WorldWideScience

Sample records for high activity wastes

  1. High activity waste disposal

    International Nuclear Information System (INIS)

    Gaul, W.C.

    1990-01-01

    Chem-Nuclear Environmental Services (CNES) has developed a container that is capable of containing high activity waste and can be shipped as a regular DOT Type A shipment. By making the container special form the amount of activity that can be transported in a Type A shipment is greatly enhanced. Special form material presents an extra degree of protection to the environment by requiring the package to be destroyed to get access to the radioactive material and must undergo specific testing requirements, whereas normal form material can allow access to the radioactive material. With the special form container up to 10 caries of radium can be transported in a single package. This paper will describe the considerations that were taken to develop these products

  2. Solidification of highly active wastes

    International Nuclear Information System (INIS)

    Morris, J.B.

    1986-07-01

    This document contains the annual reports for the contracts: (A) Glass Technology; (B) Calcination of Highly Active Waste Liquors; (C) Formation and Trapping of Volatile Ruthenium; (D) Deposition of Ruthenium; (E) Enhancement of Off-Gas Aerosol Collection; (F) Volatilisation of Cs, Tc and Te in High Level Waste Vitrification. (author)

  3. Solidification of highly active wastes

    International Nuclear Information System (INIS)

    Morris, J.B.

    1984-11-01

    Final reports are presented on work on the following topics: glass technology; enhancement of off-gas aerosol collection; formation and trapping of volatile ruthenium; volatilisation of caesium, technetium and tellurium in high-level waste vitrification; deposition of ruthenium; and calcination of high-level waste liquors. (author)

  4. Disposal of high-activity nuclear wastes

    International Nuclear Information System (INIS)

    Hamilton, E.I.

    1983-01-01

    A discussion is presented on the deep sea ocean disposal for high-activity nuclear wastes. The following topics are covered: effect of ionizing radiation on marine ecosystems; pathways by which radionuclides are transferred to man from the marine environment; information about releases of radioactivity to the sea; radiological protection; storage and disposal of radioactive wastes and information needs. (U.K.)

  5. Underground disposal of high active waste

    International Nuclear Information System (INIS)

    Engelmann, H.J.

    1982-01-01

    This paper is concerned with the engineering aspects relating to the deep burial of high active waste in stable geological formations. The design of a repository depends upon a number of factors not least of which is the type of rock in which it is to be constructed. High level wastes must be isolated from man's environment for such periods that subsequent release will not result in an unacceptable hazard to human population. Design aspects of repositories are reviewed and conceptual design are present in relation to the geological formations under consideration. Over long time periods the most probable mode of release of radionuclides is through groundwater contacting the waste. The proposed concepts therefore include the use of engineered and natural barriers to delay the eventual release of waterborne radionuclides into mans environment. In all cases the ultimate barrier will be the geological formation. Nevertheless, depending upon the type of host rock, use will be made of various additional engineered barriers to delay water contacting the high level waste for several hundreds of years. During this time the level of radiation and associated heat emitted by the waste, will fall by several orders of magnitude and the rock temperatures within a repository will be returning to ambient. Thereafter the residual activity will mainly arise from the actinides. Containment may be enhanced by surrounding the canisters with materials having high sorption capabilities for many of the radionuclides involved. The depth at which a repository is excavated must be sufficient to ensure that the overburden will withstand changes in environmental conditions. The depth of cover required in different rock types may vary. In clay excavating at depth of up to -250 m appears feasible, while in hard rocks and salts working at depth of up to -1000 m is entirely practicable. (orig./RW)

  6. High-active waste (HAW) data report

    International Nuclear Information System (INIS)

    Duijves, K.A.

