mixed low-level waste disposal at the Nevada Test Site. For decades, the Nevada National Security Site (NNSS) has served as a vital disposal resource in the nation-wide cleanup of former nuclear research and testing facilities. State-of-the-art waste management sites at the NNSS offer a safe, permanent disposal option for U.S. Department of Energy/U.S. Department of Defense facilities generating cleanup-related radioactive waste.

  19. GIVE THE PUBLIC SOMETHING, SOMETHING MORE INTERESTING THAN RADIOACTIVE WASTE

    SciTech Connect

    Codee, Hans D.K.

    2003-02-27

    In the Netherlands the policy to manage radioactive waste is somewhat different from that in other countries, although the practical outcome is not much different. Long-term, i.e. at least 100 years, storage in above ground engineered structures of all waste types is the first element in the Dutch policy. Second element, but equally important, is that deep geologic disposal is foreseen after the storage period. This policy was brought out in the early eighties and was communicated to the public as a practical, logical and feasible management system for the Dutch situation. Strong opposition existed at that time to deep disposal in salt domes in the Netherlands. Above ground storage at principle was not rejected because the need to do something was obvious. Volunteers for a long term storage site did not automatically emerge. A site selection procedure was followed and resulted in the present site at Vlissingen-Oost. The waste management organization, COVRA, was not really welcomed here , but was tolerated. In the nineties facilities for low and medium level waste were erected and commissioned. In the design of the facilities much attention was given to emotional factors. The first ten operational years were needed to gain trust from the local population. Impeccable conduct and behavior was necessary as well as honesty and full openness to the public Now, after some ten years, the COVRA facilities are accepted. And a new phase is entered with the commissioning of the storage facility for high level waste, the HABOG facility. A visit to that facility will not be very spectacular, activities take place only during loading and unloading. Furthermore it is a facility for waste, so unwanted material will be brought into the community. In order to give the public something more interesting the building itself is transformed into a piece of art and in the inside a special work of art will be displayed. Together with that the attitude of the company will change. We are

  20. Bibliographic Data on Low-Level Radioactive Waste Documents

    Energy Science and Technology Software Center (ESTSC)

    1995-11-10

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

  1. Commentary: Radioactive Wastes and Damage to Marine Communities

    ERIC Educational Resources Information Center

    Wallace, Bruce

    1974-01-01

    Discusses the effects of radioactive wastes on marine communities, with particular reference to the fitness of populations and the need for field and laboratory studies to provide evidence of ecological change. (JR)

  2. Radioactive Waste Management in Non-Nuclear Countries - 13070

    SciTech Connect

    Kubelka, Dragan; Trifunovic, Dejan

    2013-07-01

    This paper challenges internationally accepted concepts of dissemination of responsibilities between all stakeholders involved in national radioactive waste management infrastructure in the countries without nuclear power program. Mainly it concerns countries classified as class A and potentially B countries according to International Atomic Energy Agency. It will be shown that in such countries long term sustainability of national radioactive waste management infrastructure is very sensitive issue that can be addressed by involving regulatory body in more active way in the infrastructure. In that way countries can mitigate possible consequences on the very sensitive open market of radioactive waste management services, comprised mainly of radioactive waste generators, operators of end-life management facilities and regulatory body. (authors)

  3. ACTINIDE-ALUMINATE SPECIATION IN ALKALINE RADIOACTIVE WASTE

    EPA Science Inventory

    Highly alkaline radioactive waste tanks contain a number of transuranic species, in particular U, Np, Pu, and Am - the exact forms of which are currently unknown. Knowledge of actinide speciation under highly alkaline conditions is essential towards understanding and predicting ...

  4. Natural diatomite process for removal of radioactivity from liquid waste.

    PubMed

    Osmanlioglu, Ahmet Erdal

    2007-01-01

    Diatomite has a number of unique physical properties and has found diversified industrial utilization. The filtration characteristics are particularly significant in the purification of liquids. The purpose of this study was to test natural diatomaceous earth (diatomite) as an alternative material that could be used for removal of radioactivity from liquid waste. A pilot-scale column-type device was designed. Natural diatomite samples were ground, sieved and prepared to use as sorption media. In this study, real waste liquid was used as radioactive liquid having special conditions. The liquid waste contained three radionuclides (Cs-137, Cs-134 and Co-60). Following the treatment by diatomite, the radioactivity of liquid waste was reduced from the initial 2.60 Bq/ml to less than 0.40 Bq/ml. The results of this study show that most of the radioactivity was removed from the solution by processing with diatomite. PMID:17049259

  5. Foaming and Antifoaming in Radioactive Waste Pretreatment and Immobilization Processes

    SciTech Connect

    Wasan, Darsh T.; Nikolov, Alex; Lambert, Dan; Calloway, T. Bond, Jr.

    2003-06-05

    The objective of this research is to develop a fundamental understanding of the physico-chemical mechanisms that cause foaminess in the DOE High Level (HLW) and Low Activity radioactive waste separation processes and to develop and test advanced antifoam/defoaming agents. Antifoams developed for this research will be tested using simulated defense HLW radioactive wastes obtained from the Hanford and Savannah River sites.

  6. Foaming and Antifoaming in Radioactive Waste Pretreatment and Immobilization Processes

    SciTech Connect

    Wasan, Darsh T.

    2002-08-01

    The objective of this research is to develop a fundamental understanding of the physico-chemical mechanisms that cause foaminess in the DOE High Level (HLW) and Low Activity radioactive waste separation processes and to develop and test advanced antifoam/defoaming agents. Antifoams developed for this research will be tested using simulated defense HLW radioactive wastes obtained from the Hanford and Savannah River sites.

  7. Radioactive Liquid Waste Treatment Facility: Environmental Information Document

    SciTech Connect

    Haagenstad, H.T.; Gonzales, G.; Suazo, I.L.

    1993-11-01

    At Los Alamos National Laboratory (LANL), the treatment of radioactive liquid waste is an integral function of the LANL mission: to assure U.S. military deterrence capability through nuclear weapons technology. As part of this mission, LANL conducts nuclear materials research and development (R&D) activities. These activities generate radioactive liquid waste that must be handled in a manner to ensure protection of workers, the public, and the environment. Radioactive liquid waste currently generated at LANL is treated at the Radioactive Liquid Waste Treatment Facility (RLWTF), located at Technical Area (TA)-50. The RLWTF is 30 years old and nearing the end of its useful design life. The facility was designed at a time when environmental requirements, as well as more effective treatment technologies, were not inherent in engineering design criteria. The evolution of engineering design criteria has resulted in the older technology becoming less effective in treating radioactive liquid wastestreams in accordance with current National Pollutant Discharge Elimination System (NPDES) and Department of Energy (DOE) regulatory requirements. Therefore, to support ongoing R&D programs pertinent to its mission, LANL is in need of capabilities to efficiently treat radioactive liquid waste onsite or to transport the waste off site for treatment and/or disposal. The purpose of the EID is to provide the technical baseline information for subsequent preparation of an Environmental Impact Statement (EIS) for the RLWTF. This EID addresses the proposed action and alternatives for meeting the purpose and need for agency action.

  8. The advantages of a salt/bentonite backfill for Waste Isolation Pilot Plant disposal rooms

    SciTech Connect

    Butcher, B.M.; Novak, C.F. ); Jercinovic, M. )

    1991-04-01

    A 70/30 wt% salt/bentonite mixture is shown to be preferable to pure crushed salt as backfill for disposal rooms in the Waste Isolation Pilot Plant (WIPP). This report discusses several selection criteria used to arrive at this conclusion: the need for low permeability and porosity after closure, chemical stability with the surroundings, adequate strength to avoid shear erosion from human intrusion, ease of emplacement, and sorption potential for brine and radionuclides. Both salt and salt/bentonite are expected to consolidate to a final state of impermeability (i.e., {le} 10{sup {minus}18}m{sup 2}) adequate for satisfying federal nuclear regulations. Any advantage of the salt/bentonite mixture is dependent upon bentonite's potential for sorbing brine and radionuclides. Estimates suggest that bentonite's sorption potential for water in brine is much less than for pure water. While no credit is presently taken for brine sorption in salt/bentonite backfill, the possibility that some amount of inflowing brine would be chemically bound is considered likely. Bentonite may also sorb much of the plutonium, americium, and neptunium within the disposal room inventory. Sorption would be effective only if a major portion of the backfill is in contact with radioactive brine. Brine flow from the waste out through highly localized channels in the backfill would negate sorption effectiveness. Although the sorption potentials of bentonite for both brine and radionuclides are not ideal, they are distinctly beneficial. Furthermore, no detrimental aspects of adding bentonite to the salt as a backfill have been identified. These two observations are the major reasons for selecting salt/bentonite as a backfill within the WIPP. 39 refs., 16 figs., 6 tabs.

  9. Laboratory simulation of salt dissolution during waste removal

    SciTech Connect

    Wiersma, B.J.; Parish, W.R.

    1997-01-01

    Laboratory experiments were performed to support the field demonstration of improved techniques for salt dissolution in waste tanks at the Savannah River Site. The tests were designed to investigate three density driven techniques for salt dissolution: (1) Drain-Add-Sit-Remove, (2) Modified Density Gradient, and (3) Continuous Salt Mining. Salt dissolution was observed to be a very rapid process as salt solutions with densities between 1.38-1.4 were frequently removed. Slower addition and removal rates and locating the outlet line at deeper levels below the top of the saltcake provided the best contact between the dissolution water and the saltcake. It was observed that dissolution with 1 M sodium hydroxide solution resulted in salt solutions that were within the current inhibitor requirements for the prevention of stress corrosion cracking. This result was independent of the density driven technique. However, if inhibited water (0.01 M sodium hydroxide and 0.011 M sodium nitrite) was utilized, the salt solutions were frequently outside the inhibitor requirements. Corrosion testing at conditions similar to the environments expected during waste removal was recommended.

  10. 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.

  11. Gasification characteristics of organic waste by molten salt

    NASA Astrophysics Data System (ADS)

    Sugiura, Kimihiko; Minami, Keishi; Yamauchi, Makoto; Morimitsu, Shinsuke; Tanimoto, Kazumi

    Recently, along with the growth in economic development, there has been a dramatic accompanying increase in the amount of sludge and organic waste. The disposal of such is a significant problem. Moreover, there is also an increased in the consumption of electricity along with economic growth. Although new energy development, such as fuel cells, has been promoted to solve the problem of power consumption, there has been little corresponding promotion relating to the disposal of sludge and organic waste. Generally, methane fermentation comprises the primary organic waste fuel used in gasification systems. However, the methane fermentation method takes a long time to obtain the fuel gas, and the quality of the obtained gas is unstable. On the other hand, gasification by molten salt is undesirable because the molten salt in the gasification gas corrodes the piping and turbine blades. Therefore, a gasification system is proposed by which the sludge and organic waste are gasified by molten salt. Moreover, molten carbonate fuel cells (MCFC) are needed to refill the MCFC electrolyte volatilized in the operation. Since the gasification gas is used as an MCFC fuel, MCFC electrolyte can be provided with the fuel gas. This paper elucidates the fundamental characteristics of sludge and organic waste gasification. A crucible filled with the molten salt comprising 62 Li 2CO 3/38 K 2CO 3, is installed in the reaction vessel, and can be set to an arbitrary temperature in a gas atmosphere. In this instance, the gasifying agent gas is CO 2. Sludge or the rice is supplied as organic waste into the molten salt, and is gasified. The chemical composition of the gasification gas is analyzed by a CO/CO 2 meter, a HC meter, and a SO x meter gas chromatography. As a result, although sludge can generate CO and H 2 near the chemical equilibrium value, all of the sulfur in the sludge is not fixed in the molten salt, because the sludge floats on the surface of the carbonate by the specific

  12. In situ electrochemical characterization of grouted radioactive waste

    SciTech Connect

    Gu, Jingyan; Shen, Wu-Mian; Tomkiewicz, Micha; Kruger, A.A.

    1993-04-01

    At the Hanford Site, twenty-eight double-shell tanks (DST) and one hundred and forty nine single-shell tanks (SST) are used for storage of radioactive liquid and sludge wastes and sat cake. A fundamental goal of the Westinghouse Hanford Company is to end the current storage practice for liquid wastes and to permanently dispose of the waste. The Hanford Defense Waste Environmental Impact Statement and subsequent record of decision has identified a cement-based waste form for disposal of DST low-level liquid waste. The low level radioactive fractions of these wastes will be immobilized in a cementitious grout at the Hanford Grout Processing Facility and disposed of in concrete vaults of the Grout Disposal Facility. Prior to closing each vault, postcuring verification will show that the final product meets the performance requirements. Any long term disposal system of radioactive waste will require monitoring to warn against structural deterioration and/or leach of the radioactive or hazardous components into the environment. We are investigating the possibility of monitoring the degree of immobilization of the waste by embedding a grid of long-lasting electrodes in grout. This work describes our ongoing attempts to understand the physics and chemistry of charge carriers in the grout under various load conditions.

  13. Office of Civilian Radioactive Waste Management annual report to Congress

    SciTech Connect

    1990-12-01

    This seventh Annual Report to Congress by the Office of Civilian Radioactive Waste Management (OCRWM) describes activities and expenditures of the Office during fiscal years (FY) 1989 and 1990. In November 1989, OCRWM is responsible for disposing of the Nation`s spent nuclear fuel and high-level radioactive waste in a manner that protects the health and safety of the public and the quality of the environment. To direct the implementation of its mission, OCRWM has established the following objectives: (1) Safe and timely disposal: to establish as soon as practicable the ability to dispose of radioactive waste in a geologic repository licensed by the NRC. (2) Timely and adequate waste acceptance: to begin the operation of the waste management system as soon as practicable in order to obtain the system development and operational benefits that have been identified for the MRS facility. (3) Schedule confidence: to establish confidence in the schedule for waste acceptance and disposal such that the management of radioactive waste is not an obstacle to the nuclear energy option. (4) System flexibility: to ensure that the program has the flexibility necessary for adapting to future circumstances while fulfilling established commitments. To achieve these objectives, OCRWM is developing a waste management system consisting of a geologic repository for permanent disposed deep beneath the surface of the earth, a facility for MRS, and a system for transporting the waste.

  14. Direct Grout Stabilization of High Cesium Salt Waste: Salt Alternative Phase III Feasibility Study

    SciTech Connect

    Langton, C.A.

    1998-12-07

    The direct grout alternative is a viable option for treatment/stabilization and disposal of salt waste containing Cs-137 concentrations of 1-3 Ci/gal. The composition of the direct grout salt solution is higher in sodium salts and contains up to a few hundred ppm Cs-137 more than the current reference salt solution. However it is still similar to the composition of the current reference salt solution. Consequently, the processing, setting, and leaching properties (including TCLP for Cr and Hg) of the direct grout and current saltstone waste forms are very similar. The significant difference between these waste solutions is that the high cesium salt solution will contain between 1 and 3 Curies of Cs-137 per gallon compared to a negligible amount in the current salt solution. This difference will require special engineering and shielding for a direct grout processing facility and disposal units to achieve acceptable radiation exposure conditions. The Cs-137 concentration in the direct grout salt solution will also affect the long-term curing temperature of the waste form since 4.84 Watts of energy are generated per 1000 Ci of Cs-137. The temperature rise of the direct grout during long-term curing has been calculated by A. Shaddy, SRTC.1 The effect of curing temperature on the strength, leaching and physical durability of the direct grout saltstone is described in this report. At the present time, long term curing at 90 degrees C appears to be unacceptable because of cracking which will affect the structural integrity as evaluated in the immersion test. (The experiments conducted in this feasibility study do not address the effect of cracking on leaching of contaminants other than Cr, Hg, and Cs.) No cracking of the direct grout or reference saltstone waste forms was observed for samples cured at 70 degrees C. At the present time the implications of waste form cracking at elevated curing temperatures has not been fully addressed. The direct grout falls within the

  15. Engineering Options Assessment Report: Nitrate Salt Waste Stream Processing

    SciTech Connect

    Anast, Kurt Roy

    2015-11-18

    This report examines and assesses the available systems and facilities considered for carrying out remediation activities on remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The assessment includes a review of the waste streams consisting of 60 RNS, 29 aboveground UNS, and 79 candidate belowground UNS containers that may need remediation. The waste stream characteristics were examined along with the proposed treatment options identified in the Options Assessment Report . Two primary approaches were identified in the five candidate treatment options discussed in the Options Assessment Report: zeolite blending and cementation. Systems that could be used at LANL were examined for housing processing operations to remediate the RNS and UNS containers and for their viability to provide repackaging support for remaining LANL legacy waste.

  16. Engineering Options Assessment Report. Nitrate Salt Waste Stream Processing

    SciTech Connect

    Anast, Kurt Roy

    2015-11-13

    This report examines and assesses the available systems and facilities considered for carrying out remediation activities on remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The assessment includes a review of the waste streams consisting of 60 RNS, 29 above-ground UNS, and 79 candidate below-ground UNS containers that may need remediation. The waste stream characteristics were examined along with the proposed treatment options identified in the Options Assessment Report . Two primary approaches were identified in the five candidate treatment options discussed in the Options Assessment Report: zeolite blending and cementation. Systems that could be used at LANL were examined for housing processing operations to remediate the RNS and UNS containers and for their viability to provide repackaging support for remaining LANL legacy waste.

  17. Dismantlement and Radioactive Waste Management of DPRK Nuclear Facilities

    SciTech Connect

    Jooho, W.; Baldwin, G. T.

    2005-04-01

    One critical aspect of any denuclearization of the Democratic People’s Republic of Korea (DPRK) involves dismantlement of its nuclear facilities and management of their associated radioactive wastes. The decommissioning problem for its two principal operational plutonium facilities at Yongbyun, the 5MWe nuclear reactor and the Radiochemical Laboratory reprocessing facility, alone present a formidable challenge. Dismantling those facilities will create radioactive waste in addition to existing inventories of spent fuel and reprocessing wastes. Negotiations with the DPRK, such as the Six Party Talks, need to appreciate the enormous scale of the radioactive waste management problem resulting from dismantlement. The two operating plutonium facilities, along with their legacy wastes, will result in anywhere from 50 to 100 metric tons of uranium spent fuel, as much as 500,000 liters of liquid high-level waste, as well as miscellaneous high-level waste sources from the Radiochemical Laboratory. A substantial quantity of intermediate-level waste will result from disposing 600 metric tons of graphite from the reactor, an undetermined quantity of chemical decladding liquid waste from reprocessing, and hundreds of tons of contaminated concrete and metal from facility dismantlement. Various facilities for dismantlement, decontamination, waste treatment and packaging, and storage will be needed. The shipment of spent fuel and liquid high level waste out of the DPRK is also likely to be required. Nuclear facility dismantlement and radioactive waste management in the DPRK are all the more difficult because of nuclear nonproliferation constraints, including the call by the United States for “complete, verifiable and irreversible dismantlement,” or “CVID.” It is desirable to accomplish dismantlement quickly, but many aspects of the radioactive waste management cannot be achieved without careful assessment, planning and preparation, sustained commitment, and long

  18. Annual Report - FY 2000, Radioactive Waste Shipments to and from the Nevada Test Site, March 2001

    SciTech Connect

    U.S. Department of Energy, Nevada Operations Office

    2001-03-01

    This document reports the low-level radioactive waste, mixed low-level radioactive waste, and Polychlorinated Biphenyl contaminated low-level waste transported to or from the Nevada Test Site during fiscal year 2000.

  19. Waste characterization for radioactive liquid waste evaporators at Argonne National Laboratory - West.

    SciTech Connect

    Christensen, B. D.

    1999-02-15

    Several facilities at Argonne National Laboratory - West (ANL-W) generate many thousand gallons of radioactive liquid waste per year. These waste streams are sent to the AFL-W Radioactive Liquid Waste Treatment Facility (RLWTF) where they are processed through hot air evaporators. These evaporators remove the liquid portion of the waste and leave a relatively small volume of solids in a shielded container. The ANL-W sampling, characterization and tracking programs ensure that these solids ultimately meet the disposal requirements of a low-level radioactive waste landfill. One set of evaporators will process an average 25,000 gallons of radioactive liquid waste, provide shielding, and reduce it to a volume of six cubic meters (container volume) for disposal. Waste characterization of the shielded evaporators poses some challenges. The process of evaporating the liquid and reducing the volume of waste increases the concentrations of RCIU regulated metals and radionuclides in the final waste form. Also, once the liquid waste has been processed through the evaporators it is not possible to obtain sample material for characterization. The process for tracking and assessing the final radioactive waste concentrations is described in this paper, The structural components of the evaporator are an approved and integral part of the final waste stream and they are included in the final waste characterization.

  20. Effect of sodium salt on anaerobic digestion of kitchen waste.

    PubMed

    Anwar, Naveed; Wang, Wen; Zhang, Jie; Li, Yeqing; Chen, Chang; Liu, Guangqing; Zhang, Ruihong

    2016-01-01

    The effect of different sodium salt concentration on anaerobic digestion of kitchen waste was investigated. The methane production performance, the corresponding methane production model and sodium salt inhibition model were studied, and the degradation efficiency was analyzed. With the increase of sodium salt concentration, the methane yield and the maximal methane production rate decreased along with the increase of lag phase time. The highest methane yield of 594 mL/g-VSadded (VS: volatile solids) was found with no sodium salt addition while the lowest was obtained with addition of 16 g/L NaCl. The declines of the methane yield were negligible when the sodium salt concentration was below 8 g/L, which corresponded to <10% inhibiting efficiency. In contrast, a sharp decrease of methane yield was observed with addition of >8 g/L NaCl (causing 17-80% inhibition). Five kinds of regression models were developed to describe the sodium salt inhibition efficiency, and the cubic regression model of y = 0.508 + 2.401x - 0.369x(2) + 0.033x(3) showed the best fitting. The volatile fatty acids/ethanol gradually accumulated along with the increase of the sodium salt concentration, and the volatile solid removal efficiency represented a gradual decline accordingly. It is recommended that the sodium salt concentration in the anaerobic digesters should be controlled below 8 g/L in order to avoid intense methane inhibition. PMID:27120641

  1. FOAMING IN RADIOACTIVE WASTE TREATMENT AND IMMOBILIZATION PROCESSES

    EPA Science Inventory

    The physical mechanisms of the formation of foam in radioactive waste treatment and waste immobilization processes are poorly understood. The objective of this research is to develop a basic understanding of the mechanisms that produce foaming, to identify the key parameters whic...

  2. Process for immobilizing radioactive boric acid liquid wastes

    SciTech Connect

    Greenhalgh, W.O.

    1986-06-17

    A method is described for immobilizing radioactive boric acid waste solutions comprising: neutralizing a boric acid waste solution containing radionuclides with calcium hydroxide and forming a precipitate, evaporating the precipitate to near dryness, and firing the dry precipitate to form a calcium borate glass product containing the radionuclides.

  3. 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.

  4. Discussions about safety criteria and guidelines for radioactive waste management.

    PubMed

    Yamamoto, Masafumi

    2011-07-01

    In Japan, the clearance levels for uranium-bearing waste have been established by the Nuclear Safety Commission (NSC). The criteria for uranium-bearing waste disposal are also necessary; however, the NSC has not concluded the discussion on this subject. Meanwhile, the General Administrative Group of the Radiation Council has concluded the revision of its former recommendation 'Regulatory exemption dose for radioactive solid waste disposal', the dose criteria after the institutional control period for a repository. The Standardization Committee on Radiation Protection in the Japan Health Physics Society (The Committee) also has developed the relevant safety criteria and guidelines for existing exposure situations, which are potentially applicable to uranium-bearing waste disposal. A new working group established by The Committee was initially aimed at developing criteria and guidelines specifically for uranium-bearing waste disposal; however, the aim has been shifted to broader criteria applicable to any radioactive wastes. PMID:21531746

  5. Issues in radioactive-waste management for fusion power

    SciTech Connect

    Maninger, R.C.; Dorn, D.W.

    1982-10-12

    Analysis of recent conceptual designs reveals that commercial fusion power systems will raise issues of occupational and public health and safety. This paper focuses on radioactive wastes from fusion reactor materials activated by neutrons. The analysis shows that different selections of materials and neutronic designs can make differences in orders-of-magnitude of the kinds and amounts of radioactivity to be expected. By careful and early evaluation of the impacts of the selections on waste management, designers can produce fusion power systems with radiation from waste well below today's limits for occupational and public health and safety.

  6. Industrial-Scale Processes For Stabilizing Radioactively Contaminated Mercury Wastes

    SciTech Connect

    Broderick, T. E.; Grondin, R.

    2003-02-24

    This paper describes two industrial-scaled processes now being used to treat two problematic mercury waste categories: elemental mercury contaminated with radionuclides and radioactive solid wastes containing greater than 260-ppm mercury. The stabilization processes were developed by ADA Technologies, Inc., an environmental control and process development company in Littleton, Colorado. Perma-Fix Environmental Services has licensed the liquid elemental mercury stabilization process to treat radioactive mercury from Los Alamos National Laboratory and other DOE sites. ADA and Perma-Fix also cooperated to apply the >260-ppm mercury treatment technology to a storm sewer sediment waste collected from the Y-12 complex in Oak Ridge, TN.

  7. Educational support programs: Office of Civilian Radioactive Waste Management

    SciTech Connect

    Williamson, R.C.

    1989-01-01

    The Office of Civilian Radioactive Waste Management (OCRWM) currently sponsors two educationally related programs: the Radioactive Waste Management Fellowship Program and the Radioactive Waste Management Research Program for Historically Black Colleges and Universities (HBCU). The graduate fellowship program was implemented in 1985 to meet the US Department of Energy's (DOE's) expected manpower needs for trained scientists and engineers to assist in carrying out the activities of the Nuclear Waste Policy Act. It is recognized that a shortage of master's and doctoral level scientists and engineers in disciplines supportive of the nation's high-level radioactive waste management (RWM) program may impede the DOE's ability to properly carry out its mission under the act. The fellowship program encourages talented undergraduate students to enter graduate programs designed to educate and train them in fields directly related to RWM. The program supports graduate students in various disciplines, including nuclear science and engineering, health physics, and certain area of geology and chemical engineering. It also encourages universities to support and improve research activities and academic programs related to the management of spent nuclear fuel and high-level radioactive waste.

  8. Comparison of modified sulfur cement and hydraulic cement for encapsulation of radioactive and mixed wastes

    SciTech Connect

    Kalb, P.D.; Heiser, J.H. III; Colombo, P.

    1990-01-01

    The majority of solidification/stabilization systems for low-level radioactive waste (LLW) and mixed waste, both in the commercial sector and at Department of Energy (DOE) facilities, utilize hydraulic cement (such as portland cement) to encapsulate waste materials and yield a monolithic solid waste form for disposal. A new and innovative process utilizing modified sulfur cement developed by the US Bureau of Mines has been applied at Brookhaven National Laboratory (BNL) for the encapsulation of many of these problem'' wastes. Modified sulfur cement is a thermoplastic material, and as such, it can be heated above it's melting point (120{degree}C), combined with dry waste products to form a homogeneous mixture, and cooled to form a monolithic solid product. Under sponsorship of the DOE, research and development efforts at BNL have successfully applied the modified sulfur cement process for treatment of a range of LLWs including sodium sulfate salts, boric acid salts, and incinerator bottom ash and for mixed waste contaminated incinerator fly ash. Process development studies were conducted to determine optimal waste loadings for each waste type. Property evaluation studies were conducted to test waste form behavior under disposal conditions by applying relevant performance testing criteria established by the Nuclear Regulatory Commission (for LLW) and the Environmental Protection Agency (for hazardous wastes). Based on both processing and performance considerations, significantly greater waste loadings were achieved using modified sulfur cement when compared with hydraulic cement. Technology demonstration of the modified sulfur cement encapsulation system using production-scale equipment is scheduled for FY 1991. 12 refs., 8 figs., 3 tabs.