    1991-06-01

    Data are presented from the High Active Waste (HAW) experiment, a large-scale, in situ test being performed underground at the Asse salt mine in Remlingen, FRG. These data include selected field information, the test configuration, instrumentation activities and comprehensive results from a large number of gauges. The results are measured data obtained from gap meters, thermocouples, linear displacement trans-ducers, extensometers, inclinometers and pressure gauges. Data certification practices have been described together with the quality assurance of the data reduction and of the data base management system. The experiment began on November 8, 1988 and will continue for five years. Data in this report cover the period from July 1st, 1990 to December 31, 1990. (author). 4 refs.; 100 figs.; 8 tabs

  7. Conversion of highly active waste to solids

    International Nuclear Information System (INIS)

    Scheffler, K.

    Borosilicate glasses were selected as matrix material for solidification of highly radioactive wastes. Current laboratory work on the VERA process is described. Goals were met by a five-component glass VG-38 and a glass-ceramic VC-15. The VERA process is described: flowsheet, denitration, calcinator, fusion facility

  8. Development plan. High activity-long living wastes project. Abstract

    International Nuclear Information System (INIS)

    2007-01-01

    This brochure presents the actions that the ANDRA (the French national agency of radioactive wastes) has to implement in the framework of the project of high activity-long living (HALL) radioactive wastes (HAVL project) conformably to the requirements of the program defined in the law from June 28, 2006 (law no 2006-739). This law precises the three, complementary, research paths to explore for the management of this type of wastes: separation and transmutation of long-living radioactive elements, reversible disposal in deep geologic underground, and long duration storage. The ANDRA's action concerns the geologic disposal aspect. The following points are presented: the HALL wastes and their containers, the reversible disposal procedure, the HAVL project: financing of researches, storage concepts, development plan of the project (dynamics, information and dialogue approach, input data, main steps, schedule); the nine programs of the HAVL project (laboratory experiments and demonstration tests, surface survey, scientific program, simulation program, surface engineering studies and technological tests, information and communication program, program of environment and facilities surface observation and monitoring, waste packages management, monitoring and transport program, disposal program); the five transverse technical and scientific activities (safety, reversibility, cost, health and occupational safety, impact study). (J.S.)

  9. ''Cold crucible'' vitrification projects for low and high active waste

    International Nuclear Information System (INIS)

    Roux, P.; Jouan, A.

    1998-01-01

    In continuity of the CEA HLW vitrification process experienced for more than 20 years in industrial operations in Cogema reprocessing plants (Marcoule and La Hague), CEA has developed an advanced extended performance cold crucible glass melter to address a wider range of waste like LLW, ILW and in particular waste with very corrosive species or requiring glass with higher elaboration temperature. In the cold crucible melter the bath of molten glass is directly heated by induction while the walls are cooled in order to freeze a protective glass layer. This technology subsequently allows high glass throughput while keeping the flexibility, the maintainability and low secondary waste generation related to a small metallic melter. Its recent use in the glass industry and the thousands of hours of pilot tests performed on inactive surrogates have demonstrated the maturity of this technology and its flexibility of use for processing most of the waste generated at nuclear facilities. SGN has therefore proposed this technology in Italy and Korea and in USA in the frame of the Hanford Privatization phase 1 A feasibility study. Main features of this study but also tests results with Hanford surrogates and active samples are discussed. (author)

  10. Borosilicate glasses for the high activity waste vetrification

    International Nuclear Information System (INIS)

    Cantale, C.; Donato, A.; Guidi, G.

    1984-01-01

    Some results concerning the researches carried out on the high-level wastes vitrification at ENEA, Comb-Mepis-Rifiu laboratory are reported. A fission product solution referred to power plant nuclear fuel reprocessing has been selected and simulated with no radioactive chemicals. Some glass composition have been tested for the vitrification of this solution, the best of them being taken into consideration for real active tests at the hot bench scale plant ESTER in Ispra. The final glasses have been characterized from the chemical and physical point of view; moreover some microstructural investigations have been performed in order to identify few microsegregations and to test the degree of amorphousness of the products

  11. Lead corrosion evaluation in high activity nuclear waste container (Argentina)

    International Nuclear Information System (INIS)

    Guasp, R.; Lanzani, L.; Bruzzoni, P.; Cufre, W.; Semino, C.J.

    2000-01-01

    This report describes a study of high activity nuclear waste canister corrosion in a deep geological disposal. In this canister design, the vitrified nuclear waste stainless steel container is shielded by a 100 mm thick lead wall. For mechanical resistance, the canister will also have a thin carbon steel external liner. Experimental and mathematical modeling studies are aimed to asses the corrosion kinetics of the carbon steel liner in first instance and then, once this liner has been corroded away, the corrosion kinetics of the main lead barrier. Being that oxygen reduction is the main cathodic reaction that supports the anodic oxidation of iron, a model is described predicting the rate of oxygen consumption in a sealed deep nuclear waste disposal vault as a result of the canister corrosion. Oxidation processes other than container corrosion, and that can account also for oxygen depletion, are not taken into consideration. Corrosion experimental studies on lead and its alloys in groundwater are also reported. These experiments are aimed to improve the corrosion resistance of commercial lead in groundwater. (author)

  12. High active waste (HAW) data report January-June 1991

    International Nuclear Information System (INIS)

    Duijves, K.A.

    1992-04-01

    Data are presented from the High Active Waste (HAW) experiment, a large-scale, in situ test being performed underground at the Asse salt mine in Remlingen, FRG. These data include selected field information, the test configuration, instrumentation activities and comprehensive results from a large number of gauges. The results are measured data obtained from gap meters, thermocouples, linear displacement transducers, extensometers, inclinometers and pressure gauges. Data certification practices have been described together with the quality assurance of the data reduction and of the data base management system. The experiment began on November 8, 1988 and will continue for five years. Data in this report cover the period from January 1st, 1991 to June 30, 1991. (author). 4 refs.; 43 figs.; 8 tabs

  13. Technical baseline description of high-level waste and low-activity waste feed mobilization and delivery

    International Nuclear Information System (INIS)

    Papp, I.G.

    1997-01-01

    This document is a compilation of information related to the high-level waste (HLW) and low-activity waste (LAW) feed staging, mobilization, and transfer/delivery issues. Information relevant to current Tank Waste Remediation System (TWRS) inventories and activities designed to feed the Phase I Privatization effort at the Hanford Site is included. Discussions on the higher level Phase II activities are offered for a perspective on the interfaces

  14. Development of Simulants to Support Mixing Tests for High Level Waste and Low Activity Waste

    International Nuclear Information System (INIS)

    EIBLING, RUSSELLE.

    2004-01-01

    The objectives of this study were to develop two different types of simulants to support vendor agitator design studies and mixing studies. The initial simulant development task was to develop rheologically-bounding physical simulants and the final portion was to develop a nominal chemical simulant which is designed to match, as closely as possible, the actual sludge from a tank. The physical simulants to be developed included a lower and upper rheologically bounded: pretreated low activity waste (LAW) physical simulant; LAW melter feed physical simulant; pretreated high level waste (HLW) physical simulant; HLW melter feed physical simulant. The nominal chemical simulant, hereafter referred to as the HLW Precipitated Hydroxide simulant, is designed to represent the chemical/physical composition of the actual washed and leached sludge sample. The objective was to produce a simulant which matches not only the chemical composition but also the physical properties of the actual waste sample. The HLW Precipitated Hydroxide simulant could then be used for mixing tests to validate mixing, homogeneity and representative sampling and transferring issues. The HLW Precipitated Hydroxide simulant may also be used for integrated nonradioactive testing of the WTP prior to radioactive operation

  15. Alternatives evaluation of high activity radioactive wastes disposal

    International Nuclear Information System (INIS)

    Ciallella, N.R.; Petraitis, E.J.

    1989-01-01

    Different alternatives considered in the world to be used as barriers to isolate the high level radioactive from the environment wastes produced during the electric energy generation of nuclear origin are presented. Engineering and geologic barriers, are analyzed, considering nuclear fuel cycles with or without plutonium recycling; to that purpose the consideration of elements such as durability and resistance of the various engineering, availability of the fabrication processes, associated radiological impact, geological media apt to be used as geological barrier. Finally, the scopes of the Feasibility Study and Engineering draft are presented for the construction of a repository for high-level radioactive wastes, for the Argentine Nuclear Program needs, which contemplates the construction of six nuclear power plants with a potential installed towards the year 2000 GW( e ), with natural and/or lowly enriched uranium power plants and recycling of plutonium generated in the cycle. (Author) [es

  16. The safety of high activity long life nuclear waste

    International Nuclear Information System (INIS)

    Devillers, Ch.