  9. Injector nozzle for molten salt destruction of energetic waste materials

    DOEpatents

    Brummond, W.A.; Upadhye, R.S.

    1996-02-13

    An injector nozzle has been designed for safely injecting energetic waste materials, such as high explosives, propellants, and rocket fuels, into a molten salt reactor in a molten salt destruction process without premature detonation or back burn in the injection system. The energetic waste material is typically diluted to form a fluid fuel mixture that is injected rapidly into the reactor. A carrier gas used in the nozzle serves as a carrier for the fuel mixture, and further dilutes the energetic material and increases its injection velocity into the reactor. The injector nozzle is cooled to keep the fuel mixture below the decomposition temperature to prevent spontaneous detonation of the explosive materials before contact with the high-temperature molten salt bath. 2 figs.

  10. Injector nozzle for molten salt destruction of energetic waste materials

    DOEpatents

    Brummond, William A.; Upadhye, Ravindra S.

    1996-01-01

    An injector nozzle has been designed for safely injecting energetic waste materials, such as high explosives, propellants, and rocket fuels, into a molten salt reactor in a molten salt destruction process without premature detonation or back burn in the injection system. The energetic waste material is typically diluted to form a fluid fuel mixture that is injected rapidly into the reactor. A carrier gas used in the nozzle serves as a carrier for the fuel mixture, and further dilutes the energetic material and increases its injection velocity into the reactor. The injector nozzle is cooled to keep the fuel mixture below the decomposition temperature to prevent spontaneous detonation of the explosive materials before contact with the high-temperature molten salt bath.

  11. Foaming and Antifoaming in Radioactive Waste Pretreatment and Immobilization Processes

    SciTech Connect

    Darsh T. Wasan; Alex D. Nikolov; D.P. Lamber; T. Bond Calloway; M.E. Stone

    2005-03-12

    Savannah River National Laboratory (SRNL) has reported severe foaminess in the bench scale evaporation of the Hanford River Protection - Waste Treatment Plant (RPP-WPT) envelope C waste. Excessive foaming in waste evaporators can cause carryover of radionuclides and non-radioactive waste to the condensate system. The antifoams used at Hanford and tested by SRNL are believed to degrade and become inactive in high pH solutions. Hanford wastes have been known to foam during evaporation causing excessive down time and processing delays.

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

    DOEpatents

    Cowan, Richard G.; Blasewitz, Albert G.

    1982-01-01

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

  13. 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.

  14. Commercial low-level radioactive waste disposal in the US

    SciTech Connect

    Smith, P.

    1995-10-01

    Why are 11 states attempting to develop new low-level radioactive waste disposal facilities? Why is only on disposal facility accepting waste nationally? What is the future of waste disposal? These questions are representative of those being asked throughout the country. This paper attempts to answer these questions in terms of where we are, how we got there, and where we might be going.

  15. Removal of radioactive and other hazardous material from fluid waste

    DOEpatents

    Tranter, Troy J.; Knecht, Dieter A.; Todd, Terry A.; Burchfield, Larry A.; Anshits, Alexander G.; Vereshchagina, Tatiana; Tretyakov, Alexander A.; Aloy, Albert S.; Sapozhnikova, Natalia V.

    2006-10-03

    Hollow glass microspheres obtained from fly ash (cenospheres) are impregnated with extractants/ion-exchangers and used to remove hazardous material from fluid waste. In a preferred embodiment the microsphere material is loaded with ammonium molybdophosphonate (AMP) and used to remove radioactive ions, such as cesium-137, from acidic liquid wastes. In another preferred embodiment, the microsphere material is loaded with octyl(phenyl)-N-N-diisobutyl-carbamoylmethylphosphine oxide (CMPO) and used to remove americium and plutonium from acidic liquid wastes.

  16. Process for the encapsulation and stabilization of radioactive, hazardous and mixed wastes

    DOEpatents

    Colombo, Peter; Kalb, Paul D.; Heiser, III, John H.

    1997-11-14

    The present invention provides a method for encapsulating and stabilizing radioactive, hazardous and mixed wastes in a modified sulfur cement composition. The waste may be incinerator fly ash or bottom ash including radioactive contaminants, toxic metal salts and other wastes commonly found in refuse. The process may use glass fibers mixed into the composition to improve the tensile strength and a low concentration of anhydrous sodium sulfide to reduce toxic metal solubility. The present invention preferably includes a method for encapsulating radioactive, hazardous and mixed wastes by combining substantially anhydrous wastes, molten modified sulfur cement, preferably glass fibers, as well as anhydrous sodium sulfide or calcium hydroxide or sodium hydroxide in a heated double-planetary orbital mixer. The modified sulfur cement is preheated to about 135.degree..+-.5.degree. C., then the remaining substantially dry components are added and mixed to homogeneity. The homogeneous molten mixture is poured or extruded into a suitable mold. The mold is allowed to cool, while the mixture hardens, thereby immobilizing and encapsulating the contaminants present in the ash.

  17. Low-level radioactive waste disposal facility closure

    SciTech Connect

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

    1990-11-01

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

  18. Review of geochemical measurement techniques for a nuclear waste repository in bedded salt

    SciTech Connect

    Knauss, K.G.; Steinborn, T.L.

    1980-05-22

    A broad, general review is presented of geochemical measurement techniques that can provide data necessary for site selection and repository effectiveness assessment for a radioactive waste repository in bedded salt. The available measurement techniques are organized according to the parameter measured. The list of geochemical parameters include all those measurable geochemical properties of a sample whole values determine the geochemical characteristics or behavior of the system. For each technique, remarks are made pertaining to the operating principles of the measurement instrument and the purpose for which the technique is used. Attention is drawn to areas where further research and development are needed.

  19. Method for aqueous radioactive waste treatment

    DOEpatents

    Bray, L.A.; Burger, L.L.

    1994-03-29

    Plutonium, strontium, and cesium found in aqueous waste solutions resulting from nuclear fuel processing are removed by contacting the waste solutions with synthetic zeolite incorporating up to about 5 wt % titanium as sodium titanate in an ion exchange system. More than 99.9% of the plutonium, strontium, and cesium are removed from the waste solutions. 3 figures.

  20. Method for aqueous radioactive waste treatment

    DOEpatents

    Bray, Lane A.; Burger, Leland L.

    1994-01-01

    Plutonium, strontium, and cesium found in aqueous waste solutions resulting from nuclear fuel processing are removed by contacting the waste solutions with synthetic zeolite incorporating up to about 5 wt % titanium as sodium titanate in an ion exchange system. More than 99.9% of the plutonium, strontium, and cesium are removed from the waste solutions.

  1. Validation of Stress Corrosion Cracking Model for High Level Radioactive-Waste Packages

    SciTech Connect

    Lu, S; Gordon, G; Andresen, P

    2004-04-22

    A stress corrosion cracking (SCC) model has been adapted for performance prediction of high level radioactive-waste packages to be emplaced in the proposed Yucca Mountain radioactive-waste repository. SCC is one form of environmentally assisted cracking resulting from the presence of three factors: metallurgical susceptibility, critical environment, and tensile stresses. For waste packages of the proposed Yucca Mountain repository, the outer barrier material is the highly corrosion-resistant Alloy UNS-N06022, the environment is represented by the water film present on the surface of the waste package from dripping or deliquescence of soluble salts present in any surface deposits, and the stress is principally the weld induced residual stress. SCC has historically been separated into 'initiation' and 'propagation' phases. Initiation of SCC will not occur on a smooth surface if the surface stress is below a threshold value defined as the threshold stress. Cracks can also initiate at and propagate from flaws (or defects) resulting from manufacturing processes (such as welding). To account for crack propagation, the slip dissolution/film rupture (SDFR) model is adopted to provide mathematical formulae for prediction of the crack growth rate. Once the crack growth rate at an initiated SCC is determined, it can be used by the performance assessment (not in the scope of this paper) to determine the time to through-wall penetration for the waste package. This paper presents the development and validation of the SDFR crack growth rate model based on technical information in the literature as well as experimentally determined crack growth rates developed specifically for Alloy UNS- N06022 in environments relevant to high level radioactive-waste packages of the proposed Yucca Mountain radioactive-waste repository.

  2. Novel waste printed circuit board recycling process with molten salt

    PubMed Central

    Riedewald, Frank; Sousa-Gallagher, Maria

    2015-01-01

    The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450–470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl–KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. • The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept. • This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L. • The treated PCBs can be removed via leg B while the process is on-going. PMID:26150977

  3. Novel waste printed circuit board recycling process with molten salt.

    PubMed

    Riedewald, Frank; Sousa-Gallagher, Maria

    2015-01-01

    The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450-470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl-KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. •The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept.•This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L.•The treated PCBs can be removed via leg B while the process is on-going. PMID:26150977

  4. Radioactive waste management information for 1996 and record-to-date

    SciTech Connect

    French, D.L.; Lisee, D.J.; Taylor, K.A.

    1997-07-01

    This document presents detailed data, bar graphs, and pie charts on volume, radioactivity, isotopic identity, origin, and status of radioactive waste for calendar year 1996. It also summarizes the radioactive waste data records compiled from 1952 to present for the Idaho National Engineering and Environmental Laboratory (INEEL). The data presented are from the INEEL Radioactive Waste Management Information System.

  5. Radioactive waste management information for 1993 and record-to-date

    SciTech Connect

    Taylor, K.A.

    1994-07-01

    This document presents detailed data, bar graphs, and pie charts on volume, radioactivity, isotopic identity, origin, and decay status of radioactive waste for the calendar year 1993. It also summarizes the radioactive waste data records compiled from 1952 to present for the Idaho National Engineering Laboratory (INEL). The data presented are from the INEL Radioactive Waste Management Information System.

  6. Civilian Radioactive Waste Management System Requirements Document

    SciTech Connect

    C.A. Kouts

    2006-05-10

    The CRD addresses the requirements of Department of Energy (DOE) Order 413.3-Change 1, ''Program and Project Management for the Acquisition of Capital Assets'', by providing the Secretarial Acquisition Executive (Level 0) scope baseline and the Program-level (Level 1) technical baseline. The Secretarial Acquisition Executive approves the Office of Civilian Radioactive Waste Management's (OCRWM) critical decisions and changes against the Level 0 baseline; and in turn, the OCRWM Director approves all changes against the Level 1 baseline. This baseline establishes the top-level technical scope of the CRMWS and its three system elements, as described in section 1.3.2. The organizations responsible for design, development, and operation of system elements described in this document must therefore prepare subordinate project-level documents that are consistent with the CRD. Changes to requirements will be managed in accordance with established change and configuration control procedures. The CRD establishes requirements for the design, development, and operation of the CRWMS. It specifically addresses the top-level governing laws and regulations (e.g., ''Nuclear Waste Policy Act'' (NWPA), 10 Code of Federal Regulations (CFR) Part 63, 10 CFR Part 71, etc.) along with specific policy, performance requirements, interface requirements, and system architecture. The CRD shall be used as a vehicle to incorporate specific changes in technical scope or performance requirements that may have significant program implications. Such may include changes to the program mission, changes to operational capability, and high visibility stakeholder issues. The CRD uses a systems approach to: (1) identify key functions that the CRWMS must perform, (2) allocate top-level requirements derived from statutory, regulatory, and programmatic sources, and (3) define the basic elements of the system architecture and operational concept. Project-level documents address CRD requirements by further

  7. Computer-based supervisory control and data acquisition system for the radioactive waste evaporator

    SciTech Connect

    Pope, N.G.; Schreiber, S.B.; Yarbro, S.L.; Gomez, B.G.; Nekimken, H.L.; Sanchez, D.E.; Bibeau, R.A.; Macdonald, J.M.

    1994-12-01

    The evaporator process at TA-55 reduces the amount of transuranic liquid radioactive waste by separating radioactive salts from relatively low-level radioactive nitric acid solution. A computer-based supervisory control and data acquisition (SCADA) system has been installed on the process that allows the operators to easily interface with process equipment. Individual single-loop controllers in the SCADA system allow more precise process operation with less human intervention. With this system, process data can be archieved in computer files for later analysis. Data are distributed throughout the TA-55 site through a local area network so that real-time process conditions can be monitored at multiple locations. The entire system has been built using commercially available hardware and software components.

  8. A New Storage Facility for Institutional Radioactive Wastes at IPEN.

    PubMed

    Vicente, Roberto; Dellamano, José Claudio; Potiens, Ademar José

    2015-08-01

    IPEN, the Nuclear and Energy Research Institute in Sao Paulo, Brazil, has been managing the radioactive wastes generated in its own activities of research and radioisotope production as well as those received from many radioisotope users in the country since its start up in 1958. Final disposal options are presently unavailable for the wastes that cannot be managed by release after decay. Treated and untreated wastes including disused sealed radioactive sources and solid and liquid wastes containing radionuclides of the uranium and thorium series or fission and activation products are among the categories that are under safe and secure storage. This paper discusses the aspects considered in the design and describes the startup of a new storage facility for these wastes. PMID:26102323

  9. Radioactive waste disposal in simulated peat bog repositories

    SciTech Connect

    Schell, W.R.; Massey, C.D.

    1987-01-01

    The Low Level Radioactive Waste Policy Act of 1980 and the Low Level Radioactive Waste Policy Amendments Act of 1985 have required state governments to be responsible for providing low-level waste (LLW) disposal facilities in their respective areas. Questions are (a) is the technology sufficiently advanced to ensure that radioactive wastes can be stored for 300 to 1000 yr without entering into any uncontrolled area. (b) since actual experience does not exist for nuclear waste disposal over this time period, can the mathematical models developed be tested and verified using unequivocal data. (c) how can the public perception of the problem be addressed and the potential risk assessment of the hazards be communicated. To address the technical problems of nuclear waste disposal in the acid precipitation regions of the Northern Hemisphere, a project was initiated in 1984 to evaluate an alternative method of nuclear waste disposal that may not rely completely on engineered barriers to protect the public. Certain natural biogeochemical systems have been retaining deposited materials since the last Ice Age (12,000 to 15,000 yr). It is the authors belief that the biogeochemical system of wetlands and peat bogs may provide an example of an analogue for a nuclear waste repository system that can be tested and verified over a sufficient time period, at least for the LLW disposal problem.

  10. An evaluation of air effluent and workplace radioactivity monitoring at the Waste Isolation Pilot Plant

    SciTech Connect

    Bartlett, W.T. Environmental Evaluation Group, Albuquerque, NM )

    1993-02-01

    Improvements are needed in the Waste Isolation Pilot Plant (WIPP) air effluent and workplace radioactivity monitoring prior to receipt of radioactive wastes. This report provides a detailed review Zf radioactivity air monitoring regulatory requirements and related facility design requirements. Air monitoring data, supplied by the Westinghouse Isolation Division, are analyzed. The WIPP Final Safety Analysis Report (FSAR) requires that the WIPP radiological facilities always have multiple confinement barriers to prevent the accidental release of radioactive material to the environment. The Waste Handling Building has standard confinement barriers that satisfy the regulatory requirements, but the underground confinement barriers.include a more complex system for filtering air in the event of-an accidental release. A continuous air monitor (CAM) is an integral part of the underground confinement barrier strategy. For the last four years'' the reliability and sensitivity of the CAMs have been the subject of numerous reports and meetings which are summarized in this report. Data supplied to the Environmental Evaluation Group (EEG) show that the Station A CAM, which monitors the underground.exhaust, does not satisfy the requirements of the FSAR. The CAM system is not fail-safe, and operations appear to be affected by high levels of salt aerosol and poor detector performance. Additional test information is needed to establish the limits of CAM performance. Findings and recommendations are also provided on alternative monitoring methods, procedures and calculations.

  11. An evaluation of air effluent and workplace radioactivity monitoring at the Waste Isolation Pilot Plant

    SciTech Connect

    Bartlett, W.T. |

    1993-02-01

    Improvements are needed in the Waste Isolation Pilot Plant (WIPP) air effluent and workplace radioactivity monitoring prior to receipt of radioactive wastes. This report provides a detailed review Zf radioactivity air monitoring regulatory requirements and related facility design requirements. Air monitoring data, supplied by the Westinghouse Isolation Division, are analyzed. The WIPP Final Safety Analysis Report (FSAR) requires that the WIPP radiological facilities always have multiple confinement barriers to prevent the accidental release of radioactive material to the environment. The Waste Handling Building has standard confinement barriers that satisfy the regulatory requirements, but the underground confinement barriers.include a more complex system for filtering air in the event of-an accidental release. A continuous air monitor (CAM) is an integral part of the underground confinement barrier strategy. For the last four years`` the reliability and sensitivity of the CAMs have been the subject of numerous reports and meetings which are summarized in this report. Data supplied to the Environmental Evaluation Group (EEG) show that the Station A CAM, which monitors the underground.exhaust, does not satisfy the requirements of the FSAR. The CAM system is not fail-safe, and operations appear to be affected by high levels of salt aerosol and poor detector performance. Additional test information is needed to establish the limits of CAM performance. Findings and recommendations are also provided on alternative monitoring methods, procedures and calculations.

  12. Results of a radioactive waste course for high school teachers

    SciTech Connect

    Binney, S.E.; Dodd, B.; Ringle, J.C. )

    1992-01-01

    A week-long three-credit graduate-level course was taught in the summer of 1991 for high school science and social science teachers. Instruction was at the layperson's level, and no prior knowledge was presumed. The objectives of the course were to present factual information about radioactive waste, measure recipient's changes in perception about this topic, and measure the effect of the course on others. The technical half of the course involved discussions of basic nuclear physics, radiation control, biological effects of radiation, nuclear fuel cycle, waste management principles, Hanford waste issues, and transport of radioactive waste. The social half of the course included a discussion of values and attitudes, risk perception and communication, and public participation in a decision-making role-playing session. Discussions about radioactive-waste issues with high school teachers, especially of an extended nature such as discussed in this paper, are an effective tool because high school teachers are a critical link to reaching the next generation. Armed with facts presented in an open, nonconfrontational format, these teachers tend to become more understanding and accepting of radioactive-waste issues.

  13. Decontamination processes for low level radioactive waste metal objects

    SciTech Connect

    Longnecker, E.F.; Ichikawa, Sekigo; Kanamori, Osamu

    1996-12-31

    Disposal and safe storage of contaminated nuclear waste is a problem of international scope. Although the greatest volume of such waste is concentrated in the USA and former Soviet Union, Western Europe and Japan have contaminated nuclear waste requiring attention. Japan`s radioactive nuclear waste is principally generated at nuclear power plants since it has no nuclear weapons production. However, their waste reduction, storage and disposal problems may be comparable to that of the USA on an inhabited area basis when consideration is given to population density where Japan`s population, half that of the USA, lives in an area slightly smaller than that of California`s. If everyone`s backyard was in California, the USA might have insoluble radioactive waste reduction, storage and disposal problems. Viewing Japan`s contaminated nuclear waste as a national problem requiring solutions, as well as an economic opportunity, Morikawa began research and development for decontaminating low level radioactive nuclear waste seven years ago. As engineers and manufacturers of special machinery for many years Morikawa brings special electro/mechanical/pneumatic Skills and knowledge to solving these unique problems. Genden Engineering Services and Construction Company (GESC), an affiliate of Japan Atomic Power Company, recently joined with Morikawa in this R&D effort to decontaminate low level radioactive nuclear waste (LLW) and to substantially reduce the volume of such nuclear waste requiring long term storage. This paper will present equipment with both mechanical and chemical processes developed over these several years by Morikawa and most recently in cooperation with GESC.

  14. 78 FR 7818 - Request To Amend a License To Export Radioactive Waste

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-04

    ... COMMISSION Request To Amend a License To Export Radioactive Waste Pursuant to 10 CFR 110.70 (b) ``Public... radioactive The total Amend to: 1) Remove Mexico. December 28, 2012; January waste as slightly quantity... the (ETI) facility, the Class A radioactive secondary waste will waste imported in either be...

  15. Crystallization of rhenium salts in a simulated low-activity waste borosilicate glass

    SciTech Connect

    Riley, Brian J.; McCloy, John S.; Goel, Ashutosh; Liezers, Martin; Schweiger, Michael J.; Liu, Juan; Rodriguez, Carmen P.; Kim, Dong-Sang

    2013-04-01

    This study presents a new method for looking at the solubility of volatile species in simulated low-activity waste glass. The present study looking at rhenium salts is also applicable to real applications involving radioactive technetium salts. In this synthesis method, oxide glass powder is mixed with the volatiles species, vacuum-sealed in a fused quartz ampoule, and then heat-treated under vacuum in a furnace. This technique restricts the volatile species to the headspace above the melt but still within the sealed ampoule, thus maximizing the volatile concentration in contact with the glass. Various techniques were used to measure the solubility of rhenium in glass and include energy dispersive spectroscopy, wavelength dispersive spectroscopy, laser ablation inductively-coupled plasma mass spectroscopy, and inductively-coupled plasma optical emission spectroscopy. The Re-solubility in this glass was determined to be ~3004 parts per million Re atoms. Above this concentration, the salts separated out of the melt as inclusions and as a low viscosity molten salt phase on top of the melt observed during and after cooling. This salt phase was analyzed with X-ray diffraction, scanning electron microscopy as well as some of the other aforementioned techniques and identified to be composed of alkali perrhenate and alkali sulfate.

  16. Thermochemical Processing of Radioactive Waste Using Powder Metal Fuels

    SciTech Connect

    Ojovan, M. I.; Sobolev, I. A.; Dmitriev, S. A.; Panteleev, V. I.; Karlina, O. K.; Klimov. V. L.

    2003-02-25

    Problematic radioactive wastes were generated during various activities of both industrial facilities and research institutions usually in relative small amounts. These can be spent ion exchange resins, inorganic absorbents, wastes from research nuclear reactors, irradiated graphite, mixed, organic or chlorine-containing radioactive waste, contaminated soils, un-burnable heavily surface-contaminated materials, etc. Conventional treatment methods encounter serious problems concerning processing efficiency of such waste, e.g. complete destruction of organic molecules and avoiding of possible emissions of radionuclides, heavy metals and chemically hazardous species. Some contaminations cannot be removed from surface using common decontamination methods. Conditioning of ash residues obtained after treatment of solid radioactive waste including ashes received from treating problematic wastes also is a complicated task. Moreover due to relative small volume of specific type radioactive waste the development of target treatment procedures and facilities to conduct technological processes and their deployment could be economically unexpedient and ecologically no justified. Thermochemical processing technologies are used for treating and conditioning problematic radioactive wastes. The thermochemical processing uses powdered metal fuels (PMF) that are specifically formulated for the waste composition and react chemically with the waste components. The composition of the PMF is designed in such a way as to minimize the release of hazardous components and radionuclides in the off gas and to confine the contaminants in the ash residue. The thermochemical procedures allow decomposition of organic matter and capturing hazardous radionuclides and chemical species simultaneously. A significant advantage of thermochemical processing is its autonomy. Thermochemical treatment technologies use the energy of exothermic reactions in the mixture of radioactive or hazardous waste with PMF

  17. Commission operation. National Low-Level Radioactive Waste Management Program

    NASA Astrophysics Data System (ADS)

    1984-09-01

    Since Congress enacted the Low-Level Radioactive Waste Policy Act, the states have prepared to meet their responsibilities for management of low-level radioactive waste by entering into regional compacts. This option document is intended to provide a framework for the operation of a compact commission formed as the governing body of a low-level radioactive waste compact. The document is designed to be easily modified to meet the needs of various regional compacts. The ideas and format presented were taken in general from the Federal Administrative procedures Act, various state administrative procedures, and the state regulatory agencies' rules of procedure. Requirements of filing, time frames, and standard language are written from a legal perspective.

  18. Radioactive waste and contamination in the former Soviet Union

    SciTech Connect

    Suokko, K.; Reicher, D. )

    1993-04-01

    Decades of disregard for the hazards of radioactive waste have created contamination problems throughout the former Soviet Union rivaled only by the Chernobyl disaster. Although many civilian activities have contributed to radioactive waste problems, the nuclear weapons program has been by far the greatest culprit. For decades, three major weapons production facilities located east of the Ural Mountains operated in complete secrecy and outside of environmental controls. Referred to until recently only by their postal abbreviations, the cities of Chelyabinsk-65, Tomsk-7, and Krasnoyarsk-26 were open only to people who worked in them. The mismanagement of waste at these sites has led to catastrophic accidents and serious releases of radioactive materials. Lack of public disclosure, meanwhile, has often prevented proper medical treatment and caused delays in cleanup and containment. 5 refs.

  19. Radioactive solid waste handling at the Plutonium Finishing Plant

    SciTech Connect

    Manthos, E.J.

    1990-05-01

    The Plutonium Finishing Plant is located on the Hanford Site in the southeast section of Washington State. It has been in operation since 1949. The mission of the plant is to produce plutonium metal and related products for the US Department of Energy defense programs. Solid transuranic, low-level, and mixed wastes are generated at the plant, the radioactive contaminants in the waste being primarily alpha emitting. This paper discusses present waste-handling methods at the plant and recent changes that were made to improve waste characterization. 2 refs.

  20. The Constitution, waste facility performance standards, and radioactive waste classification: Is equal protection possible?

    SciTech Connect

    Eye, R.V.

    1993-03-01

    The process for disposal of so-called low-level radioactive waste is deadlocked at present. Supporters of the proposed near-surface facilities assert that their designs will meet minimum legal and regulatory standards currently in effect. Among opponents there is an overarching concern that the proposed waste management facilities will not isolate radiation from the biosphere for an adequate length of time. This clash between legal acceptability and a perceived need to protect the environment and public health by requiring more than the law demand sis one of the underlying reasons why the process is deadlocked. Perhaps the most exhaustive public hearing yet conducted on low-level radioactive waste management has recently concluded in Illinois. The Illinois Low-Level Radioactive Waste Disposal Facility Sitting Commission conducted 71 days of fact-finding hearings on the safety and suitability of a site near Martinsville, Illinois, to serve as a location for disposition of low-level radioactive waste. Ultimately, the siting commission rejected the proposed facility site for several reasons. However, almost all the reasons were related, to the prospect that, as currently conceived, the concrete barrier/shallow-land burial method will not isolate radioactive waste from the biosphere. This paper reviews the relevant legal framework of the radioactive waste classification system and will argue that it is inadequate for long-lived radionuclides. Next, the paper will present a case for altering the classification system based on high-level waste regulatory considerations.

  1. Radioactive waste management in the former USSR. Volume 3

    SciTech Connect

    Bradley, D.J.

    1992-06-01

    Radioactive waste materials--and the methods being used to treat, process, store, transport, and dispose of them--have come under increased scrutiny over last decade, both nationally and internationally. Nuclear waste practices in the former Soviet Union, arguably the world`s largest nuclear waste management system, are of obvious interest and may affect practices in other countries. In addition, poor waste management practices are causing increasing technical, political, and economic problems for the Soviet Union, and this will undoubtedly influence future strategies. this report was prepared as part of a continuing effort to gain a better understanding of the radioactive waste management program in the former Soviet Union. the scope of this study covers all publicly known radioactive waste management activities in the former Soviet Union as of April 1992, and is based on a review of a wide variety of literature sources, including documents, meeting presentations, and data base searches of worldwide press releases. The study focuses primarily on nuclear waste management activities in the former Soviet Union, but relevant background information on nuclear reactors is also provided in appendixes.