    1998-01-01

    The article concerns the deep geological storage for managing high activity long life nuclear waste. He puts forward a context giving a structure to the discussions of those involved concerning an assessment of the safety of a deep geological deposit project. Three main aspects are put forward. The risks for future generations and the time scales to be considered: briefly, the deposit needs to satisfy two functions for protecting man and the environment, namely firstly isolating high activity radionuclides from the biosphere during the time required for their radioactive decay (about ten thousands years), and secondly delay and dilute long life radionuclides without any a priori time limit so as to reduce their effects in the biosphere to extremely low levels. The risks are linked to possible failures of the containment barriers whose causes need to be analysed and be provided against by suitable provisions concerning their design. The definition of these design provisions requires an in depth examination of uncertain elements. The main causes of uncertainty are listed according to the scale of time in question, that is O-10,000 years, 10,000-100,000 years and beyond 100,000 years, stressing the importance of selecting a stable geological site and more generally a solid concept that is not very sensitive in uncertainties. Beyond 100,000 years the extent of uncertainties no longer makes it possible to make realistic predictions. It is thus necessary to consider the alternative scenarios concerning geological and climatic changes and the corresponding increasing risks of radionuclides. The risks in question may be relativized by realizing that on this time scale, the residual activities of soluble and insoluble alpha and beta emitters are comparable to those of a storage centre located on the surface at the end of the monitoring period. Finally, the article considers the approach put forward concerning the safety of a deep geological storage advocated by the French

  17. Demonstration of sulfur solubility determinations in high waste loading, low-activity waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-04-25

    A method recommended by Pacific Northwest National Laboratory (PNNL) for sulfate solubility determinations in simulated low-activity waste glasses was demonstrated using three compositions from a recent Hanford high waste loading glass study. Sodium and sulfate concentrations in the glasses increased after each re-melting step. Visual observations of the glasses during the re-melting process reflected the changes in composition. The measured compositions showed that the glasses met the targeted values. The amount of SO3 retained in the glasses after washing was relatively high, ranging from 1.6 to 2.6 weight percent (wt %). Measured SnO2 concentrations were notably low in all of the study glasses. The composition of the wash solutions should be measured in future work to determine whether SnO2 is present with the excess sulfate washed from the glass. Increases in batch size and the amount of sodium sulfate added did not have a measureable impact on the amount of sulfate retained in the glass, although this was tested for only a single glass composition. A batch size of 250 g and a sodium sulfate addition targeting 7 wt %, as recommended by PNNL, will be used in future experiments.

  18. Melter feed viscosity during conversion to glass: Comparison between low-activity waste and high-level waste feeds

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Tongan [Pacific Northwest National Laboratory, Richland Washington; Chun, Jaehun [Pacific Northwest National Laboratory, Richland Washington; Dixon, Derek R. [Pacific Northwest National Laboratory, Richland Washington; Kim, Dongsang [Pacific Northwest National Laboratory, Richland Washington; Crum, Jarrod V. [Pacific Northwest National Laboratory, Richland Washington; Bonham, Charles C. [Pacific Northwest National Laboratory, Richland Washington; VanderVeer, Bradley J. [Pacific Northwest National Laboratory, Richland Washington; Rodriguez, Carmen P. [Pacific Northwest National Laboratory, Richland Washington; Weese, Brigitte L. [Pacific Northwest National Laboratory, Richland Washington; Schweiger, Michael J. [Pacific Northwest National Laboratory, Richland Washington; Kruger, Albert A. [U.S. Department of Energy, Office of River Protection, Richland Washington; Hrma, Pavel [Pacific Northwest National Laboratory, Richland Washington