  2. Modeling Solute Thermokinetics in LiCI-KCI Molten Salt for Nuclear Waste Separation

    SciTech Connect

    Morgan, Dane; Eapen, Jacob

    2013-10-01

    Recovery of actinides is an integral part of a closed nuclear fuel cycle. Pyrometallurgical nuclear fuel recycling processes have been developed in the past for recovering actinides from spent metallic and nitride fuels. The process is essentially to dissolve the spent fuel in a molten salt and then extract just the actinides for reuse in a reactor. Extraction is typically done through electrorefining, which involves electrochemical reduction of the dissolved actinides and plating onto a cathode. Knowledge of a number of basic thermokinetic properties of salts and salt-fuel mixtures is necessary for optimizing present and developing new approaches for pyrometallurgical waste processing. The properties of salt-fuel mixtures are presently being studied, but there are so many solutes and varying concentrations that direct experimental investigation is prohibitively time consuming and expensive (particularly for radioactive elements like Pu). Therefore, there is a need to reduce the number of required experiments through modeling of salt and salt-fuel mixture properties. This project will develop first-principles-based molecular modeling and simulation approaches to predict fundamental thermokinetic properties of dissolved actinides and fission products in molten salts. The focus of the proposed work is on property changes with higher concentrations (up to 5 mol%) of dissolved fuel components, where there is still very limited experimental data. The properties predicted with the modeling will be density, which is used to assess the amount of dissolved material in the salt; diffusion coefficients, which can control rates of material transport during separation; and solute activity, which determines total solubility and reduction potentials used during electrorefining. The work will focus on La, Sr, and U, which are chosen to include the important distinct categories of lanthanides, alkali earths, and actinides, respectively. Studies will be performed using LiCl-KCl salt

  3. FINAL REPORT. POLYOXOMETALATES FOR RADIOACTIVE WASTE TREATMENT

    EPA Science Inventory

    The research was directed primarily towards the use of polyoxometalate complexes for separationof lanthanide, actinide, and technetium species from aqueous waste solutions, such as the HanfordTank Wastes. Selective binding of these species responsible for much of the high level...

  4. What a public-relations-model regarding radioactive waste implicates

    SciTech Connect

    Ohnishi, Teruaki |

    1996-12-31

    The behavior of public attitude to radioactive waste with time was investigated by using a mathematical model which was developed for estimating the extent of attitude change, being based on the assumption that the change of public attitude to a certain subject is caused by the information environment mainly formed by the newsmedia. Investigations were also made on the extent the public relations activity can contribute to the change of public opinion for the radioactive waste, and on the method of assortment and execution of various types of activity which brings the maximum change of attitude under a given condition of budget.

  5. Potential of pottery materials in manufacturing radioactive waste containers.

    PubMed

    Helal, A A; Alian, A M; Aly, H M; Khalifa, S M

    2003-07-01

    Various pottery materials were evaluated for possible use in manufacturing containers for radioactive waste. Their potential was examined from the viewpoints of the effectiveness of disposal and the changes induced in them by gamma rays. Samples of these materials were irradiated with high-energy neutrons and gamma rays in a reactor near its core. the physical and mechanical properties of the materials before and after gamma irradiation (in a 60Co gamma cell) were compared. The study showed that pottery materials are resistant to radiation. Therefore, they were proposed for manufacturing drums for disposal of radioactive waste of high gamma activity. PMID:12878117

  6. In-Situ Chemical Precipitation of Radioactive Liquid Waste - 12492

    SciTech Connect

    Osmanlioglu, Ahmet Erdal

    2012-07-01

    This paper presented in-situ chemical precipitation for radioactive liquid waste by using chemical agents. Results are reported on large-scale implementation on the removal of {sup 137}Cs, {sup 134}Cs and {sup 60}Co from liquid radioactive waste generating from Nuclear Research and Training Centre. Total amount of liquid radioactive waste was 35 m{sup 3} and main radionuclides were Cs-137, Cs- 134 and Co-60. Initial radioactivity concentration of the liquid waste was 2264, 17 and 9 Bq/liter for Cs-137, Cs-134 and Co-60 respectively. Potassium ferro cyanide was selected as chemical agent at high pH levels 8-10 according to laboratory tests. After the process, radioactive sludge precipitated at the bottom of the tank and decontaminated clean liquid was evaluated depending on discharge limits. By this precipitation method decontamination factors were determined as 60, 9 and 17 for Cs-137, Cs-134 and Co-60 respectively. At the bottom of the tank radioactive sludge amount was 0.98 m{sup 3}. It was transferred by sludge pumps to cementation unit for solidification. By in situ chemical processing 97% of volume reduction was achieved. Using the optimal concentration of 0.75 M potassium ferro cyanide about 98% of the {sup 137}Cs can be removed at pH 8. The Potassium ferro cyanide precipitation method could be used successfully in large scale applications with nickel and ferrum agents for removal of Cs-137, Cs-134 and Co- 60. Although DF values of laboratory test were much higher than in-situ implementation, liquid radioactive waste was decontaminated successfully by using potassium ferro cyanide. Majority of liquid waste were discharged as clean liquid. %97.2 volumetric amount of liquid waste was cleaned and discharged at the original site. Reduced amount of sludge transportation in drums is more economical and safer method than liquid transportation. Although DF values could be different for each of applications related to main specifications of original liquid waste, this

  7. Use plan for demonstration radioactive-waste incinerator

    SciTech Connect

    Cooley, L.R.; McCampbell, M.R.; Thompson, J.D.

    1982-04-01

    The University of Maryland at Baltimore was awarded a grant from the Department of Energy to test a specially modified incinerator to burn biomedical radioactive waste. In preparation for the incinerator, the Radiation Safety Office devised a comprehensive plan for its safe and effective use. The incinerator plan includes a discussion of regulations regarding on-site incineration of radioactive waste, plans for optimum use in burning four principal waste forms, controlled air incineration technology, and standard health physics safety practices; a use plan, including waste categorization and segregation, processing, and ash disposition; safety procedures, including personnel and area monitoring; and methods to evaluate the incinerator's effectiveness by estimating its volume reduction factors, mass and activity balances, and by determining the cost effectiveness of incineration versus commercial shallow land burial.

  8. The Defense Waste Processing Facility: Two Years of Radioactive Operation

    SciTech Connect

    Marra, S.L.; Gee, J.T.; Sproull, J.F.

    1998-05-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site in Aiken, SC is currently immobilizing high level radioactive sludge waste in borosilicate glass. The DWPF began vitrification of radioactive waste in May, 1996. Prior to that time, an extensive startup test program was completed with simulated waste. The DWPF is a first of its kind facility. The experience gained and data collected during the startup program and early years of operation can provide valuable information to other similar facilities. This experience involves many areas such as process enhancements, analytical improvements, glass pouring issues, and documentation/data collection and tracking. A summary of this experience and the results of the first two years of operation will be presented.

  9. 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.

  10. Method of encapsulating solid radioactive waste material for storage

    DOEpatents

    Bunnell, Lee Roy; Bates, J. Lambert

    1976-01-01

    High-level radioactive wastes are encapsulated in vitreous carbon for long-term storage by mixing the wastes as finely divided solids with a suitable resin, formed into an appropriate shape and cured. The cured resin is carbonized by heating under a vacuum to form vitreous carbon. The vitreous carbon shapes may be further protected for storage by encasement in a canister containing a low melting temperature matrix material such as aluminum to increase impact resistance and improve heat dissipation.

  11. Novel Solvent for the Simultaneous recovery of Radioactive Nuclides from Liquid Radioactive Wastes

    SciTech Connect

    Romanovskiy, Valeriy Nicholiavich; Smirnov, Lgor V.; Babain, Vasiliy A.; Todd, Terry A.; Brewer, Ken N.

    1999-10-07

    The present invention relates to solvents, and methods, for selectively extracting and recovering radionuclides, especially cesium and strontium, rare earths and actinides from liquid radioactive wastes. More specifically, the invention relates to extracting agent solvent compositions comprising complex organoboron compounds, substituted polyethylene glycols, and neutral organophosphorus compounds in a diluent. The preferred solvent comprises a chlorinated cobalt dicarbollide, diphenyl-dibutylmethylenecarbamoylphosphine oxide, PEG-400, and a diluent of phenylpolyfluoroalkyl sulfone. The invention also provides a method of using the invention extracting agents to recover cesium, strontium, rare earths and actinides from liquid radioactive waste.

  12. Automated NDT techniques in radioactive waste management

    SciTech Connect

    Barna, B.A.; Brown, B.W.; Anderson, B.C.

    1983-01-01

    The prime NDT method selected for characterization of the waste is real-time x-radiography (RTR). An RTR system specifically designed for the TRU waste inspection is currently being used to develop the best techniques for waste certification. It is based on a standard 420 kV constant potential x-ray machine with a rare-earth fluorescing screen (gadolinium oxysulfide) functioning as an image converter. The low-light-level image produced on the screen is picked up by a CCTV camera with an image intensifier coupled to a plumbicon imaging tube. The system was designed for automated waste container handling and translation. Image analysis is not currently automated, although the CCTV image is digitized to allow signal averaging and edge enhancement through digital filtering. The digitized image is available through an IEEE 488 I/O port for more sophisticated computerized analysis.

  13. Commercial radioactive waste minimization program development guidance

    SciTech Connect

    Fischer, D.K.

    1991-01-01

    This document is one of two prepared by the EG&G Idaho, Inc., Waste Management Technical Support Program Group, National Low-Level Waste Management Program Unit. One of several Department of Energy responsibilities stated in the Amendments Act of 1985 is to provide technical assistance to compact regions Host States, and nonmember States (to the extent provided in appropriations acts) in establishing waste minimization program plans. Technical assistance includes, among other things, the development of technical guidelines for volume reduction options. Pursuant to this defined responsibility, the Department of Energy (through EG&G Idaho, Inc.) has prepared this report, which includes guidance on defining a program, State/compact commission participation, and waste minimization program plans.

  14. Commercial radioactive waste minimization program development guidance

    SciTech Connect

    Fischer, D.K.

    1991-01-01

    This document is one of two prepared by the EG G Idaho, Inc., Waste Management Technical Support Program Group, National Low-Level Waste Management Program Unit. One of several Department of Energy responsibilities stated in the Amendments Act of 1985 is to provide technical assistance to compact regions Host States, and nonmember States (to the extent provided in appropriations acts) in establishing waste minimization program plans. Technical assistance includes, among other things, the development of technical guidelines for volume reduction options. Pursuant to this defined responsibility, the Department of Energy (through EG G Idaho, Inc.) has prepared this report, which includes guidance on defining a program, State/compact commission participation, and waste minimization program plans.

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

    SciTech Connect

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

    1980-10-01

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

  16. Beneficial role of conflict in radioactive waste management programs

    SciTech Connect

    Payne, B.A.; Williams, R.G.

    1985-01-01

    Of the technical, political, and social problems associated with radioactive waste management, least is known about the latter two. Lay persons tend to generalize negative attitudes about other nuclear activity to radioactive waste management. Thus, conflict appears inevitable between the general public, citizen action groups, and decision-makers on radioactive waste management. The basis of conflict, we believe, can be found in the value orientation of certain groups and in differing perceptions of risk. Research on similar controversial issues reveals that conflict may be beneficial in the long run by contributing to the public's participation level and understanding of the issues, and to the decision-makers' appreciation of the lay perspective. The paper is in three parts. First, we review the sources of conflict over radioactive waste management issues. The negative attitudes and fears of the public toward different types of projects involving radioactivity, value conflicts, and differential perceptions of risk are cited as sources. Next we discuss the consequences of conflict in terms of sociological theory. Finally, we discuss how conflict can be directed and managed to produce an informed decision-making process. When the public is sensitized to an issue, when prevailing attitudes on the issue are negative, and when perceived risks are high - all of which are characteristic of waste management issues - specific steps should be taken to establish a legitimate process of communication and interaction between the public and the sponsor agency. When conflict is recognized as inevitable, the goal of a communications program is no longer to avoid it. It is to use the increased awareness to increase knowledge about waste management issues and public participation in decisions so that the final solution is acceptable at some level to all parties.

  17. 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).

  18. Remote automated material handling of radioactive waste containers

    SciTech Connect

    Greager, T.M.

    1994-09-01

    To enhance personnel safety, improve productivity, and reduce costs, the design team incorporated a remote, automated stacker/retriever, automatic inspection, and automated guidance vehicle for material handling at the Enhanced Radioactive and Mixed Waste Storage Facility - Phase V (Phase V Storage Facility) on the Hanford Site in south-central Washington State. The Phase V Storage Facility, scheduled to begin operation in mid-1997, is the first low-cost facility of its kind to use this technology for handling drums. Since 1970, the Hanford Site`s suspect transuranic (TRU) wastes and, more recently, mixed wastes (both low-level and TRU) have been accumulating in storage awaiting treatment and disposal. Currently, the Hanford Site is only capable of onsite disposal of radioactive low-level waste (LLW). Nonradioactive hazardous wastes must be shipped off site for treatment. The Waste Receiving and Processing (WRAP) facilities will provide the primary treatment capability for solid-waste storage at the Hanford Site. The Phase V Storage Facility, which accommodates 27,000 drum equivalents of contact-handled waste, will provide the following critical functions for the efficient operation of the WRAP facilities: (1) Shipping/Receiving; (2) Head Space Gas Sampling; (3) Inventory Control; (4) Storage; (5) Automated/Manual Material Handling.

  19. High level radioactive waste glass production and product description

    SciTech Connect

    Sproull, J.F.; Marra, S.L.; Jantzen, C.M.

    1993-12-01

    This report examines borosilicate glass as a means of immobilizing high-level radioactive wastes. Borosilicate glass will encapsulate most of the defense and some of the commercial HLW in the US. The resulting waste forms must meet the requirements of the WA-SRD and the WAPS, which include a short term PCT durability test. The waste form producer must report the composition(s) of the borosilicate waste glass(es) produced but can choose the composition(s) to meet site-specific requirements. Although the waste form composition is the primary determinant of durability, the redox state of the glass; the existence, content, and composition of crystals; and the presence of glass-in-glass phase separation can affect durability. The waste glass should be formulated to avoid phase separation regions. The ultimate result of this effort will be a waste form which is much more stable and potentially less mobile than the liquid high level radioactive waste is currently.

  20. The Use of Induction Melting for the Treatment of Metal Radioactive Waste - 13088

    SciTech Connect

    Zherebtsov, Alexander; Pastushkov, Vladimir; Poluektov, Pavel; Smelova, Tatiana; Shadrin, Andrey

    2013-07-01

    The aim of the work is to assess the efficacy of induction melting metal for recycling radioactive waste in order to reduce the volume of solid radioactive waste to be disposed of, and utilization of the metal. (authors)

  1. Disposal of soluble salt waste from coal gasification

    SciTech Connect

    McKnight, C.E.

    1980-06-01

    This paper addresses pollutants in the form of soluble salts and resource recovery in the form of water and land. A design for disposal of soluble salts has been produced. The interactions of its parameters have been shown by a process design study. The design will enable harmonious compliance with United States Public Laws 92-500 and 94-580, relating to water pollution and resource recovery. In the disposal of waste salt solutions, natural water resources need not be contaminated, because an encapsulation technique is available which will immobilize the salts. At the same time it will make useful landforms available, and water as a resource can be recovered. There is a cost minimum when electrodialysis and evaporation are combined, which is not realizable with evaporation alone, unless very low-cost thermal energy is available or unless very high-cost pretreatment for electrodialysis is required. All the processes making up the proposed disposal process are commercially available, although they are nowhere operating commercially as one process. Because of the commercial availability of the processes, the proposed process may be a candidate 'best commercially available treatment' for soluble salt disposal.

  2. Radioactive and mixed waste - risk as a basis for waste classification. Symposium proceedings No. 2

    SciTech Connect

    1995-06-21

    The management of risks from radioactive and chemical materials has been a major environmental concern in the United states for the past two or three decades. Risk management of these materials encompasses the remediation of past disposal practices as well as development of appropriate strategies and controls for current and future operations. This symposium is concerned primarily with low-level radioactive wastes and mixed wastes. Individual reports were processed separately for the Department of Energy databases.

  3. Method of handling radioactive alkali metal waste

    DOEpatents

    Wolson, Raymond D.; McPheeters, Charles C.

    1980-01-01

    Radioactive alkali metal is mixed with particulate silica in a rotary drum reactor in which the alkali metal is converted to the monoxide during rotation of the reactor to produce particulate silica coated with the alkali metal monoxide suitable as a feed material to make a glass for storing radioactive material. Silica particles, the majority of which pass through a 95 mesh screen or preferably through a 200 mesh screen, are employed in this process, and the preferred weight ratio of silica to alkali metal is 7 to 1 in order to produce a feed material for the final glass product having a silica to alkali metal monoxide ratio of about 5 to 1.

  4. Method of handling radioactive alkali metal waste

    DOEpatents

    Wolson, R.D.; McPheeters, C.C.

    Radioactive alkali metal is mixed with particulate silica in a rotary drum reactor in which the alkali metal is converted to the monoxide during rotation of the reactor to produce particulate silica coated with the alkali metal monoxide suitable as a feed material to make a glass for storing radioactive material. Silica particles, the majority of which pass through a 95 mesh screen or preferably through a 200 mesh screen, are employed in this process, and the preferred weight ratio of silica to alkali metal is 7 to 1 in order to produce a feed material for the final glass product having a silica to alkali metal monoxide ratio of about 5 to 1.

  5. Method of storing radioactive wastes using modified tobermorite

    DOEpatents

    Komarneni, Sridhar; Roy, Della M.

    1985-01-01

    A new cation exchanger is a modified tobermorite containing aluminum isomorphously substituted for silicon and containing sodium or potassium. The exchanger is selective for lead, rubidium, cobalt and cadmium and is selective for cesium over calcium or sodium. The tobermorites are compatable with cement and are useful for the long-term fixation and storage of radioactive nuclear wastes.

  6. Driving Forces and Priorities in the Hungarian Radioactive Waste Management

    SciTech Connect

    Takats, F.; Ormai, P.

    2002-02-26

    Hungary, being a candidate state to the European Union, pays particular attention to the measures that are typically considered as good practice within the EU when developing and implementing its national program for the safe management of spent fuel and radioactive waste. The Public Agency for Radioactive Waste Management (PURAM) has been designated to carry out the multilevel tasks in the field of radioactive waste management. In accordance with changes in infrastructure, Hungary is about to make significant strategic and technical decisions. There are several technical priorities for the coming years, such as improving the existing L/ILW repository, construction of a new repository for L/ILW, extension of the interim storage facility for spent fuel and setting up a revised back-end policy. Preparations for decommissioning of the nuclear facilities have to be developed as well. The paper outlines the main problem areas as well as the approach to managing radioactive wastes. It will be concluded that priorities can be set, but key dates and deadlines will always contain an element of uncertainty due to public and political acceptance problems.

  7. 77 FR 26991 - Low-Level Radioactive Waste Management Issues

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-08

    ... Volume Reduction and Low-Level Radioactive Waste Management'' (76 FR 50500; August 15, 2011); and the... regulations were published in the Federal Register on December 27, 1982 (47 FR 57446). The rule applies to any... (74 FR 30175; Docket ID NRC-2009-0257), the NRC staff subsequently developed a technical...

  8. Ukranian program of radioactive waste disposal in geological formations

    SciTech Connect

    Khrushchov, D.P.; Pavlovsky, M.A.; Starodoumov, V.M.

    1996-12-01

    On the initiative of State Committee on Nuclear Power Utilization the purposeful investigations in the frames of interinstitutional program `isolation of radioactive waste in geologic formations` has been started in 1998. A preparatory stage of R&D program has been completed.

  9. 40 CFR 227.30 - High-level radioactive waste.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false High-level radioactive waste. 227.30 Section 227.30 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING CRITERIA FOR THE EVALUATION OF PERMIT APPLICATIONS FOR OCEAN DUMPING OF MATERIALS Definitions § 227.30...

  10. International Surveillance Mechanism for Sea Dumping of Radioactive Waste

    ERIC Educational Resources Information Center

    OECD Observer, 1977

    1977-01-01

    The OECD consultation and surveillance mechanism is discussed in detail in this article. Four phases are identified and examined: (1) Notification, (2) Consultation, (3) Supervision, (4) Post-operation. This system is designed to provide the safest possible conditions for sea dumping of radioactive wastes. (MA)

  11. Mitigation of plant penetration into radioactive waste utilizing herbicides

    SciTech Connect

    Cox, G.R.

    1982-01-01

    This paper describes the use of herbicides as an effective method of precluding plant root penetration into buried radioactive wastes. The discussed surface applications are selective herbicides to control broadleaf vegetation in grasses; nonselective herbicides, which control all vegetation; and slow-release forms of these herbicides to prolong effectiveness.

  12. Groundwater Impacts of Radioactive Wastes and Associated Environmental Modeling Assessment

    SciTech Connect

    Ma, Rui; Zheng, Chunmiao; Liu, Chongxuan

    2012-11-01

    This article provides a review of the major sources of radioactive wastes and their impacts on groundwater contamination. The review discusses the major biogeochemical processes that control the transport and fate of radionuclide contaminants in groundwater, and describe the evolution of mathematical models designed to simulate and assess the transport and transformation of radionuclides in groundwater.

  13. Radioactive Waste...The Problem and Some Possible Solutions

    ERIC Educational Resources Information Center

    Olivier, Jean-Pierre

    1977-01-01

    Nuclear safety is a highly technical and controversial subject that has caused much heated debate and political concern. This article examines the problems involved in managing radioactive wastes and the techniques now used. Potential solutions are suggested and the need for international cooperation is stressed. (Author/MA)

  14. Disposal of Radioactive Waste at Hanford Creates Problems

    ERIC Educational Resources Information Center

    Chemical and Engineering News, 1978

    1978-01-01

    Radioactive storage tanks at the Hanford facility have developed leaks. The situation is presently considered safe, but serious. A report from the National Academy of Science has recommended that the wastes be converted to stable solids and stored at another site on the Hanford Reservation. (Author/MA)

  15. Ion-exchange material and method of storing radioactive wastes

    DOEpatents

    Komarneni, S.; Roy, D.M.

    1983-10-31

    A new cation exchanger is a modified tobermorite containing aluminum isomorphously substituted for silicon and containing sodium or potassium. The exchanger is selective for lead, rubidium, cobalt, and cadmium and is selective for cesium over calcium or sodium. The tobermorites are compatible with cement and are useful for the long-term fixation and storage of radioactive nuclear wastes.

  16. ANNUAL REPORT. ACTINIDE-ALUMINATE SPECIATION IN ALKALINE RADIOACTIVE WASTE

    EPA Science Inventory

    Highly alkaline radioactive waste tanks contain a number of transuranic species, in particular U, Np, Pu, and Am-the exact forms of which are currently unknown. Knowledge of actinide speciation under highly alkaline conditions is essential towards understanding and predicting the...

  17. Guidelines for generators of hazardous chemical waste at LBL and Guidelines for generators of radioactive and mixed waste at LBL

    SciTech Connect

    Not Available

    1991-07-01

    The purpose of this document is to provide the acceptance criteria for the transfer of hazardous chemical, radioactive, and mixed waste to Lawrence Berkeley Laboratory's (LBL) Hazardous Waste Handling Facility (HWHF). These guidelines describe how a generator of wastes can meet LBL's acceptance criteria for hazardous chemical, radioactive, and mixed waste. 9 figs.

  18. Summary of radioactive solid waste received in the 200 Areas during calendar year 1995

    SciTech Connect

    Hladek, K.L.

    1996-06-06

    Westinghouse Hanford Company manages and operates the Hanford Site 200 Area radioactive solid waste storage and disposal facilities for the US Department of Energy, Richland Operations Office. These facilities include radioactive solid waste disposal sites and radioactive solid waste storage areas. This document summarizes the amount of radioactive materials that have been buried and stored in the 200 Area radioactive solid waste storage and disposal facilities since startup in 1944 through calendar year 1995. This report does not include backlog waste, solid radioactive wastes in storage or disposed of in other areas, or facilities such as the underground tank farms. Unless packaged within the scope of WHC-EP-0063, Hanford Site Solid Waste Acceptance Criteria, liquid waste data are not included in this document. This annual report provides a summary of the radioactive solid waste received in the both the 200-East and 200-West Areas during the calendar year 1995.

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

    SciTech Connect

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

    1992-08-01

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

  20. Method for utilizing decay heat from radioactive nuclear wastes

    DOEpatents

    Busey, H.M.

    1974-10-14

    Management of radioactive heat-producing waste material while safely utilizing the heat thereof is accomplished by encapsulating the wastes after a cooling period, transporting the capsules to a facility including a plurality of vertically disposed storage tubes, lowering the capsules as they arrive at the facility into the storage tubes, cooling the storage tubes by circulating a gas thereover, employing the so heated gas to obtain an economically beneficial result, and continually adding waste capsules to the facility as they arrive thereat over a substantial period of time.

  1. Integrated approach to hazardous and radioactive waste remediation

    SciTech Connect

    Hyde, R.A.; Reece, W.J.

    1994-11-01

    The US Department of Energy Office of Technology Development is supporting the demonstration, and evaluation of a suite of waste retrieval technologies. An integration of leading-edge technologies with commercially available baseline technologies will form a comprehensive system for effective and efficient remediation of buried waste throughout the complex of DOE nuclear facilities. This paper discusses the complexity of systems integration, addressing organizational and engineering aspects of integration as well as the impact of human operators, and the importance of using integrated systems in remediating buried hazardous and radioactive waste.

  2. The problem of burying radioactive wastes containing transplutonium elements (TPE)

    SciTech Connect

    Bryzgalova, R.V.; Krivokhatskii, A.S.; Rogozin, Y.M.; Sinitsyna, G.S.

    1986-09-01

    This paper discusses the problem of burying radioactive wastes containing TPE. The most acceptable and developed method at present is that of disposal into continental, deep-lying, geological formatins. Based on an analysis of estimates of the thermal conditions on burying highly active wastes, including TPE concentrates, data on the filtration and sorption characteristics of rocks, estimates of the diffusion of radionuclide species capable of migrating, and taking into account the retention powers of rocks it is concluded that it is possible to bury such wastes in weakly permeable geological formations possessing shielding characteristics which ensure reliability and safety in burial.

  3. Retrieval & Transfer of Stored Radioactive Waste Content of Process Vessels

    SciTech Connect

    GIBBONS, P.W.

    2002-05-01

    The overall objective of this book is to provide guidance on the retrieval and transfer of stored, bulk radioactive waste in tanks, silos, or similar containment systems. Information is based on the experiences of particular Member States and is intended to provide the people planning retrieval operations with the information they need to develop the most appropriate strategy and supporting processes for their application. It can also provide those in ongoing programs with information to measure their progress and identify additional resources. To this end, a generic methodology for addressing waste retrieval in specific situations is presented with information on the waste retrieval and transportation processes.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. A risk analysis model for radioactive wastes.

    PubMed

    Külahcı, Fatih

    2011-07-15

    Hazardous wastes affect natural environmental systems to a significant extend, and therefore, it is necessary to control their harm through risk analysis. Herein, an effective risk methodology is proposed by considering their uncertain behaviors on stochastic, statistical and probabilistic bases. The basic element is attachment of a convenient probability distribution function (pdf) to a given waste quality measurement sequence. In this paper, (40)K contaminant measurements are adapted for risk assessment application after derivation of necessary fundamental formulations. The spatial contaminant distribution of (40)K is presented in the forms of maps and three-dimensional surfaces. PMID:21571428

  9. Selection of containment systems for commercial high-level radioactive waste management

    SciTech Connect

    Kaplan, M. F.; Giuffre, M. S.; Bartlett, J. W.

    1981-05-01

    This document reports the results of a study aimed at determining the best strategy for providing containment during management of commercial high-level radioactive wastes. Containment to assure public and worker safety is needed for all storage, transport, handling, and disposal operations. There are several thousand containment system options; this work determined, in overview rather than detail, which options should be pursued. This work shows that the geologic and engineered barriers in repositories in different geologic media, such as salt and granite, play very different roles in preserving long-term containment. In sum, there is no common engineered waste package that is suitable for disposal in all geologic media, each package must be tailored to the specific repository system. The need to make the waste package specific to the repository system leads to the key elements of waste management containment strategy: perform final packaging at the disposal site, and deliver to the site a waste that is in a form suitable for disposal and in a container that is (a) appropriate for the process that produced the waste form, (b) satisfactory for transport, and (c) suitable as the common basis for custom tailoring the waste package for any repository system. As described in this report, mild carbon steel is a container material that can be expected to meet these requirements.

  10. Radioactive Waste Management Information for 1991 and Record-to-Date

    SciTech Connect

    Litteer, D.L.; Peterson, C.N.; Sims, A.M.

    1993-04-01

    This document presents detailed data, bar graphs, and pie charts on volume, radioactivity, isotopic identity, origin, and decay status of radioactive waste for the calendar year 1991. It also summarizes the radiative waste data records compiled from 1952 to present for the Idaho National Engineering Laboratory (INEL). The data presented are from the INEL Radioactive Waste Management Information System.

  11. Defense Waste Processing Facility -- Radioactive operations -- Part 3 -- Remote operations

    SciTech Connect

    Barnes, W.M.; Kerley, W.D.; Hughes, P.D.

    1997-06-01

    The Savannah River Site`s Defense Waste Processing Facility (DWPF) near Aiken, South Carolina is the nation`s first and world`s largest vitrification facility. Following a ten year construction period and nearly three years of non-radioactive testing, the DWPF began radioactive operations in March 1996. Radioactive glass is poured from the joule heated melter into the stainless steel canisters. The canisters are then temporarily sealed, decontaminated, resistance welded for final closure, and transported to an interim storage facility. All of these operations are conducted remotely with equipment specially designed for these processes. This paper reviews canister processing during the first nine months of radioactive operations at DWPF. The fundamental design consideration for DWPF remote canister processing and handling equipment are discussed as well as interim canister storage.

  12. Applications Where Snap is BPM for Radioactive Waste Assay

    SciTech Connect

    Miller, T.J.

    2008-07-01

    Historically, the Atomic Weapons Establishment (AWE) at Aldermaston in the United Kingdom (UK), has used a variety of assay techniques to measure the radioactive content of a diverse range of waste packages from decommissioning, operational and legacy sources. The regulator, the Environment Agency in the UK, places conditions and limits on AWE through an authorisation within the Radioactive Substances Act (RSA93). The conditions and limits require Best Practical Means (BPM) measurements to be used to demonstrate compliance with the authorisation. Hence, the assay technique employed needs to achieve a balance between risk of exposure, environmental considerations, technological considerations, health and safety considerations and cost effectiveness, without being grossly disproportionate in terms of money, time or trouble. Recently published work has concluded that the Spectral Non-destructive Assay Platform (SNAP) assay system is BPM for Depleted Uranium (DU) waste assay at AWE (1) and low level plutonium in soft drummed waste, HEPA filters and soils (2-4). The purpose of this paper is to highlight other applications where SNAP represents BPM for radioactive waste assay. This has been done by intercomparison studies of SNAP with other assay techniques, such as Segmented Gamma Scanner (SGS) and Passive Neutron Coincidence Counter (PNCC). It has been concluded that, for a large range of waste packages encountered at AWE, SNAP is BPM. (author)

  13. Radioactive waste management treatments: A selection for the Italian scenario

    SciTech Connect

    Locatelli, G.; Mancini, M.

    2012-07-01

    The increased attention for radioactive waste management is one of the most peculiar aspects of the nuclear sector considering both reactors and not power sources. The aim of this paper is to present the state-of-art of treatments for radioactive waste management all over the world in order to derive guidelines for the radioactive waste management in the Italian scenario. Starting with an overview on the international situation, it analyses the different sources, amounts, treatments, social and economic impacts looking at countries with different industrial backgrounds, energetic policies, geography and population. It lists all these treatments and selects the most reasonable according to technical, economic and social criteria. In particular, a double scenario is discussed (to be considered in case of few quantities of nuclear waste): the use of regional, centralized, off site processing facilities, which accept waste from many nuclear plants, and the use of mobile systems, which can be transported among multiple nuclear sites for processing campaigns. At the end the treatments suitable for the Italian scenario are presented providing simplified work-flows and guidelines. (authors)

  14. Development of radioactive waste management licensing review assistant

    SciTech Connect

    Wei-Whua Loa; Suan Chen; Wei-Chu Yu

    1992-12-31

    Regulations on radioactive waste disposal are now in urgent need due to our increasing consumption of electric power from nuclear origin. It is set forth that actually applying the regulations to evaluate the license application of new repositories for interim storage and final disposal of High-Level Waste and Low-Level Waste before the year of 2000. In the mean time, it is expected to establish the basis for the decision on issuing the license. The license review procedure can be very complicated, because too many factors must be taken into consideration. For the time being, licensing review is as much an art as it is a science. The authority usually faces three major problems; (1) the availability of domain expert, (2) maintaining of high quality and consistent reviews, and (3) the documentation of the review process. However, to maintain a more efficient, accurate, and systematic review procedure, and at the same time to reduce costs, the Artificial Intelligence (AI) techniques may be used. An expert system is designed as a radioactive waste management licensing review aid for the staff those are in charge of the license application. Tasks such as completeness checking, functional areas of review distribution, participation confirmation, knowledge acquisition, review comment collection, weighting calculation, and degree of satisfaction are considered. In this paper we will discuss the development of the radioactive waste management licensing review assistant.

  15. Stress Corrosion Cracking Model for High Level Radioactive-Waste Packages

    SciTech Connect

    P. Andresen; G. Gordon; S. Lu

    2004-10-05

    A stress corrosion cracking (SCC) model has been adapted for performance prediction of high level radioactive-waste packages to be emplaced in the proposed Yucca Mountain repository. For waste packages of the proposed Yucca Mountain repository, the outer barrier material is the highly corrosion-resistant Alloy UNS-N06022 (Alloy 22), the environment is represented by aqueous brine films present on the surface of the waste package from dripping or deliquescence of soluble salts present in any surface deposits, and the tensile stress is principally from weld induced residual stress. SCC has historically been separated into ''initiation'' and ''propagation'' phases. Initiation of SCC will not occur on a smooth surface if the surface stress is below a threshold value defined as the threshold stress. Cracks can also initiate at and propagate from flaws (or defects) resulting from manufacturing processes (such as welding); or that develop from corrosion processes such as pitting or dissolution of inclusions. To account for crack propagation, the slip dissolution/film rupture (SDFR) model is adopted to provide mathematical formulae for prediction of the crack growth rate. Once the crack growth rate at an initiated SCC is determined, it can be used by the performance assessment to determine the time to through-wall penetration for the waste package. This paper presents the development of the SDFR crack growth rate model based on technical information in the literature as well as experimentally determined crack growth rates developed specifically for Alloy UNS-N06022 in environments relevant to high level radioactive-waste packages of the proposed Yucca Mountain radioactive-waste repository. In addition, a seismic damage related SCC crack opening area density model is briefly described.

  16. ISOLATION OF RADIOACTIVE METALS FROM LIQUID WASTES

    EPA Science Inventory

    Metals are present in many waste streams, and pose challenges with regard to their disposal. Release of metals into the environment presents both human health and ecological concerns. As a result, efforts are directed at reducing their toxicity, bioavailability, and environment...

  17. Iraq liquid radioactive waste tanks maintenance and monitoring program plan.

    SciTech Connect

    Dennis, Matthew L.; Cochran, John Russell; Sol Shamsaldin, Emad

    2011-10-01

    The purpose of this report is to develop a project management plan for maintaining and monitoring liquid radioactive waste tanks at Iraq's Al-Tuwaitha Nuclear Research Center. Based on information from several sources, the Al-Tuwaitha site has approximately 30 waste tanks that contain varying amounts of liquid or sludge radioactive waste. All of the tanks have been non-operational for over 20 years and most have limited characterization. The program plan embodied in this document provides guidance on conducting radiological surveys, posting radiation control areas and controlling access, performing tank hazard assessments to remove debris and gain access, and conducting routine tank inspections. This program plan provides general advice on how to sample and characterize tank contents, and how to prioritize tanks for soil sampling and borehole monitoring.

  18. Corrosion of radioactive waste tanks containing washed sludge and precipitates

    SciTech Connect

    Bickford, D.F.; Congdon, J.W.; Oblath, S.B.

    1988-05-01

    At the US Department of Energy (DOE) Savannah River Plant, the corrosion of carbon steel storage tanks containing alkaline, high-level radioactive waste is controlled by specification of limits on waste composition and temperature. Laboratory tests, conducted to determine minimum corrosion inhibitor levels, indicated pitting of carbon steel near the waterline for proposed storage conditions. In situ electrochemical measurements of full-scale radioactive process demonstrations were conducted to assess the validity of laboratory tests. The in situ results are compared to those of laboratory tests, with particular regard given to simulated solution composition. Transition metal hydroxide sludge contains strong passivating species for carbon steel. Washed precipitate contains organic species that lower solution pH and tend to reduce passivating films, requiring higher inhibitor concentrations than the 0.01 M nitrite required for reactor fuel reprocessing wastes.

  19. Summary of national and international fuel cycle and radioactive waste management programs, 1984

    SciTech Connect

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

    1984-07-01

    Worldwide activities related to nuclear fuel cycle and radioactive waste management programs are summarized. Several trends have developed in waste management strategy: All countries having to dispose of reprocessing wastes plan on conversion of the high-level waste (HLW) stream to a borosilicate glass and eventual emplacement of the glass logs, suitably packaged, in a deep geologic repository. Countries that must deal with plutonium-contaminated waste emphasize pluonium recovery, volume reduction and fixation in cement or bitumen in their treatment plans and expect to use deep geologic repositories for final disposal. Commercially available, classical engineering processing are being used worldwide to treat and immobilize low- and intermediate-level wastes (LLW, ILW); disposal to surface structures, shallow-land burial and deep-underground repositories, such as played-out mines, is being done widely with no obvious technical problems. Many countries have established extensive programs to prepare for construction and operation of geologic repositories. Geologic media being studied fall into three main classes: argillites (clay or shale); crystalline rock (granite, basalt, gneiss or gabbro); and evaporates (salt formations). Most nations plan to allow 30 years or longer between discharge of fuel from the reactor and emplacement of HLW or spent fuel is a repository to permit thermal and radioactive decay. Most repository designs are based on the mined-gallery concept, placing waste or spent fuel packages into shallow holes in the floor of the gallery. Many countries have established extensive and costly programs of site evaluation, repository development and safety assessment. Two other waste management problems are the subject of major R and D programs in several countries: stabilization of uranium mill tailing piles; and immobilization or disposal of contaminated nuclear facilities, namely reactors, fuel cycle plants and R and D laboratories.

  20. First use of in situ vitrification on radioactive wastes

    SciTech Connect

    Bowlds, L.

    1992-03-01

    A high-temperature method for containing hazardous wastes, which was first developed in the 1980s, is being adapted for the in situ treatment of buried radioactive wastes by the US DOE's Idaho National Engineering Laboratory (INEL), following its recent report on successful preliminary tests. The method, called in situ vitrification (ISV), is an electrically induced thermal process that melts and fuses soil and wastes into a glass-like material at least as strong as natural obsidian or granite. Gases released during the process are captured and treated by an off-gas treatment system. After the wastes are vitrified, they could be left in place, or the mass could be broken up and transported to a disposal site. The glass-like substance would be chemically and physically similar to obsidian and from 4 to 10 times more durable than typical borosilicate glasses used to immobolize high-level nuclear wastes.

  1. Characteristics of low-level radioactive decontamination waste

    SciTech Connect

    Akers, D.W.; McConnell, J.W. Jr.; Morcos, N. )

    1993-02-01

    This document addresses the work performed during fiscal year 1992 at the Idaho National Engineering Laboratory by the Low-Level Radioactive Waste -- Decontamination Waste Program (FIN A6359), which is funded by the US Nuclear Regulatory Commission. The program evaluates the physical stability and leachability of solidified waste streams generated in the decontamination process of primary coolant systems in operating nuclear power stations. The data in this document include the chemical composition and characterization of waste streams from Peach Bottom Atomic Power Station Unit 3 and from Nine Mile Point Nuclear Plant Unit 1. The results of compressive strength testing on immersed and unimmersed solidified waste-form specimens from peach Bottom, and the results of leachate analysis are addressed. Cumulative fractional release rates and leachability indexes of those specimens were calculated and are included in this report.

  2. Building 251 Radioactive Waste Characterization by Process Knowledge

    SciTech Connect

    Dominick, J L

    2002-05-29

    Building 251 is the Lawrence Livermore National Laboratory Heavy Elements Facility. Operations that involved heavy elements with uncontained radioisotopes including transuranic elements took place inside of glove boxes and fume hoods. These operations included process and solution chemistry, dissolutions, titrations, centrifuging, etc., and isotope separation. Operations with radioactive material which presently take place outside of glove boxes include storage, assaying, packing and unpacking and inventory verification. Wastes generated inside glove boxes will generally be considered TRU or Greater Than Class C (GTCC). Wastes generated in the RMA, outside glove boxes, is presumed to be low level waste. This process knowledge quantification method may be applied to waste generated anywhere within or around B251. The method is suitable only for quantification of waste which measures below the MDA of the Blue Alpha meter (i.e. only material which measures as Non-Detect with the blue alpha is to be characterized by this method).

  3. RETENTION OF SULFATE IN HIGH LEVEL RADIOACTIVE WASTE GLASS

    SciTech Connect

    Fox, K.

    2010-09-07

    High level radioactive wastes are being vitrified at the Savannah River Site for long term disposal. Many of the wastes contain sulfate at concentrations that can be difficult to retain in borosilicate glass. This study involves efforts to optimize the composition of a glass frit for combination with the waste to improve sulfate retention while meeting other process and product performance constraints. The fabrication and characterization of several series of simulated waste glasses are described. The experiments are detailed chronologically, to provide insight into part of the engineering studies used in developing frit compositions for an operating high level waste vitrification facility. The results lead to the recommendation of a specific frit composition and a concentration limit for sulfate in the glass for the next batch of sludge to be processed at Savannah River.

  4. Radioactive waste reality as revealed by neutron measurements

    SciTech Connect

    Schultz, F.J.

    1995-12-31

    To comprehend certain aspects of the contents of a radioactive waste container is not a trivial matter, especially if one is not allowed to open the container and peer inside. One of the suite of tools available to a practioner in the art of nondestructive assay is based upon neutron measurements. Neutrons, both naturally occuring and induced, are penertrating radiations that can be detected external to the waste container. The practioner should be skilled in applying the proper technique(s) to selected waste types. Available techniques include active and passive neutron measurements, each with their own strengths and weaknesses. The waste material itself can compromise the assay results by occluding a portion of the mass of fissile material present, or by multiplying the number of neutrons produced by a spontaneously fissioning mass. This paper will discuss the difficult, but albeit necessary marriage, between radiioactive waste types and alternative neutron measurement techniques.

  5. Incineration of Low Level Radioactive Vegetation for Waste Volume Reduction

    SciTech Connect

    Malik, N.P.S.; Rucker, G.G.; Looper, M.G.

    1995-03-01

    The DOE changing mission at Savannah River Site (SRS) are to increase activities for Waste Management and Environmental Restoration. There are a number of Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) locations that are contaminated with radioactivity and support dense vegetation, and are targeted for remediation. Two such locations have been studied for non-time critical removal actions under the National Contingency Plan (NCP). Both of these sites support about 23 plant species. Surveys of the vegetation show that radiation emanates mainly from vines, shrubs, and trees and range from 20,000 to 200,000 d/m beta gamma. Planning for removal and disposal of low-level radioactive vegetation was done with two principal goals: to process contaminated vegetation for optimum volume reduction and waste minimization, and for the protection of human health and environment. Four alternatives were identified as candidates for vegetation removal and disposal: chipping the vegetation and packing in carbon steel boxes (lined with synthetic commercial liners) and disposal at the Solid Waste Disposal Facility at SRS; composting the vegetation; burning the vegetation in the field; and incinerating the vegetation. One alternative `incineration` was considered viable choice for waste minimization, safe handling, and the protection of the environment and human health. Advantages and disadvantages of all four alternatives considered have been evaluated. For waste minimization and ultimate disposal of radioactive vegetation incineration is the preferred option. Advantages of incineration are that volume reduction is achieved and low-level radioactive waste are stabilized. For incineration and final disposal vegetation will be chipped and packed in card board boxes and discharged to the rotary kiln of the incinerator. The slow rotation and longer resident time in the kiln will ensure complete combustion of the vegetative material.

  6. Drying radioactive wastewater salts using a thin film dryer

    SciTech Connect

    Scully, D.E.

    1998-03-19

    This paper describes the operational experience in drying brines generated at a radioactive wastewater treatment facility. The brines are composed of aqueous ammonium sulfate/sodium sulfate and aqueous sodium nitrate/sodium sulfate, The brine feeds receive pretreatment to preclude dryer bridging and fouling. The dryer products are a distillate and a powder. The dryer is a vertical thin film type consisting of a steam heated cylinder with rotor. Maintenance on the dryer has been minimal. Although many operability problems have had to be overcome, dryer performance can now be said to be highly reliable.

  7. Systematic approach to radioactive waste characterization at Belgoprocess

    SciTech Connect

    Huys, T.; Gielen, P.

    2007-07-01

    Belgoprocess is capable of processing almost every type of low and medium level radioactive waste and thereby covering a large segment from the back-end of the nuclear fuel cycle. Waste from numerous producers is treated and conditioned into a stable end product. Such processes lead inevitably to the generation of a large number of different waste streams. Each of these streams is uniquely defined by its radiological and physicochemical characteristics. From regulatory point of view and in order to select appropriate processing and conditioning techniques it is essential to characterize each of these waste streams. Because of the labour-intensive nature of the work and to keep a trustworthy traceability, Belgoprocess has decided to automate this task as far as possible. Therefore it has developed a system that seamlessly integrates waste-accounting and radiological characterization into one system. The use of generic methodologies, isotope vectors and a measurement database makes it possible to characterize most waste packages without elaborate knowledge of radiological characterization. A nuclear engineer develops generic methodologies and defines isotope vectors and appropriate measurements. These combinations are documented in procedures and used by the waste-accounting team to characterize the waste packages. The whole system is designed and programmed in such a way that it offers maximum flexibility and traceability. For example, changes in characterization of the previously processed and conditioned waste will propagate through the system until the changes reach the end product. This kind of systematic approach to radioactive waste characterization is found to be very fruitful. (authors)

  8. Vitrification of an incinerator blowdown waste containing both chloride salts and carbon

    SciTech Connect

    Resce, J.L.; Ragsdale, R. G.; Overcamp, T.J.

    1996-10-01

    A low-level, mixed-waste simulant, derived from incinerator blowdown residue, has been vitrified in a series of crucible tests. Major components of this waste simulant included carbon, sodium chloride, silica, and alumina. Hazardous and surrogate radioactive metal complexes were also included. Two different formulations of additives were combined with the waste to facilitate glass formation. These glass formulations included NaNO{sub 3} to serve as a flux and to assist in the oxidation of the carbon. During vitrification, a chloride salt layer was observed on the melt surface which volatilized during the course of heating. Furthermore, significant concentrations of As, Cd, Cs, Ni, Pb, Sb, Se, Tl, and Zn had volatilized during vitrification. It is postulated that this was due, in part, to the formation of volatile metal chlorides. Upon quenching, small metal nuggets, containing Pb, Sb, Ag, and Ni separated from the glasses. This reduction of the metal oxides was caused by the presence of carbon in the simulant. The chemical durability was evaluated by both the Toxicity Characteristic Leaching Procedure (TCLP) and the Product Consistency Test (PCT). The TCLP results were below the limits for the toxicity characteristic for a hazardous waste and also met the Universal Treatment Standards for listed wastes such as F006. 11 refs., 5 tabs.

  9. Properties of radioactive wastes and waste containers. [Marlex CL-100

    SciTech Connect

    Arora, H.S.; Dayal, R.

    1984-01-01

    Major tasks in this NRC-sponsored program include: (1) an evaluation of the acceptability of low-level solidified wastes with respect to minimizing radionuclide releases after burial; and (2) an assessment of the influence of pertinent environmental stresses on the performance of high-integrity radwaste container (HIC) materials. The waste form performance task involves studies on small-scale laboratory specimens to predict and extrapolate: (1) leachability for extended time periods; (2) leach behavior of full-size forms; (3) performance of waste forms under realistic leaching conditions; and (4) leachability of solidified reactor wastes. The results show that leach data derived from testing of small-scale specimens can be extrapolated to estimate leachability of a full-scale specimen and that radionuclide release data derived from testing of simulants can be employed to predict the release behavior of reactor wastes. Leaching under partially saturated conditions exhibits lower releases of radionuclides than those observed under the conventional IAEA-type or ANS 16.1 leach tests. The HIC assessment task includes the characterization of mechanical properties of Marlex CL-100, a candidate radwaste high density polyethylene material. Tensile strength and creep rupture tests have been carried out to determine the influence of specific waste constituents as well as gamma irradiation on material performance. Emphasis in ongoing tests is being placed on studying creep rupture while the specimens are in contact with a variety of chemicals including radiolytic by-products of irradiated resin wastes. 12 references 6 figures, 2 tables.

  10. Pyrolytic conversion of plastic and rubber waste to hydrocarbons with basic salt catalysts

    DOEpatents

    Wingfield, Jr., Robert C.; Braslaw, Jacob; Gealer, Roy L.

    1985-01-01

    The invention relates to a process for improving the pyrolytic conversion of waste selected from rubber and plastic to low molecular weight olefinic materials by employing basis salt catalysts in the waste mixture. The salts comprise alkali or alkaline earth compounds, particularly sodium carbonate, in an amount of greater than about 1 weight percent based on the waste feed.

  11. Resistance of class C fly ash belite cement to simulated sodium sulphate radioactive liquid waste attack.

    PubMed

    Guerrero, A; Goñi, S; Allegro, V R

    2009-01-30

    The resistance of class C fly ash belite cement (FABC-2-W) to concentrated sodium sulphate salts associated with low level wastes (LLW) and medium level wastes (MLW) is discussed. This study was carried out according to the Koch and Steinegger methodology by testing the flexural strength of mortars immersed in simulated radioactive liquid waste rich in sulphate (48,000 ppm) and demineralised water (used as a reference), at 20 degrees C and 40 degrees C over a period of 180 days. The reaction mechanisms of sulphate ion with the mortar was carried out through a microstructure study, which included the use of Scanning electron microscopy (SEM), porosity and pore-size distribution and X-ray diffraction (XRD). The results showed that the FABC mortar was stable against simulated sulphate radioactive liquid waste (SSRLW) attack at the two chosen temperatures. The enhancement of mechanical properties was a result of the formation of non-expansive ettringite inside the pores and an alkaline activation of the hydraulic activity of cement promoted by the ingress of sulphate. Accordingly, the microstructure was strongly refined. PMID:18524482

  12. Permeability of natural rock salt from the Waste Isolation Pilot Plant (WIPP) during damage evolution and healing

    SciTech Connect

    Pfeifle, T.W.

    1998-06-01

    The US Department of Energy has developed the Waste Isolation Pilot Plant (WIPP) in the bedded salt of southeastern New Mexico to demonstrate the safe disposal of radioactive transuranic wastes. Four vertical shafts provide access to the underground workings located at a depth of about 660 meters. These shafts connect the underground facility to the surface and potentially provide communication between lithologic units, so they will be sealed to limit both the release of hazardous waste from and fluid flow into the repository. The seal design must consider the potential for fluid flow through a disturbed rock zone (DRZ) that develops in the salt near the shafts. The DRZ, which forms initially during excavation and then evolves with time, is expected to have higher permeability than the native salt. The closure of the shaft openings (i.e., through salt creep) will compress the seals, thereby inducing a compressive back-stress on the DRZ. This back-stress is expected to arrest the evolution of the DRZ, and with time will promote healing of damage. This paper presents laboratory data from tertiary creep and hydrostatic compression tests designed to characterize damage evolution and healing in WIPP salt. Healing is quantified in terms of permanent reduction in permeability, and the data are used to estimate healing times based on considerations of first-order kinetics.

  13. Gas generation phenomena in radioactive waste transportation packaging

    SciTech Connect

    Nigrey, P.J.

    1997-11-01

    The interaction of radiation from radioactive materials with the waste matrix can lead to the deterioration of the waste form resulting in the possible formation of gaseous species. Depending on the type and characteristics of the radiation source, the generation of hydrogen may predominate. Since the interaction of alpha particles with the waste form results in significant energy transfer, other gases such as carbon oxides, methane, nitrogen oxides, oxygen, water, and helium are possible. The type of gases produced from the waste forms is determined by the mechanisms involved in the waste degradation. For transuranic wastes, the identified degradation mechanisms are reported to be caused by radiolysis, thermal decomposition or dewatering, chemical corrosion, and bacterial action. While all these mechanisms may be responsible for the buildup of gases during the storage of wastes, radiolysis and thermal decomposition appear to be the main contributors during waste transport operations. In this paper, the authors provide a review of applicable gas generation data resulting from the degradation of various waste forms under conditions typical for transport. The effects of radiolytic and thermal degradation mechanisms will be discussed in the context of transportation safety.

  14. Lessons Learned from Radioactive Waste Storage and Disposal Facilities

    SciTech Connect

    Esh, David W.; Bradford, Anna H.

    2008-01-15

    The safety of radioactive waste disposal facilities and the decommissioning of complex sites may be predicated on the performance of engineered and natural barriers. For assessing the safety of a waste disposal facility or a decommissioned site, a performance assessment or similar analysis is often completed. The analysis is typically based on a site conceptual model that is developed from site characterization information, observations, and, in many cases, expert judgment. Because waste disposal facilities are sited, constructed, monitored, and maintained, a fair amount of data has been generated at a variety of sites in a variety of natural systems. This paper provides select examples of lessons learned from the observations developed from the monitoring of various radioactive waste facilities (storage and disposal), and discusses the implications for modeling of future waste disposal facilities that are yet to be constructed or for the development of dose assessments for the release of decommissioning sites. Monitoring has been and continues to be performed at a variety of different facilities for the disposal of radioactive waste. These include facilities for the disposal of commercial low-level waste (LLW), reprocessing wastes, and uranium mill tailings. Many of the lessons learned and problems encountered provide a unique opportunity to improve future designs of waste disposal facilities, to improve dose modeling for decommissioning sites, and to be proactive in identifying future problems. Typically, an initial conceptual model was developed and the siting and design of the disposal facility was based on the conceptual model. After facility construction and operation, monitoring data was collected and evaluated. In many cases the monitoring data did not comport with the original site conceptual model, leading to additional investigation and changes to the site conceptual model and modifications to the design of the facility. The following cases are discussed

  15. Anaerobic microbial transformations of radioactive wastes in subsurface environments

    SciTech Connect

    Francis, A.J.

    1984-01-01

    Radioactive wastes disposed of in subsurface environments contain a variety of radionuclides and organic compounds. Microorganisms play a major role in the transformation of organic and inorganic constituents of the waste and are partly responsible for the problems encountered at the waste disposal sites. These include microbial degradation of waste forms resulting in trench cover subsidence, migration of radionuclides, and production of radioactive gases such as /sup 14/CO/sub 2/, /sup 14/CH/sub 4/, HT, and CH/sub 3/T. Microbial processes involved in solubilization, mobilization, and immobilization of toxic metals under aerobic and anaerobic conditions are reviewed. Complexing agents and several organic acids produced by microbial action affect mobilization of radionuclides and heavy metals from the wastes. Microorganisms play a significant role in the transformation and cycling of tritium in the environment by (i) oxidation of tritium and tritiated methane under aerobic conditions and (ii) production of tritium and tritiated methane from wastes containing tritiated water and organic compounds under anaerobic conditions. 23 references, 2 figures, 2 tables.

  16. NRC Monitoring of Salt Waste Disposal at the Savannah River Site - 13147

    SciTech Connect

    Pinkston, Karen E.; Ridge, A. Christianne; Alexander, George W.; Barr, Cynthia S.; Devaser, Nishka J.; Felsher, Harry D.

    2013-07-01

    As part of monitoring required under Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA), the NRC staff reviewed an updated DOE performance assessment (PA) for salt waste disposal at the Saltstone Disposal Facility (SDF). The NRC staff concluded that it has reasonable assurance that waste disposal at the SDF meets the 10 CFR 61 performance objectives for protection of individuals against intrusion (chap.61.42), protection of individuals during operations (chap.61.43), and site stability (chap.61.44). However, based on its evaluation of DOE's results and independent sensitivity analyses conducted with DOE's models, the NRC staff concluded that it did not have reasonable assurance that DOE's disposal activities at the SDF meet the performance objective for protection of the general population from releases of radioactivity (chap.61.41) evaluated at a dose limit of 0.25 mSv/yr (25 mrem/yr) total effective dose equivalent (TEDE). NRC staff also concluded that the potential dose to a member of the public is expected to be limited (i.e., is expected to be similar to or less than the public dose limit in chap.20.1301 of 1 mSv/yr [100 mrem/yr] TEDE) and is expected to occur many years after site closure. The NRC staff used risk insights gained from review of the SDF PA, its experience monitoring DOE disposal actions at the SDF over the last 5 years, as well as independent analysis and modeling to identify factors that are important to assessing whether DOE's disposal actions meet the performance objectives. Many of these factors are similar to factors identified in the NRC staff's 2005 review of salt waste disposal at the SDF. Key areas of interest continue to be waste form and disposal unit degradation, the effectiveness of infiltration and erosion controls, and estimation of the radiological inventory. Based on these factors, NRC is revising its plan for monitoring salt waste disposal at the SDF in coordination with South

  17. Ceramic waste form for residues from molten salt oxidation of mixed wastes

    SciTech Connect

    Van Konynenburg, R.A.; Hopper, R.W.; Rard, J.A.

    1995-11-01

    A ceramic waste form based on Synroc-D is under development for the incorporation of the mineral residues from molten salt oxidation treatment of mixed low-level wastes. Samples containing as many as 32 chemical elements have been fabricated, characterized, and leach-tested. Universal Treatment Standards have been satisfied for all regulated elements except and two (lead and vanadium). Efforts are underway to further improve chemical durability.

  18. A Stochastic Problem Arising in the Storage of Radioactive Waste

    SciTech Connect

    Williams, M.M.R.

    2004-07-15

    Nuclear waste drums can contain a collection of radioactive components of uncertain activity and randomly dispersed in position. This implies that the dose-rate at the surface of different drums in a large assembly of similar drums can have significant variations according to the physical makeup and configuration of the waste components. The present paper addresses this problem by treating the drum, and its waste, as a stochastic medium. It is assumed that the sources in the drum contribute a dose-rate to some external point. The strengths and positions are chosen by random numbers, the dose-rate is calculated and, from several thousand realizations, a probability distribution for the dose-rate is obtained. It is shown that a very close approximation to the dose-rate probability function is the log-normal distribution. This allows some useful statistical indicators, which are of environmental importance, to be calculated with little effort.As an example of a practical situation met in the storage of radioactive waste containers, we study the problem of 'hotspots'. These arise in drums in which most of the activity is concentrated on one radioactive component and hence can lead to the possibility of large surface dose-rates. It is shown how the dose-rate, the variance, and some other statistical indicators depend on the relative activities on the sources. The results highlight the importance of such hotspots and the need to quantify their effect.

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

    SciTech Connect

    Choi, K.C.; Ahn, J.H.; Park, Y.J.; Song, K.S.

    2013-07-01

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

  20. Geological problems in radioactive waste isolation - second worldwide review

    SciTech Connect

    Witherspoon, P.A.

    1996-09-01

    The first world wide review of the geological problems in radioactive waste isolation was published by Lawrence Berkeley National Laboratory in 1991. This review was a compilation of reports that had been submitted to a workshop held in conjunction with the 28th International Geological Congress that took place July 9-19, 1989 in Washington, D.C. Reports from 15 countries were presented at the workshop and four countries provided reports after the workshop, so that material from 19 different countries was included in the first review. It was apparent from the widespread interest in this first review that the problem of providing a permanent and reliable method of isolating radioactive waste from the biosphere is a topic of great concern among the more advanced, as well as the developing, nations of the world. This is especially the case in connection with high-level waste (HLW) after its removal from nuclear power plants. The general concensus is that an adequate isolation can be accomplished by selecting an appropriate geologic setting and carefully designing the underground system with its engineered barriers. This document contains the Second Worldwide Review of Geological Problems in Radioactive Waste Isolation, dated September 1996.

  1. A robotic inspector for low-level radioactive waste

    SciTech Connect

    Byrd, J.S.; Pettus, R.O.

    1996-06-01

    The Department of Energy has low-level radioactive waste stored in warehouses at several facilities. Weekly visual inspections are required. A mobile robot inspection system, ARIES (Autonomous Robotic Inspection Experimental System), has been developed to survey and inspect the stored drums. The robot will travel through the three- foot wide aisles of drums stacked four high and perform a visual inspection, normally performed by a human operator, making decisions about the condition of the drums and maintaining a database of pertinent information about each drum. This mobile robot system will improve the quality of inspection, generate required reports, and relieve human operators from low-level radioactive exposure.

  2. Three multimedia models used at hazardous and radioactive waste sites

    SciTech Connect

    1996-01-01

    The report provides an approach for evaluating and critically reviewing the capabilities of multimedia models. The study focused on three specific models: MEPAS version 3.0, MMSOILS Version 2.2, and PRESTO-EPA-CPG Version 2.0. The approach to model review advocated in the study is directed to technical staff responsible for identifying, selecting and applying multimedia models for use at sites containing radioactive and hazardous materials. In the report, restrictions associated with the selection and application of multimedia models for sites contaminated with radioactive and mixed wastes are highlighted.

  3. Guidelines for generators of hazardous chemical waste at LBL and guidelines for generators of radioactive and mixed waste at LBL

    SciTech Connect

    Not Available

    1991-09-01

    In part one of this document the Governing Documents and Definitions sections provide general guidelines and regulations applying to the handling of hazardous chemical wastes. The remaining sections provide details on how you can prepare your waste properly for transport and disposal. They are correlated with the steps you must take to properly prepare your waste for pickup. The purpose of the second part of this document is to provide the acceptance criteria for the transfer of radioactive and mixed waste to LBL's Hazardous Waste Handling Facility (HWHF). These guidelines describe how you, as a generator of radioactive or mixed waste, can meet LBL's acceptance criteria for radioactive and mixed waste.

  4. Improving radioactive waste management: an overview of the Environmental Protection Agency's low-activity waste effort.

    PubMed

    Schultheisz, Daniel J; Czyscinski, Kenneth S; Klinger, Adam D

    2006-11-01

    Radioactive waste disposal in the United States is marked by a fragmented regulatory system, with requirements that often focus on the origin or statutory definition of the waste, rather than the hazard of the material in question. It may be possible to enhance public protection by moving toward a system that provides disposal options appropriate for the hazard presented by the waste in question. This paper summarizes aspects of an approach focusing on the potential use, with appropriate conditions, of Resource Conservation and Recovery Act Subtitle-C hazardous waste landfills for disposal of "low-activity" wastes and public comments on the suggested approach. PMID:17033466

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

    SciTech Connect

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

    1986-01-01

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

  6. Stabilization Using Phosphate Bonded Ceramics. Salt Containing Mixed Waste Treatment. Mixed Waste Focus Area. OST Reference No. 117

    SciTech Connect

    None, None

    1999-09-01

    Throughout the Department of Energy (DOE) complex there are large inventories of homogeneous mixed waste solids, such as wastewater treatment residues, fly ashes, and sludges that contain relatively high concentrations (greater than 15% by weight) of salts. The inherent solubility of salts (e.g., nitrates, chlorides, and sulfates) makes traditional treatment of these waste streams difficult, expensive, and challenging. One alternative is low-temperature stabilization by chemically bonded phosphate ceramics (CBPCs). The process involves reacting magnesium oxide with monopotassium phosphate with the salt waste to produce a dense monolith. The ceramic makes a strong environmental barrier, and the metals are converted to insoluble, low-leaching phosphate salts. The process has been tested on a variety of surrogates and actual mixed waste streams, including soils, wastewater, flyashes, and crushed debris. It has also been demonstrated at scales ranging from 5 to 55 gallons. In some applications, the CBPC technology provides higher waste loadings and a more durable salt waste form than the baseline method of cementitious grouting. Waste form test specimens were subjected to a variety of performance tests. Results of waste form performance testing concluded that CBPC forms made with salt wastes meet or exceed both RCRA and recommended Nuclear Regulatory Commission (NRC) low-level waste (LLW) disposal criteria. Application of a polymer coating to the CBPC may decrease the leaching of salt anions, but continued waste form evaluations are needed to fully assess the deteriorating effects of this leaching, if any, over time.

  7. Confinement matrices for low- and intermediate-level radioactive waste

    NASA Astrophysics Data System (ADS)

    Laverov, N. P.; Omel'Yanenko, B. I.; Yudintsev, S. V.; Stefanovsky, S. V.

    2012-02-01

    Mining of uranium for nuclear fuel production inevitably leads to the exhaustion of natural uranium resources and an increase in market price of uranium. As an alternative, it is possible to provide nuclear power plants with reprocessed spent nuclear fuel (SNF), which retains 90% of its energy resource. The main obstacle to this solution is related to the formation in the course of the reprocessing of SNF of a large volume of liquid waste, and the necessity to concentrate, solidify, and dispose of this waste. Radioactive waste is classified into three categories: low-, intermediate-, and high-level (LLW, ILW, and HLW); 95, 4.4, and 0.6% of the total waste are LLW, ILW, and HLW, respectively. Despite its small relative volume, the radioactivity of HLW is approximately equal to the combined radioactivity of LLW + ILW (LILW). The main hazard of HLW is related to its extremely high radioactivity, the occurrence of long-living radionuclides, heat release, and the necessity to confine HLW for an effectively unlimited time period. The problems of handling LILW are caused by the enormous volume of such waste. The available technology for LILW confinement is considered, and conclusion is drawn that its concentration, vitrification, and disposal in shallow-seated repositories is a necessary condition of large-scale reprocessing of SNF derived from VVER-1000 reactors. The significantly reduced volume of the vitrified LILW and its very low dissolution rate at low temperatures makes borosilicate glass an ideal confinement matrix for immobilization of LILW. At the same time, the high corrosion rate of the glass matrix at elevated temperatures casts doubt on its efficient use for immobilization of heat-releasing HLW. The higher cost of LILW vitrification compared to cementation and bitumen impregnation is compensated for by reduced expenditure for construction of additional engineering barriers, as well as by substantial decrease in LLW and ILW volume, localization of shallow

  8. Radioactive iodine separations and waste forms development.

    SciTech Connect

    Krumhansl, James Lee; Nenoff, Tina Maria; Garino, Terry J.; Rademacher, David

    2010-04-01

    Reprocessing nuclear fuel releases gaseous radio-iodine containing compounds which must be captured and stored for prolonged periods. Ag-loaded mordenites are the leading candidate for scavenging both organic and inorganic radioiodine containing compounds directly from reprocessing off gases. Alternately, the principal off-gas contaminant, I2, and I-containing acids HI, HIO3, etc. may be scavenged using caustic soda solutions, which are then treated with bismuth to put the iodine into an insoluble form. Our program is focused on using state-of-the-art materials science technologies to develop materials with high loadings of iodine, plus high long-term mechanical and thermal stability. In particular, we present results from research into two materials areas: (1) zeolite-based separations and glass encapsulation, and (2) in-situ precipitation of Bi-I-O waste forms. Ag-loaded mordenite is either commercially available or can be prepared via a simple Ag+ ion exchange process. Research using an Ag+-loaded Mordenite zeolite (MOR, LZM-5 supplied by UOP Corp.) has revealed that I2 is scavenged in one of three forms, as micron-sized AgI particles, as molecular (AgI)x clusters in the zeolite pores and as elemental I2 vapor. It was found that only a portion of the sorbed iodine is retained after heating at 95o C for three months. Furthermore, we show that even when the Ag-MOR is saturated with I2 vapor only roughly half of the silver reacted to form stable AgI compounds. However, the Iodine can be further retained if the AgI-MOR is then encapsulated into a low temperature glass binder. Follow-on studies are now focused on the sorption and waste form development of Iodine from more complex streams including organo-iodine compounds (CH3I). Bismuth-Iodate layered phases have been prepared from caustic waste stream simulant solutions. They serve as a low cost alternative to ceramics waste forms. Novel compounds have been synthesized and solubility studies have been completed

  9. Incorporation of radioactive wastes into styrenated polyester

    SciTech Connect

    Ikladious, N.E.; Ghattas, N.K.; Eskander, S.B.

    1986-01-01

    Styrenated polyester (poly(oxydiethylene maleate)) is examined as a medium for immobilization of simulated spent-ion exchange resin used at Inshas Reactor (Egypt). Compressive strength and hardness values illustrated the stability of the final products towards radiation. TG, DTG, and DTA diagrams showed the thermal instability of the final waste form at about 375/sup 0/C. Leaching experiment on incorporated blocks of active resin labelled with /sup 137/Cs, /sup 144/Ce, and /sup 106/Ru showed that the cumulative leaching rate for Ce is lower than those for Ru and Cs.

  10. Radioactive waste management approaches for developed countries

    SciTech Connect

    Patricia Paviet-Hartmann; Anthony Hechanova; Catherine Riddle

    2013-07-01

    Nuclear power has demonstrated over the last 30 years its capacity to produce base-load electricity at a low, predictable and stable cost due to the very low economic dependence on the price of uranium. However the management of used nuclear fuel remains the “Achilles’ Heel” of this energy source since the storage of used nuclear fuel is increasing as evidenced by the following number with 2,000 tons of UNF produced each year by the 104 US nuclear reactor units which equates to a total of 62,000 spent fuel assemblies stored in dry cask and 88,000 stored in pools. Two options adopted by several countries will be presented. The first one adopted by Europe, Japan and Russia consists of recycling the used nuclear fuel after irradiation in a nuclear reactor. Ninety six percent of uranium and plutonium contained in the spent fuel could be reused to produce electricity and are worth recycling. The separation of uranium and plutonium from the wastes is realized through the industrial PUREX process so that they can be recycled for re-use in a nuclear reactor as a mixed oxide (MOX) fuel. The second option undertaken by Finland, Sweden and the United States implies the direct disposal of used nuclear fuel into a geologic formation. One has to remind that only 30% of the worldwide used nuclear fuel are currently recycled, the larger part being stored (70% in pool) waiting for scientific or political decisions. A third option is emerging with a closed fuel cycle which will improve the global sustainability of nuclear energy. This option will not only decrease the volume amount of nuclear waste but also the long-term radiotoxicity of the final waste, as well as improving the long-term safety and the heat-loading of the final repository. At the present time, numerous countries are focusing on the R&D recycling activities of the ultimate waste composed of fission products and minor actinides (americium and curium). Several new chemical extraction processes, such as TRUSPEAK

  11. Tank Waste Transport Stability: Summaries of Hanford Slurry and Salt-Solution Studies in FY 2000

    SciTech Connect

    Welch, T.D.

    2002-07-08

    This report is a collection of summary articles on FY 2000 studies of slurry transport and salt-well pumping related to Hanford tank waste transfers. These studies are concerned with the stability (steady, uninterrupted flow) of tank waste transfers, a subset of the Department of Energy (DOE) Tanks Focus Area Tank (TFA) Waste Chemistry effort. This work is a collaborative effort of AEA Technology plc, the Diagnostic Instrumentation and Analysis Laboratory at Mississippi State University (DIAL-MSU), the Hemispheric Center for Environmental Technology at Florida International University (HCET-FIU), Numatec Hanford Corporation (NHC), and the Oak Ridge National Laboratory (ORNL). The purpose of this report is to provide, in a single document, an overview of these studies to help the reader identify contacts and resources for obtaining more detailed information and to help promote useful interchanges between researchers and users. Despite over 50 years of experience in transporting radioactive tank wastes to and from equipment and tanks at the Department of Energy's Hanford, Savannah River, and Oak Ridge sites, waste slurry transfer pipelines and process piping become plugged on occasion. At Hanford, several tank farm pipelines are no longer in service because of plugs. At Savannah River, solid deposits in the outlet line of the 2H evaporator have resulted in an unplanned extended downtime. Although waste transfer criteria and guidelines intended to prevent pipeline plugging are in place, they are not always adequate. To avoid pipeline plugging in the future, other factors that are not currently embodied in the transfer criteria may need to be considered. The work summarized here is being conducted to develop a better understanding of the chemical and waste flow dynamics during waste transfer. The goal is to eliminate pipeline plugs by improving analysis and engineering tools in the field that incorporate this understanding.

  12. Functional design criteria radioactive liquid waste line replacement, Project W-087. Revision 3

    SciTech Connect

    McVey, C.B.

    1994-10-13

    This document provides the functional design criteria for the 222-S Laboratory radioactive waste drain piping and transfer pipeline replacement. The project will replace the radioactive waste drain piping from the hot cells in 222-S to the 219-S Waste Handling Facility and provide a new waste transfer route from 219-S to the 244-S Catch Station in Tank Farms.

  13. Nuclear waste repositories in salt mines: a new approach to safety assessment.

    PubMed

    Memmert, G

    1996-08-01

    The long-term safety of radioactive waste repositories in rock-salt mines in the deep underground benefits significantly from the barrier effect of overlying rocks. The concentrations of radioactive substances released from the repository and migrating in the aquifer up to the biosphere are greatly reduced during passage through these rocks. In former safety analyses of waste repositories this transport has generally been modelled as a combination of the involved phenomena, e.g. convection, dispersion, adsorption, etc. The data required for a numerical evaluation of the overall effect are obtained either as (conservative) estimates based on experience or are empirical, based mainly on laboratory experiments. The approach presented here is much simpler and entirely empirical, and therefore more transparent. It makes use of the fact that the groundwater in the overlying rocks always contains dissolved salt from the salt formation and carries it continuously into the receiving channels or the drainage system. The relation between the total amount of dissolved solids present in a certain subsurface catchment area and their steady-state concentration in the receiving channels is assumed to be equivalent to the relation between the given amount of radionuclides released from the repository and their concentration in the receiving channels, the latter leading to a certain radiation exposure of the population. Two versions of this approach are discussed: version (a) assumes a continuous stream of radionuclides released from the repository, and version (b) assumes a pulse release of radionuclides from the repository. A simple calculation using data from the Gorleben exploration leads to the inequality [equation: see text] where Cmax is the maximum radionuclide concentration (with respect to time) in the receiving channels and W (Bq) is the amount of radionuclides released from the respository in a very short time. Cmax obtained from (1), is supposed to be an upper limit of

  14. Paradoxical hypertension and salt wasting in Type II Bartter syndrome.

    PubMed

    Chan, Winnie Kwai-Yu; To, Ka Fai; Tong, Joanna H M; Law, Chi Wai

    2012-06-01

    Ante/neonatal Bartter syndrome (BS) is a rare hereditary disorder. It is characterized by renal salt wasting, hypokalaemic metabolic alkalosis, high renin and aldosterone but normal blood pressure. We report a low birth weight newborn baby who presented with repeated apnoea shortly after birth as well as hyponatraemia, hypochloraemia, hyperkalaemia and metabolic acidosis. Her biochemical features mimicked pseudohypoaldosteronism but with initial hypertension, which had not been described in BS. Her subsequent genetic study confirmed two novel heterozygous mutations in the Exon 5 of KCNJ1 compatible with Type II BS. PMID:26069767

  15. Greater-confinement disposal of low-level radioactive wastes

    SciTech Connect

    Trevorrow, L.E.; Gilbert, T.L.; Luner, C.; Merry-Libby, P.A.; Meshkov, N.K.; Yu, C.

    1985-01-01

    Low-level radioactive wastes include a broad spectrum of wastes that have different radionuclide concentrations, half-lives, and physical and chemical properties. Standard shallow-land burial practice can provide adequate protection of public health and safety for most low-level wastes, but a small volume fraction (about 1%) containing most of the activity inventory (approx.90%) requires specific measures known as ''greater-confinement disposal'' (GCD). Different site characteristics and different waste characteristics - such as high radionuclide concentrations, long radionuclide half-lives, high radionuclide mobility, and physical or chemical characteristics that present exceptional hazards - lead to different GCD facility design requirements. Facility design alternatives considered for GCD include the augered shaft, deep trench, engineered structure, hydrofracture, improved waste form, and high-integrity container. Selection of an appropriate design must also consider the interplay between basic risk limits for protection of public health and safety, performance characteristics and objectives, costs, waste-acceptance criteria, waste characteristics, and site characteristics. This paper presents an overview of the factors that must be considered in planning the application of methods proposed for providing greater confinement of low-level wastes. 27 refs.

  16. Device Assembly Facility (DAF) Glovebox Radioactive Waste Characterization

    SciTech Connect

    Dominick, J L

    2001-12-18

    The Device Assembly Facility (DAF) at the Nevada Test Site (NTS) provides programmatic support to the Joint Actinide Shock Physics Experimental Research (JASPER) Facility in the form of target assembly. The target assembly activities are performed in a glovebox at DAF and include Special Nuclear Material (SNM). Currently, only activities with transuranic SNM are anticipated. Preliminary discussions with facility personnel indicate that primarily two distributions of SNM will be used: Weapons Grade Plutonium (WG-Pu), and Pu-238 enhanced WG-Pu. Nominal radionuclide distributions for the two material types are included in attachment 1. Wastes generated inside glove boxes is expected to be Transuranic (TRU) Waste which will eventually be disposed of at the Waste Isolation Pilot Plant (WIPP). Wastes generated in the Radioactive Material Area (RMA), outside of the glove box is presumed to be low level waste (LLW) which is destined for disposal at the NTS. The process knowledge quantification methods identified herein may be applied to waste generated anywhere within or around the DAF and possibly JASPER as long as the fundamental waste stream boundaries are adhered to as outlined below. The method is suitable for quantification of waste which can be directly surveyed with the Blue Alpha meter or swiped. An additional quantification methodology which requires the use of a high resolution gamma spectroscopy unit is also included and relies on the predetermined radionuclide distribution and utilizes scaling to measured nuclides for quantification.

  17. Reportable Nuclide Criteria for ORNL Radioactive Waste Management Activities - 13005

    SciTech Connect

    McDowell, Kip; Forrester, Tim; Saunders, Mark

    2013-07-01

    The U.S. Department of Energy's Oak Ridge National Laboratory (ORNL) in Oak Ridge, Tennessee generates numerous radioactive waste streams. Many of those streams contain a large number of radionuclides with an extremely broad range of concentrations. To feasibly manage the radionuclide information, ORNL developed reportable nuclide criteria to distinguish between those nuclides in a waste stream that require waste tracking versus those nuclides of such minimal activity that do not require tracking. The criteria include tracking thresholds drawn from ORNL onsite management requirements, transportation requirements, and relevant treatment and disposal facility acceptance criteria. As a management practice, ORNL maintains waste tracking on a nuclide in a specific waste stream if it exceeds any of the reportable nuclide criteria. Nuclides in a specific waste stream that screen out as non-reportable under all these criteria may be dropped from ORNL waste tracking. The benefit of these criteria is to ensure that nuclides in a waste stream with activities which meaningfully affect safety and compliance are tracked, while documenting the basis for removing certain isotopes from further consideration. (authors)

  18. Reductive Capacity Measurement of Waste Forms for Secondary Radioactive Wastes

    SciTech Connect

    Um, Wooyong; Yang, Jungseok; Serne, R. Jeffrey; Westsik, Joseph H.

    2015-09-28

    The reductive capacities of dry ingredients and final solid waste forms were measured using both the Cr(VI) and Ce(IV) methods and the results were compared. Blast furnace slag (BFS), sodium sulfide, SnF2, and SnCl2 used as dry ingredients to make various waste forms showed significantly higher reductive capacities compared to other ingredients regardless of which method was used. Although the BFS exhibits appreciable reductive capacity, it requires greater amounts of time to fully react. In almost all cases, the Ce(IV) method yielded larger reductive capacity values than those from the Cr(VI) method and can be used as an upper bound for the reductive capacity of the dry ingredients and waste forms, because the Ce(IV) method subjects the solids to a strong acid (low pH) condition that dissolves much more of the solids. Because the Cr(VI) method relies on a neutral pH condition, the Cr(VI) method can be used to estimate primarily the waste form surface-related and readily dissolvable reductive capacity. However, the Cr(VI) method does not measure the total reductive capacity of the waste form, the long-term reductive capacity afforded by very slowly dissolving solids, or the reductive capacity present in the interior pores and internal locations of the solids.

  19. Reductive capacity measurement of waste forms for secondary radioactive wastes

    NASA Astrophysics Data System (ADS)

    Um, Wooyong; Yang, Jung-Seok; Serne, R. Jeffrey; Westsik, Joseph H.

    2015-12-01

    The reductive capacities of dry ingredients and final solid waste forms were measured using both the Cr(VI) and Ce(IV) methods and the results were compared. Blast furnace slag (BFS), sodium sulfide, SnF2, and SnCl2 used as dry ingredients to make various waste forms showed significantly higher reductive capacities compared to other ingredients regardless of which method was used. Although the BFS exhibits appreciable reductive capacity, it requires greater amounts of time to fully react. In almost all cases, the Ce(IV) method yielded larger reductive capacity values than those from the Cr(VI) method and can be used as an upper bound for the reductive capacity of the dry ingredients and waste forms, because the Ce(IV) method subjects the solids to a strong acid (low pH) condition that dissolves much more of the solids. Because the Cr(VI) method relies on a neutral pH condition, the Cr(VI) method can be used to estimate primarily the waste form surface-related and readily dissolvable reductive capacity. However, the Cr(VI) method does not measure the total reductive capacity of the waste form, the long-term reductive capacity afforded by very slowly dissolving solids, or the reductive capacity present in the interior pores and internal locations of the solids.

  20. Transportation functions of the Civilian Radioactive Waste Management System

    SciTech Connect

    Shappert, L.B.; Attaway, C.R.; Pope, R.B. ); Best, R.E.; Danese, F.L. ); Dixon, L.D. , Martinez, GA ); Jones, R.H. , Los Gatos, CA ); Klimas, M.J. ); Peterson, R.W

    1992-03-01

    Within the framework of Public Law 97.425 and provisions specified in the Code of Federal Regulations, Title 10 Part 961, the US Department of Energy has the responsibility to accept and transport spent fuel and high-level waste from various organizations which have entered into a contract with the federal government in a manner that protects the health and safety of the public and workers. In implementing these requirements, the Office of Civilian Radioactive Waste Management (OCRWM) has, among other things, supported the identification of functions that must be performed by a transportation system (TS) that will accept the waste for transport to a federal facility for storage and/or disposal. This document, through the application of system engineering principles, identifies the functions that must be performed to transport waste under this law.

  1. Low-level radioactive waste form qualification testing

    SciTech Connect

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

    1998-06-01

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

  2. USING STATISTICAL PROCESS CONTROL TO MONITOR RADIOACTIVE WASTE CHARACTERIZATION AT A RADIOACTIVE FACILITY

    SciTech Connect

    WESTCOTT, J.L.; JOCHEN; PREVETTE

    2007-01-02

    Two facilities for storing spent nuclear fuel underwater at the Hanford site in southeastern Washington State are being removed from service, decommissioned, and prepared for eventual demolition. The fuel-storage facilities consist of two separate basins called K East (KE) and K West (KW) that are large subsurface concrete pools filled with water, with a containment structure over each. The basins presently contain sludge, debris, and equipment that have accumulated over the years. The spent fuel has been removed from the basins. The process for removing the remaining sludge, equipment, and structure has been initiated for the basins. Ongoing removal operations generate solid waste that is being treated as required, and then disposed. The waste, equipment and building structures must be characterized to properly manage, ship, treat (if necessary), and dispose as radioactive waste. As the work progresses, it is expected that radiological conditions in each basin may change as radioactive materials are being moved within and between the basins. It is imperative that these changing conditions be monitored so that radioactive characterization of waste is adjusted as necessary.

  3. USING STATISTICAL PROCESS CONTROL TO MONITOR RADIOACTIVE WASTE CHARACTERIZATION AT A RADIOACTIVE FACILITY

    SciTech Connect

    WESTCOTT, J.L.

    2006-11-15

    Two facilities for storing spent nuclear fuel underwater at the Hanford site in southeastern Washington State being removed from service, decommissioned, and prepared for eventual demolition. The fuel-storage facilities consist of two separate basins called K East (KE) and K West (KW) that are large subsurface concrete pools filled with water, with a containment structure over each. The basins presently contain sludge, debris, and equipment that have accumulated over the years. The spent fuel has been removed from the basins. The process for removing the remaining sludge, equipment, and structure has been initiated for the basins. Ongoing removal operations generate solid waste that is being treated as required, and then disposed. The waste, equipment and building structures must be characterized to properly manage, ship, treat (if necessary), and dispose as radioactive waste. As the work progresses, it is expected that radiological conditions in each basin may change as radioactive materials are being moved within and between the basins. It is imperative that these changing conditions be monitored so that radioactive characterization of waste is adjusted as necessary.

  4. Landsat investigations of the northern Paradox basin, Utah and Colorado: implications for radioactive waste emplacement

    USGS Publications Warehouse

    Friedman, Jules D.; Simpson, Shirley L.

    1978-01-01

    The first stages of a remote-sensing project on the Paradox basin, part of the USGS (U.S. Geological Survey) radioactive waste-emplacement program, consisted of a review and selection of the best available satellite scanner images to use in geomorphologic and tectonic investigations of the region. High-quality Landsat images in several spectral bands (E-2260-17124 and E-5165-17030), taken under low sun angle October 9 and 10, 1975, were processed via computer for planimetric rectification, histogram analysis, linear transformation of radiance values, and edge enhancement. A lineament map of the northern Paradox basin was subsequently compiled at 1:400,000 using the enhanced Landsat base. Numerous previously unmapped northeast-trending lineaments between the Green River and Yellowcat dome; confirmatory detail on the structural control of major segments of the Colorado, Gunnison, and Dolores Rivers; and new evidence for late Phanerozoic reactivation of Precambrian basement structures are among the new contributions to the tectonics of the region. Lineament trends appear to be compatible with the postulated Colorado lineament zone, with geophysical potential-field anomalies, and with a northeast-trending basement fault pattern. Combined Landsat, geologic, and geophysical field evidence for this interpretation includes the sinuousity of the composite Salt Valley anticline, the transection of the Moab-Spanish Valley anticline on its southeastern end by northeast-striking faults, and possible transection (?) of the Moab diapir. Similarly, northeast-trending lineaments in Cottonwood Canyon and elsewhere are interpreted as manifestations of structures associated with northeasterly trends in the magnetic and gravity fields of the La Sal Mountains region. Other long northwesterly lineaments near the western termination of the Ryan Creek fault zone. may be associated with the fault zone separating the Uncompahgre horst uplift from the Paradox basin. Implications of the

  5. Radioactive Waste Management information for 1994 and record-to-date

    SciTech Connect

    French, D.L.; Lisee, D.J.; Taylor, K.A.

    1995-07-01

    This document, Radioactive Waste Management Information for 1994 and Record-To-Date, contains computerized radioactive waste data records from the Idaho National Engineering Laboratory (INEL). Data are compiled from information supplied by the US Department of Energy (DOE) contractors. Data listed are on airborne and liquid radioactive effluents and solid radioactive waste that is stored, disposed, and sent to the INEL for reduction. Data are summarized for the years 1952 through 1993. Data are detailed for the calendar year 1994.

  6. Electric controlled air incinerator for radioactive wastes

    SciTech Connect

    Hootman, H.E.; Warren, J.H.

    1981-04-07

    A two-stage incinerator is provided which includes a primary combustion chamber and an afterburner chamber for off-gases. The latter is formed by a plurality of vertical tubes in combination with associated manifolds which connect the tubes together to form a continuous tortuous path. Electrically-controlled heaters surround the tubes while electrically-controlled plate heaters heat the manifolds. A gravity-type ash removal system is located at the bottom of the first afterburner tube while an air mixer is disposed in that same tube just above the outlet from the primary chamber. A ram injector in combination with rotary magazine feeds waste to a horizontal tube forming the primary combustion chamber.

  7. Natural Radioactivity in Monitoring Waste Disposals

    NASA Astrophysics Data System (ADS)

    de Meijer, R. J.; Limburg, J.; Venema, L. B.

    Monitoring large (underwater) surfaces, with strongly varying composition, requires a sampling density, exceeding the capabilities of standard techniques. These techniques involve sample collection and a number of treatments and measurements in laboratory; both steps are laborious, tedious and costly. This paper shows that a trailing detector system of natural γ-rays provides quantitative information on the dynamics at and around a waste disposal site. In this paper the technique is applied to monitor dumpsites of gold mines from an aircraft and the dispersal of dredge spoil from Rotterdam harbour dumped at the North Sea by vessel. The sea-floor monitoring has been conducted in detail, including the derivation of sediment composition and assessing by means of a mass-balance equation the transport directions and quantities in time.

  8. Electric controlled air incinerator for radioactive wastes

    DOEpatents

    Warren, Jeffery H.; Hootman, Harry E.

    1981-01-01

    A two-stage incinerator is provided which includes a primary combustion chamber and an afterburner chamber for off-gases. The latter is formed by a plurality of vertical tubes in combination with associated manifolds which connect the tubes together to form a continuous tortuous path. Electrically-controlled heaters surround the tubes while electrically-controlled plate heaters heat the manifolds. A gravity-type ash removal system is located at the bottom of the first afterburner tube while an air mixer is disposed in that same tube just above the outlet from the primary chamber. A ram injector in combination with rotary magazine feeds waste to a horizontal tube forming the primary combustion chamber.

  9. Overview of Fiscal Year 2002 Research and Development for Savannah River Site's Salt Waste Processing Facility

    SciTech Connect

    H. D. Harmon, R. Leugemors, PNNL; S. Fink, M. Thompson, D. Walker, WSRC; P. Suggs, W. D. Clark, Jr

    2003-02-26

    The Department of Energy's (DOE) Savannah River Site (SRS) high-level waste program is responsible for storage, treatment, and immobilization of high-level waste for disposal. The Salt Processing Program (SPP) is the salt (soluble) waste treatment portion of the SRS high-level waste effort. The overall SPP encompasses the selection, design, construction and operation of treatment technologies to prepare the salt waste feed material for the site's grout facility (Saltstone) and vitrification facility (Defense Waste Processing Facility). Major constituents that must be removed from the salt waste and sent as feed to Defense Waste Processing Facility include actinides, strontium, cesium, and entrained sludge. In fiscal year 2002 (FY02), research and development (R&D) on the actinide and strontium removal and Caustic-Side Solvent Extraction (CSSX) processes transitioned from technology development for baseline process selection to providing input for conceptual design of the Salt Waste Processing Facility. The SPP R&D focused on advancing the technical maturity, risk reduction, engineering development, and design support for DOE's engineering, procurement, and construction (EPC) contractors for the Salt Waste Processing Facility. Thus, R&D in FY02 addressed the areas of actual waste performance, process chemistry, engineering tests of equipment, and chemical and physical properties relevant to safety. All of the testing, studies, and reports were summarized and provided to the DOE to support the Salt Waste Processing Facility, which began conceptual design in September 2002.

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

    SciTech Connect

    Not Available

    1989-11-01

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

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

    SciTech Connect

    Not Available

    1990-11-01

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

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

    SciTech Connect

    Not Available

    1992-03-01

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

  13. DEVELOPMENT OF GLASS MATRICES FOR HLW RADIOACTIVE WASTES

    SciTech Connect

    Jantzen, C.

    2010-03-18

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

  14. Processing of solid mixed waste containing radioactive and hazardous materials

    DOEpatents

    Gotovchikov, V.T.; Ivanov, A.V.; Filippov, E.A.

    1998-05-12

    Apparatus for the continuous heating and melting of a solid mixed waste bearing radioactive and hazardous materials to form separate metallic, slag and gaseous phases for producing compact forms of the waste material to facilitate disposal includes a copper split water-cooled (cold) crucible as a reaction vessel for receiving the waste material. The waste material is heated by means of the combination of a plasma torch directed into the open upper portion of the cold crucible and an electromagnetic flux produced by induction coils disposed about the crucible which is transparent to electromagnetic fields. A metallic phase of the waste material is formed in a lower portion of the crucible and is removed in the form of a compact ingot suitable for recycling and further processing. A glass-like, non-metallic slag phase containing radioactive elements is also formed in the crucible and flows out of the open upper portion of the crucible into a slag ingot mold for disposal. The decomposition products of the organic and toxic materials are incinerated and converted to environmentally safe gases in the melter. 6 figs.

  15. Processing of solid mixed waste containing radioactive and hazardous materials

    DOEpatents

    Gotovchikov, Vitaly T.; Ivanov, Alexander V.; Filippov, Eugene A.

    1998-05-12

    Apparatus for the continuous heating and melting of a solid mixed waste bearing radioactive and hazardous materials to form separate metallic, slag and gaseous phases for producing compact forms of the waste material to facilitate disposal includes a copper split water-cooled (cold) crucible as a reaction vessel for receiving the waste material. The waste material is heated by means of the combination of a plasma torch directed into the open upper portion of the cold crucible and an electromagnetic flux produced by induction coils disposed about the crucible which is transparent to electromagnetic fields. A metallic phase of the waste material is formed in a lower portion of the crucible and is removed in the form of a compact ingot suitable for recycling and further processing. A glass-like, non-metallic slag phase containing radioactive elements is also formed in the crucible and flows out of the open upper portion of the crucible into a slag ingot mold for disposal. The decomposition products of the organic and toxic materials are incinerated and converted to environmentally safe gases in the melter.

  16. Non-Destructive Testing for Control of Radioactive Waste Package

    NASA Astrophysics Data System (ADS)

    Plumeri, S.; Carrel, F.

    2015-10-01

    Characterization and control of radioactive waste packages are important issues in the management of a radioactive waste repository. Therefore, Andra performs quality control inspection on radwaste package before disposal to ensure the compliance of the radwast characteristics with Andra waste disposal specifications and to check the consistency between Andra measurements results and producer declared properties. Objectives of this quality control are: assessment and improvement of producer radwaste packages quality mastery, guarantee of the radwaste disposal safety, maintain of the public confidence. To control radiological characteristics of radwaste package, non-destructive passive methods (gamma spectrometry and neutrons counting) are commonly used. These passive methods may not be sufficient, for instance to control the mass of fissile material contained inside radwaste package. This is particularly true for large concrete hull of heterogeneous radwaste containing several actinides mixed with fission products like 137Cs. Non-destructive active methods, like measurement of photofission delayed neutrons, allow to quantify the global mass of actinides and is a promising method to quantify mass of fissile material. Andra has performed different non-destructive measurements on concrete intermediate-level short lived nuclear waste (ILW-SL) package to control its nuclear material content. These tests have allowed Andra to have a first evaluation of the performance of photofission delayed neutron measurement and to identify development needed to have a reliable method, especially for fissile material mass control in intermediate-level long lived waste package.

  17. Waste Form Strategies for Mo-rich Radioactive Waste

    SciTech Connect

    Stewart, M.W.A.; Vance, E.R.

    2006-07-01

    This paper describes a small scoping study examining potential multiphase ceramic waste forms for wastes deriving from U-Mo research reactor fuel reprocessing. These fuels are being developed as replacements for silicide and aluminium fuels. The aim was to identify plausible phases that can be used in combination to achieve waste form monoliths with high waste loadings. These waste streams have unique challenges primarily because they have high Na, P and Mo contents. The approach taken was to utilize the Na and P and Mo to form phases that have been previously studied and are known to be durable. The Mo presents challenges because it is multivalent. In air it exists in the hexavalent state, but it can also be partially reduced to the tetravalent state and will substitute for Ti{sup 4+} in Synroc phases such as perovskite, rutile and pyrochlore. Under extremely reducing conditions it will be reduced to the metallic state. Five compositions were tested. The waste loadings ranged from 40 to {approx}77 wt%. Powellite (nominally, CaMoO{sub 4}) was one of the main phases formed in all compositions when Ca was added. Powellite was also formed by the coupled substitution of Na and Gd for Ca. Ba and Sr were also incorporated in the powellite. NZP (NaZr{sub 2}P{sub 3}O{sub 12}) and NTP (NaTi{sub 2}P{sub 3}O{sub 12}) were also found as major phases in some of the compositions tested. Attempts to incorporate Na as a Na-Gd-titanate perovskite did not work, and instead the Na tended to react with the Gd and Mo to form powellite. Left over Gd reacted with P to form monazite. Pyrochlore was formed in one sample in which it was a target phase, with Mo in the tetravalent state. This pyrochlore appears to be a Gd-Mo-Ti pyrochlore with Na and some Al incorporated, plus traces of other waste elements. The XRD pattern suggests pyrochlore although the composition as measured suggests that it is a defect pyrochlore with vacancies in the A-site. Ca phosphate phases were also detected in

  18. Migration of radioactive wastes: radionuclide mobilization by complexing agents.

    PubMed

    Means, J L; Crerar, D A; Duguid, J O

    1978-06-30

    Ion exchange, gel filtration chromatography, and gas chromatographymass spectrometry analyses have demonstrated that ethylenediaminetetraacetic acid (EDTA), an extremely strong complexing agent commonly used in decontamination operations at nuclear facilities, is causing the low-level migration of cobalt-60 from intermediate-level liquid waste disposal pits and trenches in the Oak Ridge National Laboratory burial grounds. Because it forms extremely strong complexes with rare earths and actinides, EDTA or similar chelates may also be contributing to the mobilization of these radionuclides from various terrestrial radioactive waste burial sites around the country. PMID:17757689

  19. High level radioactive waste management facility design criteria

    SciTech Connect

    Sheikh, N.A.; Salaymeh, S.R.

    1993-10-01

    This paper discusses the engineering systems for the structural design of the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). At the DWPF, high level radioactive liquids will be mixed with glass particles and heated in a melter. This molten glass will then be poured into stainless steel canisters where it will harden. This process will transform the high level waste into a more stable, manageable substance. This paper discuss the structural design requirements for this unique one of a kind facility. A special emphasis will be concentrated on the design criteria pertaining to earthquake, wind and tornado, and flooding.

  20. Office of Civilian Radioactive Waste Management annual report to Congress

    SciTech Connect

    1989-12-01

    This sixth Annual Report to Congress by the Office of Civilian Radioactive Waste Management (OCRWM) describes activities and expenditures of the Office during fiscal year 1988. An epilogue chapter reports significant events from the end of the fiscal year on September 30, 1988 through March 1989. The Nuclear Waste Policy Amendments Act (NWPA) of 1987 made significant changes to the NWPA relating to repository siting and monitored retrievable storage and added new provisions for the establishment of several institutional entities with which OCRWM will interact. Therefore, a dominant theme throughout this report is the implementation of the policy focus and specific provisions of the Amendments Act. 50 refs., 8 figs., 4 tabs.

  1. Pilot studies to achieve waste minimization and enhance radioactive liquid waste treatment at the Los Alamos National Laboratory Radioactive Liquid Waste Treatment Facility

    SciTech Connect

    Freer, J.; Freer, E.; Bond, A.

    1996-07-01

    The Radioactive and Industrial Wastewater Science Group manages and operates the Radioactive Liquid Waste Treatment Facility (RLWTF) at the Los Alamos National Laboratory (LANL). The RLWTF treats low-level radioactive liquid waste generated by research and analytical facilities at approximately 35 technical areas throughout the 43-square-mile site. The RLWTF treats an average of 5.8 million gallons (21.8-million liters) of liquid waste annually. Clarifloculation and filtration is the primary treatment technology used by the RLWTF. This technology has been used since the RLWTF became operable in 1963. Last year the RLWTF achieved an average of 99.7% removal of gross alpha activity in the waste stream. The treatment process requires the addition of chemicals for the flocculation and subsequent precipitation of radionuclides. The resultant sludge generated during this process is solidified in drums and stored or disposed of at LANL.

  2. THE USE OF POLYMERS IN RADIOACTIVE WASTE PROCESSING SYSTEMS

    SciTech Connect

    Skidmore, E.; Fondeur, F.

    2013-04-15

    The Savannah River Site (SRS), one of the largest U.S. Department of Energy (DOE) sites, has operated since the early 1950s. The early mission of the site was to produce critical nuclear materials for national defense. Many facilities have been constructed at the SRS over the years to process, stabilize and/or store radioactive waste and related materials. The primary materials of construction used in such facilities are inorganic (metals, concrete), but polymeric materials are inevitably used in various applications. The effects of aging, radiation, chemicals, heat and other environmental variables must therefore be understood to maximize service life of polymeric components. In particular, the potential for dose rate effects and synergistic effects on polymeric materials in multivariable environments can complicate compatibility reviews and life predictions. The selection and performance of polymeric materials in radioactive waste processing systems at the SRS are discussed.

  3. Defense waste processing facility radioactive operations. Part 1 - operating experience

    SciTech Connect

    Little, D.B.; Gee, J.T.; Barnes, W.M.

    1997-12-31

    The Savannah River Site`s Defense Waste Processing Facility (DWPF) near Aiken, SC is the nation`s first and the world`s largest vitrification facility. Following a ten year construction program and a 3 year non-radioactive test program, DWPF began radioactive operations in March 1996. This paper presents the results of the first 9 months of radioactive operations. Topics include: operations of the remote processing equipment reliability, and decontamination facilities for the remote processing equipment. Key equipment discussed includes process pumps, telerobotic manipulators, infrared camera, Holledge{trademark} level gauges and in-cell (remote) cranes. Information is presented regarding equipment at the conclusion of the DWPF test program it also discussed, with special emphasis on agitator blades and cooling/heating coil wear. 3 refs., 4 figs.

  4. Summary of radioactive solid waste received in the 200 Areas during calendar year 1993

    SciTech Connect

    Anderson, J.D.; Hagel, D.L.

    1994-09-01

    Westinghouse Hanford Company manages and operates the Hanford Site 200 Areas radioactive solid waste storage and disposal facilities for the US Department of Energy, Richland Operations Office. These facilities include radioactive solid waste disposal sites and radioactive solid waste storage areas. This document summarizes the amount of radioactive materials that have been buried and stored in the 200 Areas radioactive solid waste storage and disposal facilities since startup in 1944 through calendar year 1993. This report does not include backlog waste, solid radioactive waste in storage or disposed of in other areas, or facilities such as the underground tank farms. Unless packaged within the scope of WHC-EP-0063, ``Hanford Site Solid Waste Acceptance Criteria,`` (WHC 1988), liquid waste data are not included in this document.

  5. No Time Wasted. 25 years COVRA: Radioactive Waste Management in the Netherlands

    SciTech Connect

    Codee, H.D.K.; Verhoef, E.V.

    2008-07-01

    Time will render radioactive waste harmless. How can we manage the time radioactive substances remain harmful? Just 'wait and see' or 'marking time' is not an option. We need to isolate the waste from our living environment and control it as long as necessary. December 2007 was a time to commemorate, as the national waste management organisation of the Netherlands, COVRA, celebrated its 12. anniversary. During this period of 25 years a stable policy has been formulated and implemented. For the situation in the Netherlands, it was obvious that a period of long term storage was needed. Both the small volume of waste and the limited financial possibilities are determining factors. Time is needed to let the volume of waste grow and to let the money, needed for disposal, grow in a capital growth fund. A historical overview of the activities of COVRA is presented and lessons learned over a period of 25 years are given. (authors)

  6. Status of low-level radioactive waste management in Korea

    SciTech Connect

    Lee, K.J.

    1993-03-01

    The Republic of Korea has accomplished dramatic economic growth over the past three decades; demand for electricity has rapidly grown more than 15% per year. Since the first nuclear power plant, Kori-1 [587 MWe, pressurized water reactor (PWR)], went into commercial operation in 1978, the nuclear power program has continuously expanded and played a key role in meeting the national electricity demand. Nowadays, Korea has nine nuclear power plants [eight PWRs and one Canadian natural uranium reactor (CANDU)] in operation with total generating capacity of 7,616 MWe. The nuclear share of total electrical capacity is about 36%; however, about 50% of actual electricity production is provided by these nine nuclear power plants. In addition, two PWRs are under construction, five units (three CANDUs and two PWRs) are under design, and three more CANDUs and eight more PWRs are planned to be completed by 2006. With this ambitious nuclear program, the total nuclear generating capacity will reach about 23,000 MWe and the nuclear share will be about 40% of the total generating capacity in the year 2006. In order to expand the nuclear power program this ambitiously, enormous amounts of work still have to be done. One major area is radioactive waste management. This paper reviews the status of low-level radioactive waste management in Korea. First, the current and future generation of low-level radioactive wastes are estimated. Also included are the status and plan for the construction of a repository for low-level radioactive wastes, which is one of the hot issues in Korea. Then, the nuclear regulatory system is briefly mentioned. Finally, the research and development activities for LLW management are briefly discussed.

  7. Remote radioactive waste drum inspection with an autonomous mobile robot

    SciTech Connect

    Heckendorn, F.M.; Ward, C.R.; Wagner, D.G.

    1992-11-01

    An autonomous mobile robot is being developed to perform remote surveillance and inspection task on large numbers of stored radioactive waste drums. The robot will be self guided through narrow storage aisles and record the visual image of each viewable drum for subsequent off line analysis and archiving. The system will remove the personnel from potential exposure to radiation, perform the require inspections, and improve the ability to assess the long term trends in drum conditions.

  8. Remote radioactive waste drum inspection with an autonomous mobile robot

    SciTech Connect

    Heckendorn, F.M.; Ward, C.R.; Wagner, D.G.

    1992-01-01

    An autonomous mobile robot is being developed to perform remote surveillance and inspection task on large numbers of stored radioactive waste drums. The robot will be self guided through narrow storage aisles and record the visual image of each viewable drum for subsequent off line analysis and archiving. The system will remove the personnel from potential exposure to radiation, perform the require inspections, and improve the ability to assess the long term trends in drum conditions.

  9. [Board on Radioactive Waste Managements action on progress toward objectives

    SciTech Connect

    Not Available

    1994-11-28

    This report is a progress report to the US DOE from the Board on Radioactive Waste Management (BRWM), which summarizes the activities of the board during the period December 1, 1993 to May 2, 1994. The report summarizes the meetings of the board as a whole, of various of its subcommittees, and of activities it has undertaken to further its original mission. This board is associated with the National Research Council to give advice to US DOE.

  10. Xenon gamma-ray spectrometer for radioactive waste controlling complex

    NASA Astrophysics Data System (ADS)

    Ulin, S.; Novikov, A.; Dmitrenko, V.; Vlasik, K.; Krivova, K.; Petrenko, D.; Uteshev, Z.; Shustov, A.; Petkovich, E.

    2016-02-01

    Xenon detector based gamma-ray spectrometer for a radioactive waste sorting complex and its characteristics are described. It has been shown that the “thin-wall” modification of the detector allows better registration of low-energy gamma rays (tens of keV). The spectrometer is capable of operation in unfavorable field conditions and can identify radionuclides of interest in less than 1 second.

  11. Geohazards due to technologically enhanced natural radioactive wastes

    NASA Astrophysics Data System (ADS)

    Steinhäusler, Friedrich

    2010-10-01

    Human activities can modify naturally occurring radioactive material (NORM) into technologically enhanced naturally occurring radioactive material (TENORM) as a result of industrial activities. Most of these industries do not intend to work with radioactive material a priori. However, whenever a uranium- or thorium-bearing mineral is exploited, NORM-containing by-products and TENORM-contaminated wastes are created. The industrial use of NORM can result in non-negligible radiation exposure of workers and members of the public, exceeding by far the radiation exposure from nuclear technologies. For decades, millions of tons of NORM have been released into the environment without adequate control or even with the lack of any control. Various technologies have been developed for the control of NORM wastes. The paper discusses the merits and limitations of different NORM-waste management techniques, such as Containment, Immobilization, Dilution/Dispersion, Natural Attenuation, Separation, and - as an alternative - Cleaner Technologies. Each of these methods requires a comprehensive risk-benefit-cost analysis.

  12. Radioactive wastes dispersed in stabilized ash cements

    SciTech Connect

    Rubin, J.B.; Taylor, C.M.V.; Sivils, L.D.; Carey, J.W.

    1997-12-31

    One of the most widely-used methods for the solidification/stabilization of low-level radwaste is by incorporation into Type-I/II ordinary portland cement (OPC). Treating of OPC with supercritical fluid carbon dioxide (SCCO{sub 2}) has been shown to significantly increase the density, while simultaneously decreasing porosity. In addition, the process significantly reduces the hydrogenous content, reducing the likelihood of radiolytic decomposition reactions. This, in turn, permits increased actinide loadings with a concomitant reduction in disposable waste volume. In this article, the authors discuss the combined use of fly-ash-modified OPC and its treatment with SCCO{sub 2} to further enhance immobilization properties. They begin with a brief summary of current cement immobilization technology in order to delineate the areas of concern. Next, supercritical fluids are described, as they relate to these areas of concern. In the subsequent section, they present an outline of results on the application of SCCO{sub 2} to OPC, and its effectiveness in addressing these problem areas. Lastly, in the final section, they proffer their thoughts on why they believe, based on the OPC results, that the incorporation of fly ash into OPC, followed by supercritical fluid treatment, can produce highly efficient wasteforms.

  13. Civilian radioactive waste management program plan. Revision 2

    SciTech Connect

    1998-07-01

    This revision of the Civilian Radioactive Waste Management Program Plan describes the objectives of the Civilian Radioactive Waste management Program (Program) as prescribed by legislative mandate, and the technical achievements, schedule, and costs planned to complete these objectives. The Plan provides Program participants and stakeholders with an updated description of Program activities and milestones for fiscal years (FY) 1998 to 2003. It describes the steps the Program will undertake to provide a viability assessment of the Yucca Mountain site in 1998; prepare the Secretary of Energy`s site recommendation to the President in 2001, if the site is found to be suitable for development as a repository; and submit a license application to the Nuclear Regulatory Commission in 2002 for authorization to construct a repository. The Program`s ultimate challenge is to provide adequate assurance to society that an operating geologic repository at a specific site meets the required standards of safety. Chapter 1 describes the Program`s mission and vision, and summarizes the Program`s broad strategic objectives. Chapter 2 describes the Program`s approach to transform strategic objectives, strategies, and success measures to specific Program activities and milestones. Chapter 3 describes the activities and milestones currently projected by the Program for the next five years for the Yucca Mountain Site Characterization Project; the Waste Acceptance, Storage and Transportation Project; ad the Program Management Center. The appendices present information on the Nuclear Waste Policy Act of 1982, as amended, and the Energy Policy Act of 1992; the history of the Program; the Program`s organization chart; the Commission`s regulations, Disposal of High-Level Radioactive Wastes in geologic Repositories; and a glossary of terms.

  14. Identifying suitable "piercement" salt domes for nuclear waste storage sites

    SciTech Connect

    Kehle, R.

    1980-08-01

    Piercement salt domes of the northern interior salt basins of the Gulf of Mexico are being considered as permanent storage sites for both nuclear and chemically toxic wastes. The suitable domes are stable and inactive, having reached their final evolutionary configuration at least 30 million years ago. They are buried to depths far below the level to which erosion will penetrate during the prescribed storage period and are not subject to possible future reactivation. The salt cores of these domes are themselves impermeable, permitting neither the entry nor exit of ground water or other unwanted materials. In part, a stable dome may be recognized by its present geometric configuration, but conclusive proof depends on establishing its evolutionary state. The evolutionary state of a dome is obtained by reconstructing the growth history of the dome as revealed by the configuration of sedimentary strata in a large area (commonly 3,000 square miles or more) surrounding the dome. A high quality, multifold CDP reflection seismic profile across a candidate dome will provide much of the necessary information when integrated with available subsurface control. Additional seismic profiles may be required to confirm an apparent configuration of the surrounding strata and an interpreted evolutionary history. High frequency seismic data collected in the near vicinity of a dome are also needed as a supplement to the CDP data to permit accurate depiction of the configuration of shallow strata. Such data must be tied to shallow drill hole control to confirm the geologic age at which dome growth ceased. If it is determined that a dome reached a terminal configuration many millions of years ago, such a dome is incapable of reactivation and thus constitutes a stable storage site for nuclear wastes.

  15. Radioactive Liquid Waste Treatment Facility Discharges in 2011

    SciTech Connect

    Del Signore, John C.

    2012-05-16

    This report documents radioactive discharges from the TA50 Radioactive Liquid Waste Treatment Facilities (RLWTF) during calendar 2011. During 2011, three pathways were available for the discharge of treated water to the environment: discharge as water through NPDES Outfall 051 into Mortandad Canyon, evaporation via the TA50 cooling towers, and evaporation using the newly-installed natural-gas effluent evaporator at TA50. Only one of these pathways was used; all treated water (3,352,890 liters) was fed to the effluent evaporator. The quality of treated water was established by collecting a weekly grab sample of water being fed to the effluent evaporator. Forty weekly samples were collected; each was analyzed for gross alpha, gross beta, and tritium. Weekly samples were also composited at the end of each month. These flow-weighted composite samples were then analyzed for 37 radioisotopes: nine alpha-emitting isotopes, 27 beta emitters, and tritium. These monthly analyses were used to estimate the radioactive content of treated water fed to the effluent evaporator. Table 1 summarizes this information. The concentrations and quantities of radioactivity in Table 1 are for treated water fed to the evaporator. Amounts of radioactivity discharged to the environment through the evaporator stack were likely smaller since only entrained materials would exit via the evaporator stack.

  16. 10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation....

  17. 10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation....

  18. 10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation....

  19. 10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation....

  20. 10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation....

  1. Research on uranium deposits as analogies of radioactive waste repositories

    SciTech Connect

    Hardy, C.J.

    1988-01-01

    The disposal of highly radioactive waste deep underground in suitable geological formations is proposed by many countries to protect public health and safety. The study of natural analogies of nuclear waste repositories is one method of validating mathematical models and assuring that a proposed repository site and design will be safe. Since 1981, the AAEC has studied the major uranium deposits in the Alligator Rivers region of the Northern Territory of Australia as natural analogues of radioactive waste repositories. Results have been obtained on the following: (1) the migration of uranium, thorium and radium isotopes, (2) the behavior of naturally occurring levels of selected fission products and transuranium nuclides, e.g. technetium-99, iodine-129 and plutonium-239; (3) the role of specific minerals in retarding migration, and (4) the importance of colloidal material, in the migration of thorium. The AAEC has initiated a wider international project entitled The Alligator Rivers Analogue Project which will enable participating organizations to obtain additional results and to apply them in modeling, planning and regulating waste repositories.

  2. Microbial transformation of low-level radioactive waste

    SciTech Connect

    Francis, A.J.

    1980-06-01

    Microorganisms play a significant role in the transformation of the radioactive waste and waste forms disposed of at shallow-land burial sites. Microbial degradation products of organic wastes may influence the transport of buried radionuclides by leaching, solubilization, and formation of organoradionuclide complexes. The ability of indigenous microflora of the radioactive waste to degrade the organic compounds under aerobic and anaerobic conditions was examined. Leachate samples were extracted with methylene chloried and analyzed for organic compounds by gas chromatography and mass spectrometry. In general, several of the organic compounds in the leachates were degraded under aerobic conditions. Under anaerobic conditions, the degradation of the organics was very slow, and changes in concentrations of several acidic compounds were observed. Several low-molecular-weight organic acids are formed by breakdown of complex organic materials and are further metabolized by microorganisms; hence these compounds are in a dynamic state, being both synthesized and destroyed. Tributyl phosphate, a compound used in the extraction of metal ions from solutions of reactor products, was not degraded under anaerobic conditions.

  3. Teaching Radioactive Waste Management in an Undergraduate Engineering Program - 13269

    SciTech Connect

    Ikeda, Brian M.

    2013-07-01

    The University of Ontario Institute of Technology is Ontario's newest university and the only one in Canada that offers an accredited Bachelor of Nuclear Engineering (Honours) degree. The nuclear engineering program consists of 48 full-semester courses, including one on radioactive waste management. This is a design course that challenges young engineers to develop a fundamental understanding of how to manage the storage and disposal of various types and forms of radioactive waste, and to recognize the social consequences of their practices and decisions. Students are tasked with developing a major project based on an environmental assessment of a simple conceptual design for a waste disposal facility. They use collaborative learning and self-directed exploration to gain the requisite knowledge of the waste management system. The project constitutes 70% of their mark, but is broken down into several small components that include, an environmental assessment comprehensive study report, a technical review, a facility design, and a public defense of their proposal. Many aspects of the project mirror industry team project situations, including the various levels of participation. The success of the students is correlated with their engagement in the project, the highest final examination scores achieved by students with the strongest effort in the project. (authors)

  4. Location and identification of radioactive waste in Massachusetts Bay

    SciTech Connect

    Colton, D.P.; Louft, H.L.

    1993-12-31

    The accurate location and identification of hazardous waste materials dumped in the world`s oceans are becoming an increasing concern. For years, the oceans have been viewed as a convenient and economical place to dispose of all types of waste. In all but a few cases, major dump sites have been closed leaving behind years of accumulated debris. The extent of past environmental damage, the possibility of continued environmental damage, and the possibility of hazardous substances reaching the human food chain need to be carefully investigated. This paper reports an attempt to accurately locate and identify the radioactive component of the waste material. The Department of Energy`s Remote Sensing Laboratory (RSL), in support of the US Environmental Protection Agency (EPA), provided the precision navigation system and prototype underwater radiological monitoring equipment that were used during this project. The paper also describes the equipment used, presents the data obtained, and discusses future equipment development.

  5. Use of commercial robotics in radioactive waste shipping and receiving

    SciTech Connect

    Berger, J.D.

    1985-03-15

    Radioactive waste shipping and receiving facilities presently planned for commercial and defense nuclear waste will handle waste packages at frequencies far in excess of those in common practice today. Unacceptable personnel exposure to ionizing radiation is projected to occur if current limits for radiation levels at the cask surface and current handling methods are used. To reduce these exposure levels, alternate handling methods are being developed and demonstrated. The production nature of cask receiving operations suggests commercial robotics be incorporated into a remote handling system to reduce predicted worker exposure to acceptable levels, while maintaining or increasing throughput. The first phase of cask handling system development culminated in a proof-of-principle test demonstrating the feasibility of performing cask receiving and unloading operations in a remote and partially automated manner. 6 refs., 12 figs.

  6. Monitoring technologies for ocean disposal of radioactive waste

    SciTech Connect

    Triplett, M.B.; Solomon, K.A.; Bishop, C.B.; Tyce, R.C.

    1982-01-01

    The feasibility of using carefully selected subseabed locations to permanently isolate high level radioactive wastes at ocean depths greater than 4000 meters is discussed. Disposal at several candidate subseabed areas is being studied because of the long term geologic stability of the sediments, remoteness from human activity, and lack of useful natural resources. While the deep sea environment is remote, it also poses some significant challenges for the technology required to survey and monitor these sites, to identify and pinpoint container leakage should it occur, and to provide the environmental information and data base essential to determining the probable impacts of any such occurrence. Objectives and technical approaches to aid in the selective development of advanced technologies for the future monitoring of nuclear low level and high level waste disposal in the deep seabed are presented. Detailed recommendations for measurement and sampling technology development needed for deep seabed nuclear waste monitoring are also presented.

  7. Sintered bentonite ceramics for the immobilization of cesium- and strontium-bearing radioactive waste

    NASA Astrophysics Data System (ADS)

    Ortega, Luis Humberto

    The Advanced Fuel Cycle Initiative (AFCI) is a Department of Energy (DOE) program, that has been investigating technologies to improve fuel cycle sustainability and proliferation resistance. One of the program's goals is to reduce the amount of radioactive waste requiring repository disposal. Cesium and strontium are two primary heat sources during the first 300 years of spent nuclear fuel's decay, specifically isotopes Cs-137 and Sr-90. Removal of these isotopes from spent nuclear fuel will reduce the activity of the bulk spent fuel, reducing the heat given off by the waste. Once the cesium and strontium are separated from the bulk of the spent nuclear fuel, the isotopes must be immobilized. This study is focused on a method to immobilize a cesium- and strontium-bearing radioactive liquid waste stream. While there are various schemes to remove these isotopes from spent fuel, this study has focused on a nitric acid based liquid waste. The waste liquid was mixed with the bentonite, dried then sintered. To be effective sintering temperatures from 1100 to 1200°C were required, and waste concentrations must be at least 25 wt%. The product is a leach resistant ceramic solid with the waste elements embedded within alumino-silicates and a silicon rich phase. The cesium is primarily incorporated into pollucite and the strontium into a monoclinic feldspar. The simulated waste was prepared from nitrate salts of stable ions. These ions were limited to cesium, strontium, barium and rubidium. Barium and rubidium will be co-extracted during separation due to similar chemical properties to cesium and strontium. The waste liquid was added to the bentonite clay incrementally with drying steps between each addition. The dry powder was pressed and then sintered at various temperatures. The maximum loading tested is 32 wt. percent waste, which refers to 13.9 wt. percent cesium, 12.2 wt. percent barium, 4.1 wt. percent strontium, and 2.0 wt. percent rubidium. Lower loadings of waste

  8. A Challenge for Radioactive Waste Management: Memory Preservation

    SciTech Connect

    Charton, P.; Ouzounian, G.

    2008-07-01

    ANDRA, the French National Radioactive Waste Management Agency, is responsible for managing all radioactive waste in France over the long term. In the case of short-lived waste for which disposal facilities have a life expectancy of a few centuries, the Agency has set up a system for preserving the memory of those sites. Based on the historical analysis on a comparable timescale and on an appraisal of information-conservation means, a series of regulatory as well as technical provisions was made in order to ensure that sound information be transmitted to future generations. Requirements associated to the provisions deal mostly with legibility and a clear understanding of the information that must be decrypted and understood at least during the lifetime of the facilities (i.e., a few centuries). It must therefore be preserved throughout the same period. Responses to the requirements will be presented notably on various information-recording media, together with the information-diffusion strategy to the different authorities and structures within French society. A concrete illustration of the achievements made so far is the Centre de la Manche Disposal Facility, which was closed down in 1994 and is currently in its post-closure monitoring phase since 2003. In the case of deep geological repositories for long-lived radioactive waste, preserving memory takes a different aspect. First of all, timescales are much longer and are counted in hundreds of thousands of years. It is therefore much more difficult to consider how to maintain the richness of the information over such time periods than it is for short-lived waste. Both the nature and the form of the information to be transmitted must be revised. It would be risky indeed to base memory preservation over the long term on similar mechanisms beyond 1,000 years. Based on the heritage of a much more ancient history, we must seek to find appropriate means in order to develop surface markers and even more to ensure their

  9. A MODULAR STORE FOR DRUMS OF RADIOACTIVE WASTE

    SciTech Connect

    Sims, J.; Holden, G.

    2003-02-27

    Currently, the United Kingdom has no facility for the disposal of any waste above the low level category, indicating that all intermediate and high level waste, apart from spent fuel, has to be stored on the site of origin. To meet this storage requirement, nuclear sites are resorting to converting existing buildings or contemplating the construction of dedicated facilities, resulting in considerable cost implications. These financing aspects not only concern the construction strategy but also impinge on the ultimate decommissioning costs associated with each particular nuclear site. This paper reports on an investigation to apply the commercially available interlocking hollow block system to the design of a store for drums of radioactive waste. This block system can be quickly, and cost effectively, erected and filled with a choice of dense material. Later, the store can be dismantled with a minimum of disposable radioactive waste and the complete facility re - erected at another location if required, considerably reducing both capital construction and decommissioning costs. The investigation also encompassed a detailed review of the equipment required to place the drums of waste into the store, resulting in a scheme for a remotely operated vehicle that did not rely on umbilical control cables. The drum handler design included for 100% redundancy of all functions, meaning that whichever component failed, the handler was always recoverable to effect the necessary repair. The ultimate aim of the waste drum store review was to produce a facility that was as safe as a conventionally constructed unit, but at a lower overall building and decommissioning cost.

  10. Microcomputer spreadsheets for radioactive waste management and shielding analysis

    SciTech Connect

    Negin, C.A.; Worku, G.; Urland, C.S.; Deltete, C.P.

    1986-09-01

    The report describes several desktop computer applications for use in radioactive waste management and radiation engineering analyses. TMI-2 engineers identified eight calculational tasks that could be facilitated by use of microcomputer spreadsheet software. For each case, appropriate calculations and frequently used data on isotopes and materials were built into the six spreadsheet templates for calculating and thirty-one data sets. The spreadsheets are distributed on diskette to avoid the need for re-input by the user. The applications include: (1) calculation of radioactive decay for a specific time interval for often found radionuclides and their daughter radioisotopes; (2) determination of the disposability of a waste container by calculating the 10CFR61 sum-of-the-fractions concentrations; (3) estimation of the cost of a waste disposal campaign; (4) projection of the rate of hydrogen gas generation in waste liners; (5) calculation of the dose rate from a point source; and (6) inferring the radioisotopic contents of a waste container from the external radiation, and vice versa. This set of microcomputer spreadsheet applications was selected for inclusion as part of EPRI's task of transferring technology developed at TMI-2 to the industry. Waste management is a concern to all plant operators. Microcomputers are used for a variety of evaluation and recordkeeping tasks. Increasing their usefulness to engineers by tools such as these spreadsheets serves to relieve analysts of calculational tasks. As the report provides background and development histories, and the spreadsheets are presented on diskette, the product can be used as a training tool as well.

  11. Bases, Assumptions, and Results of the Flowsheet Calculations for the Decision Phase Salt Disposition Alternatives

    SciTech Connect

    Elder, H.H.

    2001-07-11

    The HLW salt waste (salt cake and supernate) now stored at the SRS must be treated to remove insoluble sludge solids and reduce the soluble concentration of radioactive cesium radioactive strontium and transuranic contaminants (principally Pu and Np). These treatments will enable the salt solution to be processed for disposal as saltstone, a solid low-level waste.

  12. Geochemical modeling (EQ3/6) plan: Office of Civilian Radioactive Waste Management Program

    SciTech Connect

    McKenzie, W.F.; Wolery, T.J.; Delany, J.M.; Silva, R.J.; Jackson, K.J.; Bourcier, W.L.; Emerson, D.O.

    1986-08-28

    This plan replaces an earlier plan for the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. It includes activities for all repository projects in the Office of Geologic Repositories: NNWSI, the Basalt Waste Isolation Project, the Salt Repository Project, and the Crystalline Project. Each of these projects is part of the Office of Civilian Radioactive Waste Management (OCRWM) Program. The scope of work for fiscal years 1986 to 1992 includes the work required to upgrade the geochemical codes and supporting data bases, to permit modeling of chemical processes associated with nuclear waste repositories in four geological environments: tuff, salt, basalt, and crystalline rock. Planned tasks include theoretical studies and code development to take account of the effects of precipitation kinetics, sorption, solid solutions, glass/water interactions, variable gas fugacities, and simple mass transport. Recent progress has been made in the ability of the codes to account for precipitation kinetics, highly-saline solutions, and solid solutions. Transition state theory was re-examined resulting in new insights that will provide the foundation for further improvements necessary to model chemical kinetics. Currently there is an increased effort that is concentrated on the supporting data base. For aqueous species and solid phases, specific to nuclear waste, requisite thermodynamic values reported in the literature are being evaluated and for cases where essential data is lacking, laboratory measurements will be carried out. Significant modifications and expansions have been made to the data base. During FY86, the total number of species in the data base has almost doubled and many improvements have been made with regard to consistency, organization, user applications, and documentation. Two Ridge computers using a RISC implementation of UNIX were installed; they are completely dedicated EQ3/6 machines.

  13. Management of salt waste from electrochemical processing of used nuclear fuel

    SciTech Connect

    Simpson, M.F.; Patterson, M.N.; Lee, J.; Wang, Y.; Versey, J.; Phongikaroon, S.

    2013-07-01

    Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electro-refiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form. (authors)

  14. Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel

    SciTech Connect

    Michael F. Simpson; Michael N. Patterson; Joon Lee; Yifeng Wang; Joshua Versey; Ammon Williams; Supathorn Phongikaroon; James Allensworth; Man-Sung Yim

    2013-10-01

    Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electrorefiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form.

  15. Properties of radioactive wastes and waste containers. Status report, October 1980-September 1981

    SciTech Connect

    Morcos, N.; Dayal, R.; Weiss, A.J.

    1982-04-01

    Licensing of near surface low-level radioactive waste disposal sites and waste forms/containers requires the ability to predict the dispersibility of radionuclides from waste forms and waste containers disposed in burial sites. Basic concerns in licensing radioactive waste forms and containers are their dimensional stability and the potential for migration of the radionuclides enclosed therein in a near- and long-term predictable fashion. To assess these concerns, a data base is needed for evaluating the acceptability of solidified low-level radioactive waste packages for disposal. Furthermore, the need to develop test procedures and methodologies exists to enable the prediction of long-term performance of waste forms based on short-term laboratory tests. The objectives of the research program at BNL are to provide an improved understanding of phenomena, testing methodology and data. This improves the NRC's capability to predict low-level waste isolation performance, and to provide a better technical basis for regulatory standards. The areas addressed to meet these objectives during the 1981 fiscal year were: leachability and compressive strength of boric acid waste in Portland III cement. The tracers used for the study were /sup 137/Cs, /sup 85/Sr, and /sup 60/Co; leachability of /sup 137/Cs, /sup 85/Sr, and /sup 60/Co from organic ion exchange resin/Portland III and Lumnite cements; displacement of /sup 137/Cs, /sup 85/Sr, and /sup 60/Co from organic ion exchange resins upon mixing with Portland II and Lumnite cements; leachability of organic ion exchange resins/Bitumen composites using resins in the H/sup +/, Na/sup +/, Cs/sup +/, Sr/sup +2/, and SO/sub 4//sup -2/ forms, and /sup 137/Cs and /sup 85/Sr tracers; correlation of /sup 137/Cs leachability from small-scale (laboratory) samples to large-scale waste forms; and hydrostatic testing of DOT 17H drums. 32 refs., 61 figs., 33 tabs.

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

    SciTech Connect

    B. C. Rogers; P. L. Walter; R. D. Baird

    1999-08-01

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

  17. 78 FR 9747 - Request To Amend A License To Import; Radioactive Waste

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-11

    ... COMMISSION Request To Amend A License To Import; Radioactive Waste Pursuant to 10 CFR 110.70 (b) ``Public... radioactive Up to 378,000 Volume reduction...... Canada Services, Inc.; January 10, mixed waste kilograms... other contaminants, free release to ship including on shipping the Canadian waste to...

  18. Process Knowledge Characterization of Radioactive Waste at the Classified Waste Landfill Remediation Project Sandia National Laboratories, Albuquerque, New Mexico

    SciTech Connect

    DOTSON,PATRICK WELLS; GALLOWAY,ROBERT B.; JOHNSON JR,CARL EDWARD

    1999-11-03

    This paper discusses the development and application of process knowledge (PK) to the characterization of radioactive wastes generated during the excavation of buried materials at the Sandia National Laboratories/New Mexico (SNL/NM) Classified Waste Landfill (CWLF). The CWLF, located in SNL/NM Technical Area II, is a 1.5-acre site that received nuclear weapon components and related materials from about 1950 through 1987. These materials were used in the development and testing of nuclear weapon designs. The CWLF is being remediated by the SNL/NM Environmental Restoration (ER) Project pursuant to regulations of the New Mexico Environment Department. A goal of the CWLF project is to maximize the amount of excavated materials that can be demilitarized and recycled. However, some of these materials are radioactively contaminated and, if they cannot be decontaminated, are destined to require disposal as radioactive waste. Five major radioactive waste streams have been designated on the CWLF project, including: unclassified soft radioactive waste--consists of soft, compatible trash such as paper, plastic, and plywood; unclassified solid radioactive waste--includes scrap metal, other unclassified hardware items, and soil; unclassified mixed waste--contains the same materials as unclassified soft or solid radioactive waste, but also contains one or more Resource Conservation and Recovery Act (RCRA) constituents; classified radioactive waste--consists of classified artifacts, usually weapons components, that contain only radioactive contaminants; and classified mixed waste--comprises radioactive classified material that also contains RCRA constituents. These waste streams contain a variety of radionuclides that exist both as surface contamination and as sealed sources. To characterize these wastes, the CWLF project's waste management team is relying on data obtained from direct measurement of radionuclide activity content to the maximum extent possible and, in cases where

  19. Radioactive Waste Storage Facility at the Armenian NPP - 12462

    SciTech Connect

    Grigoryan, G.; Amirjanyan, A.; Gondakyan, Y.; Stepanyan, A.

    2012-07-01

    We present a detailed contaminant transfer dynamics model for radionuclide in geosphere and biosphere medium. The model describes the transport of radionuclides using full equation for the processes of advection, diffusion, decay and sorption. The overall objective is to establish, from a post-closure radiological safety point of view, whether it is practical to convert an existing radioactive waste storage facility at Armenian NPP, to a waste disposal facility. The calculation includes: - Data sources for: the operational waste-source term; options for refurbishment and completion of the waste storage facility as a waste disposal facility; the site and its environs; - Development of an assessment context for the safety assessment, and identification of waste treatment options; - A description of the conceptual and mathematical models, and results calculated for the base case scenario relating to the release of contaminants via the groundwater pathway and also precipitation especially important for this site. The results of the calculations showed that the peak individual dose is < 7 E-8 Sv/y arising principally from I-129 after 700 years post closure. Other significant radionuclides, in terms of their contribution to the total dose are I-129, Tc-99 and in little C-14 (U- 234 and Po-210 are not relevant). The study does not explore all issues that might be expected to be presented in a safety case for a near surface disposal facility it mainly focuses on post- closure dose impacts. Most emphasis has been placed on the development of scenarios and conceptual models rather than the presentation and analyses of results and confidence building (only deterministic results are presented). The calculations suggest that, from a perspective the conversion of the waste-storage facility is feasible such that all the predicted doses are well below internationally recognized targets, as well as provisional Armenian regulatory objectives. This conclusion applies to the disposal

  20. RADIOACTIVE WASTE MANAGEMENT IN THE USSR: A REVIEW OF UNCLASSIFIED SOURCES, 1963-1990

    SciTech Connect

    Bradley, D. J.; Schneider, K. J.

    1990-03-01

    volume of 100-150 liters/MT of fuel reprocessed); 3) set it aside in air-cooled storage for 30-50 years; and 4} provide for its final disposal in a deep geologic repository. High-level waste solidification research started in the 1950s, and a pilot vitrification facility has operated for 10 years at Kyshtym. A fully radioactive unit also operated there for about two years until 1988 and produced 160 MT of HLW phosphate glass, containing 3.9 x 106 curies of radioactivity. [This corresponds to radioactivity in about 10 MT of 10-year-old power reactor fuel.] Problems with the glass melter electrodes have caused this latter plant to shut down, but a new design and plant are being developed. A site near Krasnoyarsk has also recently been referred to as a disposal site for low- and intermediate-level waste streams and possibly for vitrified wastes from Kyshtym. Deep geologic disposal of vitrified high-level waste has been studied, at a modest level of effort, over the last decade in the Soviet Union. Key disposal concepts under consideration include mining shafts, deep drill holes, and underground excavations. Although a variety of host rock types have been addressed, the emphasis appears to be on salt, clay, and granite. The Soviet Union appears to require the use of man-made barriers and waste packages in permanent disposal, but the geologic formation will be relied upon as the primary protective barrier. The construction schedule for a Soviet repository has not been stated. Meanwhile, a decision on the first repository site may be made in 1990, with a candidate site being at Chelyabinsk.

  1. The Murmansk low-level liquid radioactive waste treatment facility

    SciTech Connect

    Duffey, R.B.; Penzin, R.A.; Tumparov, A.; Gussgard, K.; Dyer, R.A.; Ruksha, V.V.

    1996-09-01

    Since May 1994, Russian, Norway and the US have cooperated successfully to develop the design of the LLRW treatment facility. Among the other participating organizations are the Association for Advanced Technologies, the Royal norwegian Ministry of Foreign Affairs and the US Environmental Protection Agency (EPA). The joint US/Norwegian/Russian Technical Team for the Design provide technical review and advice. A major objective of the design is to enable Russian to permanently cease disposing of this waste in Arctic waters and to formally adhere to the London Convention permanent ban. Therefore, the modifications will increase the facility`s capacity from 1,200 m{sup 3} per year to 5,000 m{sup 3} per year, will permit the facility to process high-salt wastes from the Russian Navy`s North Fleet and will improve the stabilization and interim storage of the processed wastes. The plant design utilizes novel technology for the filtration and treatment of the high saline waste streams, as well as waste streams for maintenance and decontamination activities. The discharge streams will meet national and international discharge limits for radionuclides. Following detailed design, the next phase is construction.

  2. Bonding material containing ashes after domestic waste incineration for cementation of radioactive waste

    SciTech Connect

    Dmitriev, S.A.; Varlakov, A.P.; Gorbunova, O.A.; Arustamov, A.E.; Barinov, A.S.

    2007-07-01

    It is known that cement minerals hydration is accompanied with heat emission. Heat of hardening influences formation of a cement compound structure and its properties. It is important to reduce the heat quantity at continuous cementation of waste and filling of compartments of a repository or containers by a cement grout. For reduction of heating, it is necessary to use cement of mineral additives (fuel ashes, slag and hydraulic silica). Properties of ashes after domestic waste incineration can be similar to ones of fly fuel ashes. However, ash after domestic waste incineration is toxic industrial waste as it contains toxic elements (As, Cd, Hg, Pb, Sb, Zn). Utilization of secondary waste (slag and ash) of combustion plants is an important environmental approach to solving cities' issues. Results of the research have shown that ashes of combustion plants can be used for radioactive waste conditioning. Co-processing of toxic and radioactive waste is ecologically and economically effective. At SIA 'Radon', experimental batches of cement compositions are used for cementation of oil containing waste. (authors)

  3. Magnetic nano-sorbents for fast separation of radioactive waste

    SciTech Connect

    Zhang, Huijin; Kaur, Maninder; Qiang, You

    2013-07-01

    In order to find a cost effective and environmentally benign technology to treat the liquid radioactive waste into a safe and stable form for resource recycling or ultimate disposal, this study investigates the separation of radioactive elements from aqueous systems using magnetic nano-sorbents. Our current study focuses on novel magnetic nano-sorbents by attaching DTPA molecules onto the surface of double coated magnetic nanoparticles (dMNPs), and performed preliminary sorption tests using heavy metal ions as surrogates for radionuclides. The results showed that the sorption of cadmium (Cd) and lead (Pb) onto the dMNP-DTPA conjugates was fast, the equilibrium was reached in 30 min. The calculated sorption capacities were 8.06 mg/g for Cd and 12.09 mg/g for Pb. After sorption, the complex of heavy elements captured by nano-sorbents can be easily manipulated and separated from solution in less than 1 min by applying a small external magnetic field. In addition, the sorption results demonstrate that dMNP-DTPA conjugates have a very strong chelating power in highly diluted Cd and Pb solutions (1-10 μg/L). Therefore, as a simple, fast, and compact process, this separation method has a great potential in the treatment of high level waste with low concentration of transuranic elements compared to tradition nuclear waste treatment. (authors)

  4. 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.

  5. Regulatory Approaches for Solid Radioactive Waste Storage in Russia

    SciTech Connect

    Griffith, A.; Testov, S.; Diaschev, A.; Nazarian, A.; Ustyuzhanin, A.

    2003-02-26

    The Russian Navy under the Arctic Military Environmental Cooperation (AMEC) Program has designated the Polyarninsky Shipyard as the regional recipient for solid radioactive waste (SRW) pretreatment and storage facilities. Waste storage technologies include containers and lightweight modular storage buildings. The prime focus of this paper is solid radioactive waste storage options based on the AMEC mission and Russian regulatory standards. The storage capability at the Polyarninsky Shipyard in support of Mobile Pretreatment Facility (MPF) operations under the AMEC Program will allow the Russian Navy to accumulate/stage the SRW after treatment at the MPF. It is anticipated that the MPF will operate for 20 years. This paper presents the results of a regulatory analysis performed to support an AMEC program decision on the type of facility to be used for storage of SRW. The objectives the study were to: analyze whether a modular storage building (MSB), referred in the standards as a lightweight building, would comply with the Russian SRW storage building standard, OST 95 10517-95; analyze the Russian SRW storage pad standard OST 95 10516-95; and compare the two standards, OST 95 10517-95 for storage buildings and OST 95 10516-95 for storage pads.

  6. Fifty years of federal radioactive waste management: Policies and practices

    SciTech Connect

    Bradley, R.G.

    1997-04-01

    This report provides a chronological history of policies and practices relating to the management of radioactive waste for which the US Atomic Energy Commission and its successor agencies, the Energy Research and Development Administration and the Department of Energy, have been responsible since the enactment of the Atomic Energy Act in 1946. The defense programs and capabilities that the Commission inherited in 1947 are briefly described. The Commission undertook a dramatic expansion nationwide of its physical facilities and program capabilities over the five years beginning in 1947. While the nuclear defense activities continued to be a major portion of the Atomic Energy Commission`s program, there was added in 1955 the Atoms for Peace program that spawned a multiplicity of peaceful use applications for nuclear energy, e.g., the civilian nuclear power program and its associated nuclear fuel cycle; a variety of industrial applications; and medical research, diagnostic, and therapeutic applications. All of these nuclear programs and activities generated large volumes of radioactive waste that had to be managed in a manner that was safe for the workers, the public, and the environment. The management of these materials, which varied significantly in their physical, chemical, and radiological characteristics, involved to varying degrees the following phases of the waste management system life cycle: waste characterization, storage, treatment, and disposal, with appropriate transportation linkages. One of the benefits of reviewing the history of the waste management program policies and practices if the opportunity it provides for identifying the lessons learned over the years. Examples are summarized at the end of the report and are listed in no particular order of importance.

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

    SciTech Connect

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

    2009-07-01

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

  8. [A system for decontamination of liquid radioactive waste produced in in vitro tests in nuclear medicine].

    PubMed

    Tsuchiya, T; Norimura, T; Ueno, T

    1983-06-01

    It is well known that very large storage tanks for radioactive liquids are necessary for the disposal of liquid radioactive waste. In vitro tests in radioimmunoassay in nuclear medicine are rapidly increasing for clinical examination causing marked increase in the volume of liquid radioactive waste. Thus we have developed a system for decontaminating radioactivity from liquid waste. In the first step, the liquid waste is boiled by a sterilizer and, in the second step, this sterilised liquid is filtered by a cylindrical filter (Toyo filter No. 84). After filtration, the liquid waste is passed into a beaded charcoal column and an ion exchange resin (Amberlite IRA 402) column. After these treatments, the radioactivity level of liquid waste is lowered to less than 1% of the original radioactivity. We are now in the planning stages of building an apparatus for practical use. PMID:6622764

  9. Method for acid oxidation of radioactive, hazardous, and mixed organic waste materials

    DOEpatents

    Pierce, Robert A.; Smith, James R.; Ramsey, William G.; Cicero-Herman, Connie A.; Bickford, Dennis F.

    1999-01-01

    The present invention is directed to a process for reducing the volume of low level radioactive and mixed waste to enable the waste to be more economically stored in a suitable repository, and for placing the waste into a form suitable for permanent disposal. The invention involves a process for preparing radioactive, hazardous, or mixed waste for storage by contacting the waste starting material containing at least one organic carbon-containing compound and at least one radioactive or hazardous waste component with nitric acid and phosphoric acid simultaneously at a contacting temperature in the range of about 140.degree. C. to about 210 .degree. C. for a period of time sufficient to oxidize at least a portion of the organic carbon-containing compound to gaseous products, thereby producing a residual concentrated waste product containing substantially all of said radioactive or inorganic hazardous waste component; and immobilizing the residual concentrated waste product in a solid phosphate-based ceramic or glass form.

  10. Method for acid oxidation of radioactive, hazardous, and mixed organic waste materials

    SciTech Connect

    Pierce, R.A.; Smith, J.R.; Ramsey, W.G.; Cicero-Herman, C.A.; Bickford, D.F.

    1999-09-28

    The present invention is directed to a process for reducing the volume of low level radioactive and mixed waste to enable the waste to be more economically stored in a suitable repository, and for placing the waste into a form suitable for permanent disposal. The invention involves a process for preparing radioactive, hazardous, or mixed waste for storage by contacting the waste starting material containing at least one organic carbon-containing compound and at least one radioactive or hazardous waste component with nitric acid and phosphoric acid simultaneously at a contacting temperature in the range of about 140 C to about 210 C for a period of time sufficient to oxidize at least a portion of the organic carbon-containing compound to gaseous products, thereby producing a residual concentrated waste product containing substantially all of said radioactive or inorganic hazardous waste component; and immobilizing the residual concentrated waste product in a solid phosphate-based ceramic or glass form.

  11. Electrodialysis-based separation process for salt recovery and recycling from waste water

    DOEpatents

    Tsai, Shih-Perng

    1997-01-01

    A method for recovering salt from a process stream containing organic contaminants is provided, comprising directing the waste stream to a desalting electrodialysis unit so as to create a concentrated and purified salt permeate and an organic contaminants containing stream, and contacting said concentrated salt permeate to a water-splitting electrodialysis unit so as to convert the salt to its corresponding base and acid.

  12. Electrodialysis-based separation process for salt recovery and recycling from waste water

    DOEpatents

    Tsai, S.P.

    1997-07-08

    A method for recovering salt from a process stream containing organic contaminants is provided, comprising directing the waste stream to a desalting electrodialysis unit so as to create a concentrated and purified salt permeate and an organic contaminants-containing stream, and contacting said concentrated salt permeate to a water-splitting electrodialysis unit so as to convert the salt to its corresponding base and acid. 6 figs.

  13. FINAL DISPOSAL OF RADIOACTIVE WASTE IN GERMANY: PLAN APPROVAL PROCESS OF KONRAD MINE AND ACCEPTANCE REQUIREMENTS

    SciTech Connect

    Bandt, Gabriele; Posnatzki, Britta; Beckers, Klaus-Arno

    2003-02-27

    Currently no final repository for any type of radioactive waste is operated in Germany. Preliminary Final Storage Acceptance Requirements for radioactive waste packages were published in 1995. Up to now these are the basis for treatment of radioactive waste in Germany. After licensing of the final repository these preliminary waste acceptance requirements are completed with licensing conditions. Some of these conditions affect the preliminary waste acceptance requirements, e. g. behavior of chemo-toxic substances in case of accidents in the final repository or the allowed maximum concentration of fissile material. The presented examples of radioactive waste conditioning campaigns demonstrate that no difficulties are expected in management, characterization and quality assurance of radioactive wastes due to the licensing conditions.

  14. Vitrification of radioactive waste by reaction sintering under pressure

    NASA Astrophysics Data System (ADS)

    Gong, W. L.; Lutze, W.; Abdelouas, A.; Ewing, R. C.

    1999-02-01

    Silicate nuclear waste glasses were synthesized by reaction sintering of powdered precursors under pressure. The glass samples contained a glass matrix phase with embedded zirconia (baddeleyite) particles. A waste composition with 38 wt% of ZrO 2 was prepared with a waste loading of 30-50 wt% at 800°C and 28 MPa, by hot isostatic pressing. The glass former was commercial amorphous silica powder to which simulated waste was added as calcined oxides. Phase compositions and microstructure of the sintered glass samples were characterized using scanning and analytical electron microscopy. The results show that extensive sintering took place and that a continuous glass phase was formed, particularly at higher waste loading. Waste components such as Na 2O, CaO, MnO 2, La 2O 3, Fe 2O 3, Cr 2O 3, and P 2O 5 dissolved completely in the glass phase. ZrO 2 was also dissolved but recrystallized from the glass as aggregates of baddeleyite crystallites surrounding the original silica particles. MCC-1 type chemical durability tests showed that the glasses are durable with dissolution rates similar to or lower than that of the highly durable French R7T7 borosilicate glass. This glass contains 13 wt% high-level radioactive waste from light water reactor fuel reprocessing and has a melting temperature of 1150°C. The long-term chemical durability of our sintered glasses is expected to be as high as that of rhyolitic glasses, based on hydration energies of 3.7 and 3.3 kJ/mole, respectively. Rhyolitic glasses show little alteration over geological periods of time with a typical corrosion rate of 1 μm/1000 yr.

  15. Naturally occurring crystalline phases: analogues for radioactive waste forms

    SciTech Connect

    Haaker, R.F.; Ewing, R.C.

    1981-01-01

    Naturally occurring mineral analogues to crystalline phases that are constituents of crystalline radioactive waste forms provide a basis for comparison by which the long-term stability of these phases may be estimated. The crystal structures and the crystal chemistry of the following natural analogues are presented: baddeleyite, hematite, nepheline; pollucite, scheelite;sodalite, spinel, apatite, monazite, uraninite, hollandite-priderite, perovskite, and zirconolite. For each phase in geochemistry, occurrence, alteration and radiation effects are described. A selected bibliography for each phase is included.

  16. Disposal of liquid radioactive wastes through wells or shafts

    SciTech Connect

    Perkins, B.L.

    1982-01-01

    This report describes disposal of liquids and, in some cases, suitable solids and/or entrapped gases, through: (1) well injection into deep permeable strata, bounded by impermeable layers; (2) grout injection into an impermeable host rock, forming fractures in which the waste solidifies; and (3) slurrying into excavated subsurface cavities. Radioactive materials are presently being disposed of worldwide using all three techniques. However, it would appear that if the techniques were verified as posing minimum hazards to the environment and suitable site-specific host rock were identified, these disposal techniques could be more widely used.

  17. Disposal of liquid radioactive wastes through wells or shafts

    NASA Astrophysics Data System (ADS)

    Perkins, B. L.

    1982-01-01

    Liquids and, in some cases, suitable solids and/or entrapped gases can be disposed of by: (1) well injection into deep permeable strata, bounded by impermeable layers; (2) grout injection into an impermeable host rock, forming fractures in which the waste solidifies; and (3) slurrying into excavated subsurface cavities. Radioactive materials are presently being disposed of worldwide using all three techniques; however, it would appear that if the techniques were verified as posing minimum hazards to the environment and suitable site-specific host rock were identified, these disposal techniques could be more widely used.

  18. Vapor sampling of the headspace of radioactive waste storage tanks

    SciTech Connect

    Reynolds, D.A., Westinghouse Hanford

    1996-05-22

    This paper recants the history of vapor sampling in the headspaces of radioactive waste storage tanks at Hanford. The first two tanks to receive extensive vapor pressure sampling were Tanks 241-SY-101 and 241-C-103. At various times, a gas chromatography, on-line mass spectrometer, solid state hydrogen monitor, FTIR, and radio acoustic ammonia monitor have been installed. The head space gas sampling activities will continue for the next few years. The current goal is to sample the headspace for all the tanks. Some tank headspaces will be sampled several times to see the data vary with time. Other tanks will have continuous monitors installed to provide additional data.

  19. Design of microwave vitrification systems for radioactive waste

    SciTech Connect

    White, T.L.; Wilson, C.T.; Schaich, C.R.; Bostick, T.L.

    1995-12-31

    Oak Ridge National Laboratory (ORNL) is involved in the research and development of high-power microwave heating systems for the vitrification of Department of Energy (DOE) radioactive sludges. Design criteria for a continuous microwave vitrification system capable of processing a surrogate filtercake sludge representative of a typical waste-water treatment operation are discussed. A prototype 915-MHz, 75-kW microwave vitrification system or ``microwave melter`` is described along with some early experimental results that demonstrate a 4 to 1 volume reduction of a surrogate ORNL filtercake sludge.

  20. Nondestructive examination technologies for inspection of radioactive waste storage tanks

    SciTech Connect

    Anderson, M.T.; Kunerth, D.C.; Davidson, J.R.

    1995-08-01

    The evaluation of underground radioactive waste storage tank structural integrity poses a unique set of challenges. Radiation fields, limited access, personnel safety and internal structures are just some of the problems faced. To examine the internal surfaces a sensor suite must be deployed as an end effector on a robotic arm. The purpose of this report is to examine the potential failure modes of the tanks, rank the viability of various NDE technologies for internal surface evaluation, select a technology for initial EE implementation, and project future needs for NDE EE sensor suites.

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

    SciTech Connect

    1996-06-01

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

  2. Stability High Salt Content Waste Using Sol Gel Process. Mixed Waste Focus Area. OST Reference Number 0236

    SciTech Connect

    None, None

    1999-09-01

    Mixed waste sludges, soils, and homogeneous solids containing high levels of salt ( ~ greater than 15% by weight ) have proven to be difficult to stabilize due to the soluble nature of the salts. The current stabilization technique for high salt waste, grouting with Portland cement, is limited to low waste loadings. The presence of salts interfere with the hydration and curing of the cement, cause waste form deteriorating mineral expansions, or result in an undesirable separate phase altogether. Improved technologies for the stabilization of salt waste must be able to accommodate higher salt loadings, while maintaining structural integrity, chemical durability, and leach resistance. In a joint collaboration supported by the Department of Energy’s (DOE’s) Mixed Waste Focus Area (MWFA), the Pacific Northwest National Laboratory (PNNL) and the Arizona Materials Laboratory (AML) at the University of Arizona have developed a sol-gel (wet-chemical) based, low-temperature-processing route for the stabilization of salt-containing mixed wastes. By blending and reacting liquid precursors at room temperature with salt waste, strong, impermeable “polyceram” matrices have been formed that encapsulate the environmentally hazardous waste components. As depicted by Figure 1, polycerams are hybrid organic/inorganic materials with unique properties derived from the chemical combination of polymer (organic) and ceramic (inorganic) components. For this application, the stabilizing polyceram matrices contain polybutadiene-based polymer components and silicon dioxide (SiO2) as the inorganic component. Polybutadiene (PBD) is a strong, tough, waterresistant plastic and its use in the polyceram promotes these same characteristics in the waste form. The PBD polymer component is modified to increase its reactivity with the SiO2 precursor during sol-gel processing. When combined, the polymer and SiO2 precursors react, gel, solidify, and encapsulate the

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

    SciTech Connect

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

    1994-04-01

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

  4. Investigation of Shielding Material in Radioactive Waste Management - 13009

    SciTech Connect

    OSMANLIOGLU, Ahmet Erdal

    2013-07-01

    In this study, various waste packages have been prepared by using different materials. Experimental work has been performed on radiation shielding for gamma and neutron radiation. Various materials were evaluated (e.g. concrete, boron, etc.) related to different application areas in radioactive waste management. Effects of addition boric compound mixtures on shielding properties of concrete have been investigated for neutron radiation. The effect of the mixture addition on the shielding properties of concrete was investigated. The results show that negative effects of boric compounds on the strength of concrete decreasing by increasing boric amounts. Shielding efficiency of prepared mixture added concrete up to 80% better than ordinary concretes for neutron radiation. The attenuation was determined theoretically by calculation and practically by using neutron dose rate measurements. In addition of dose rate measurements, strength tests were applied on test shielding materials. (authors)

  5. Radioactive waste disposal characteristics of candidate tokamak demonstration reactors

    SciTech Connect

    Hoffman, E.A.; Stacey, W.M.; Hertel, N.E.

    1998-08-01

    Results from the current physics, materials and blanket R and D programs are combined with physics and engineering design constraints to characterize candidate tokamak demonstration plant (DEMO) designs. Blanket designs based on the principal structural materials, breeding materials and coolants being developed for the DEMO were adapted from the literature. Neutron flux and activation calculations were performed, and several radioactive waste disposal indices were evaluated, for each design. Of the primary low-activation structural materials under development in the US, it appears that vanadium and ferritic steel alloys, and possibly silicon carbide, could lead to DEMO designs which could satisfy realistic low-level waste (LLW) criteria, provided that impurities can be controlled within plausible limits. Allowable LLW concentrations are established for the limiting alloying and impurity elements. All breeding materials and neutron multipliers considered meet the LLW criterion.

  6. On-line remote monitoring of radioactive waste repositories

    NASA Astrophysics Data System (ADS)

    Calì, Claudio; Cosentino, Luigi; Litrico, Pietro; Pappalardo, Alfio; Scirè, Carlotta; Scirè, Sergio; Vecchio, Gianfranco; Finocchiaro, Paolo; Alfieri, Severino; Mariani, Annamaria

    2014-12-01

    A low-cost array of modular sensors for online monitoring of radioactive waste was developed at INFN-LNS. We implemented a new kind of gamma counter, based on Silicon PhotoMultipliers and scintillating fibers, that behaves like a cheap scintillating Geiger-Muller counter. It can be placed in shape of a fine grid around each single waste drum in a repository. Front-end electronics and an FPGA-based counting system were developed to handle the field data, also implementing data transmission, a graphical user interface and a data storage system. A test of four sensors in a real radwaste storage site was performed with promising results. Following the tests an agreement was signed between INFN and Sogin for the joint development and installation of a prototype DMNR (Detector Mesh for Nuclear Repository) system inside the Garigliano radwaste repository in Sessa Aurunca (CE, Italy). Such a development is currently under way, with the installation foreseen within 2014.

  7. Management of radioactive waste from nuclear power plants

    SciTech Connect

    Not Available

    1993-09-01

    Even thought risk assessment is an essential consideration in all projects involving radioactive or hazardous waste, its public role is often unclear, and it is not fully utilized in the decision-making process for public acceptance of such facilities. Risk assessment should be an integral part of such projects and should play an important role from beginning to end, i.e., from planning stages to the closing of a disposal facility. A conceptual model that incorporates all potential pathways of exposure and is based on site-specific conditions is key to a successful risk assessment. A baseline comparison with existing standards determines, along with other factors, whether the disposal site is safe. Risk assessment also plays a role in setting priorities between sites during cleanup actions and in setting cleanup standards for certain contaminants at a site. The applicable technologies and waste disposal designs can be screened through risk assessment.

  8. Midwestern High-Level Radioactive Waste Transportation Project

    SciTech Connect

    Dantoin, T.S.

    1990-12-01

    For more than half a century, the Council of State Governments has served as a common ground for the states of the nation. The Council is a nonprofit, state-supported and -directed service organization that provides research and resources, identifies trends, supplies answers and creates a network for legislative, executive and judicial branch representatives. This List of Available Resources was prepared with the support of the US Department of Energy, Cooperative Agreement No. DE-FC02-89CH10402. However, any opinions, findings, conclusions, or recommendations expressed herein are those of the author(s) and do not necessarily reflect the views of DOE. The purpose of the agreement, and reports issued pursuant to it, is to identify and analyze regional issues pertaining to the transportation of high-level radioactive waste and to inform Midwestern state officials with respect to technical issues and regulatory concerns related to waste transportation.

  9. Update on Radioactive Waste Management in the UK

    SciTech Connect

    Dalton, John; McCall, Ann

    2003-02-24

    This paper provides a brief background to the current position in the United Kingdom (UK) and provides an update on the various developments and initiatives within the field of radioactive waste management that have been taking place during 2002/03. These include: The UK Government's Department of Trade and Industry (DTi) review of UK energy policy; The UK Government's (Department of Environment, Food and Rural Affairs (Defra) and Devolved Administrations*) consultation program; The UK Government's DTi White Paper, 'Managing the Nuclear Legacy: A Strategy for Action'; Proposals for improved regulation of Intermediate Level Waste (ILW) conditioning and packaging. These various initiatives relate, in Nirex's opinion, to the three sectors of the industry and this paper will provide a comment on these initiatives in light of the lessons that Nirex has learnt from past events and suggest some conclusions for the future.

  10. Site characterization for LIL radioactive waste disposal in Romania

    SciTech Connect

    Diaconu, D. R.; Birdsell, K. H.; Witkowski, M. S.

    2001-01-01

    Recent studies in radioactive waste management in Romania have focussed mainly on the disposal of low and intermediate level waste from the operation of the new nuclear power plant at Cernavoda. Following extensive geological, hydrological, seismological, physical and chemical investigations, a disposal site at Saligny has been selected. This paper presents description of the site at Saligny as well as the most important results of the site characterisation. These are reflected in the three-dimensional, stratigraphical representation of the loess and clay layers and in representative parameter values for the main layers. Based on these data, the simulation of the background, unsaturated-zone water flow at the Saligny site, calculated by the FEHM code, is in a good agreement with the measured moisture profile.

  11. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  12. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  13. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  14. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  15. Preliminary technical and legal evaluation of disposing of nonhazardous oil field waste into salt caverns

    SciTech Connect

    Veil, J.; Elcock, D.; Raivel, M.; Caudle, D.; Ayers, R.C. Jr.; Grunewald, B.

    1996-06-01

    Caverns can be readily formed in salt formations through solution mining. The caverns may be formed incidentally, as a result of salt recovery, or intentionally to create an underground chamber that can be used for storing hydrocarbon products or compressed air or disposing of wastes. The purpose of this report is to evaluate the feasibility, suitability, and legality of disposing of nonhazardous oil and gas exploration, development, and production wastes (hereafter referred to as oil field wastes, unless otherwise noted) in salt caverns. Chapter 2 provides background information on: types and locations of US subsurface salt deposits; basic solution mining techniques used to create caverns; and ways in which salt caverns are used. Later chapters provide discussion of: federal and state regulatory requirements concerning disposal of oil field waste, including which wastes are considered eligible for cavern disposal; waste streams that are considered to be oil field waste; and an evaluation of technical issues concerning the suitability of using salt caverns for disposing of oil field waste. Separate chapters present: types of oil field wastes suitable for cavern disposal; cavern design and location; disposal operations; and closure and remediation. This report does not suggest specific numerical limits for such factors or variables as distance to neighboring activities, depths for casings, pressure testing, or size and shape of cavern. The intent is to raise issues and general approaches that will contribute to the growing body of information on this subject.

  16. Disposal of oil field wastes into salt caverns: Feasibility, legality, risk, and costs

    SciTech Connect

    Veil, J.A.

    1997-10-01

    Salt caverns can be formed through solution mining in the bedded or domal salt formations that are found in many states. Salt caverns have traditionally been used for hydrocarbon storage, but caverns have also been used to dispose of some types of wastes. This paper provides an overview of several years of research by Argonne National Laboratory on the feasibility and legality of using salt caverns for disposing of oil field wastes, the risks to human populations from this disposal method, and the cost of cavern disposal. Costs are compared between the four operating US disposal caverns and other commercial disposal options located in the same geographic area as the caverns. Argonne`s research indicates that disposal of oil field wastes into salt caverns is feasible and legal. The risk from cavern disposal of oil field wastes appears to be below accepted safe risk thresholds. Disposal caverns are economically competitive with other disposal options.

  17. Expedited demonstration of molten salt mixed waste treatment technology. Final report

    SciTech Connect

    1995-02-02

    This final report discusses the molten salt mixed waste project in terms of the various subtasks established. Subtask 1: Carbon monoxide emissions; Establish a salt recycle schedule and/or a strategy for off-gas control for MWMF that keeps carbon monoxide emission below 100 ppm on an hourly averaged basis. Subtask 2: Salt melt viscosity; Experiments are conducted to determine salt viscosity as a function of ash composition, ash concentration, temperature, and time. Subtask 3: Determine that the amount of sodium carbonate entrained in the off-gas is minimal, and that any deposited salt can easily be removed form the piping using a soot blower or other means. Subtask 4: The provision of at least one final waste form that meets the waste acceptance criteria of a landfill that will take the waste. This report discusses the progress made in each of these areas.

  18. Management of low-level radioactive wastes around the world

    SciTech Connect

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

    1985-04-01

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

  19. Impact of radioactive waste heat on soil temperatures

    SciTech Connect

    Robinson, B.A.; Gable, C.W.; Lowman, J.P.

    1999-01-04

    Consideration of the impact of radioactive waste heat is necessary for many aspects of potential repository design. Waste heat will alter the mineralogy of the host rock, and may change the character of the zeolitic units below the potential repository that are likely to be the primary natural barriers to radionuclide migration. The impact of waste heat on the near-surface temperature within the soil zone is the focus of the present study. Since 1990, the Nuclear Waste Technical Review Board (NWTRB) has raised the issue of potential impacts on the aboveground ecosystem from increases in soil temperatures. This study is a first step toward understanding the relevant heat transfer processes that controls the near-surface thermal regime and to place bounds on the expected timing and magnitude of the temperature rise. Two-dimensional, site scale thermohydrologic calculations will be used to simulate the large-scale thermohydrologic processes that will feed heat to the soil zone. The potential influence of this heat on soil-zone temperatures will then be examined in a series of simplified one-dimensional model calculations. In future efforts the measured soil-zone temperature variations in the air will be used to calibrate the model, which will tighten the bounds on the possible temperature rise. This study is a precursor to more detailed, three-dimensional simulations with a calibrated model. If it is determined that direct coupling of the site scale and soil zones would be beneficial, this will be done as well.

  20. Office of Civilian Radioactive Waste Management annual report to Congress

    SciTech Connect

    1988-08-01

    This is the fifth Annual Report to Congress by the Office of Civilian Radioactive Waste Management (OCRWM). The report covers the activities and expenditures of OCRWM during fiscal year 1987, which ended on September 30, 1987. The activities and accomplishments of OCRWM during fiscal year 1987 are discussed in chapters 1 through 9 of this report. The audited financial statements of the Nuclear Waste Fund are provided in chapter 10. Since the close of the fiscal year, a number of significant events have occurred. Foremost among them was the passage of the Nuclear Waste Policy Amendments Act of 1987 (Amendments Act) on December 21, 1987, nearly 3 months after the end of the fiscal year covered by this report. As a result, some of the plans and activities discussed in chapters 1 through 9 are currently undergoing significant change or are being discontinued. Most prominent among the provisions of the Amendments Act is the designation of Yucca Mountain, Nevada, as the only candidate first repository site to be characterized. Therefore, the site characterization plans for Deaf Smith, Texas, and Hanford, Washington, discussed in chapter 3, will not be issued. The refocusing of the waste management program under the Amendments Act is highlighted in the epilogue, chapter 11. 68 refs., 7 figs., 7 tabs.