    2017-12-07

    During nuclear waste vitrification, a melter feed (generally a slurry-like mixture of a nuclear waste and various glass forming and modifying additives) is charged into the melter where undissolved refractory constituents are suspended together with evolved gas bubbles from complex reactions. Knowledge of flow properties of various reacting melter feeds is necessary to understand their unique feed-to-glass conversion processes occurring within a floating layer of melter feed called a cold cap. The viscosity of two low-activity waste (LAW) melter feeds were studied during heating and correlated with volume fractions of undissolved solid phase and gas phase. In contrast to the high-level waste (HLW) melter feed, the effects of undissolved solid and gas phases play comparable roles and are required to represent the viscosity of LAW melter feeds. This study can help bring physical insights to feed viscosity of reacting melter feeds with different compositions and foaming behavior in nuclear waste vitrification.

  19. Long-term α-hazard of high activity waste from nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Girardi, F.; Bertozzi, G.

    1974-01-01

    The concentration and decay of α-emitters in high activity waste arising from spent nuclear fuel reprocessing was calculated under specified reference conditions. An attempt to evaluate the long-term hazard of such waste is being made by applying the ''barrier'' approach derived from reactor safety studies. Four barriers were identified, which could be evaluated in a probabilistic way by taking into account the great uncertainties present in each of them. The barriers are: 1) quality of the segregation afforded by deep geological formations, 2) stability of conditioned waste (chemical and physical), 3) retention by immediate surrounding, 4) distribution pattern of actinides in the environment. The analysis of a fictional accident showed that the uncertainties connected with the evaluation of the barriers' value are rather large. Additional studies particularly on the stability of conditioned waste and ecological properties of the environment towards actinides, would considerably improve our knowledge of the value of the barrier system. Chemical separation of actinides from high activity waste would be an additional option of undoubted value for the disposal of high activity waste. Its value for the overall safety of the entire waste inventory depends on many factors which need better evaluation, such as safety of the disposal of the separated actinides and the amount and quality of the additional waste generated by the separation process. An analysis of various levels of possible separations suggests that a reasonable target might be: Pu, Am and Cm, decontamination factor 10 3 ; Np, coextraction with U and Pu with a 90% yield

  20. The denitration of simulated fast reactor highly active liquor waste

    International Nuclear Information System (INIS)

    Saum, C.J.; Ford, L.H.; Platts, N.

    1981-11-01

    Vitrification of the highly active arisings from PFR fuel reprocessing is proposed as the optimum long-term solution to the disposal problem. During vitrification ruthenium volatilises as the tetroxide. Evidence is presented which indicates that a substantial reduction in volatility can be effected by denitration of the liquid feed by treatment with formic acid. The kinetics and stoichiometry of the reactions involved in denitration are examined and empirical rate equations developed. The predictions of the empirical rate equation have been confirmed using a one-tenth scale continuous denitrator, thus giving confidence for the design of full-scale units. (author)

  1. Radioprotection and physical surveillance during activities of liquid wastes of high and low activity in italian ITREC plant

    International Nuclear Information System (INIS)

    Petagna, Edoardo; Tortorelli, Pietro

    1997-03-01

    Many studies were made in ITREC Plant, located in ENEA - Trisaia Research Center, in the field of the nuclear fuel reprocessing, in the past years. During these activities liquid wastes of high and low activity were yielded and stored in the special area of tanks named Waste-1. In order to condition the low activity liquid wastes, essentially fission products, beta and gamma emitters, was built the SIRTE Plant (Integrate System for the Raise and Effluents Treatment) based on cementation process. In the present work, the radiological monitoring performed within the plant during the first campaign of cementation, is showed

  2. Performance of cement solidification with barium for high activity liquid waste including sulphate

    International Nuclear Information System (INIS)

    Waki, Toshikazu; Yamada, Motoyuki; Horikawa, Yoshihiko; Kaneko, Masaaki; Saso, Michitaka; Haruguchi, Yoshiko; Yamashita, Yu; Sakai, Hitoshi

    2009-01-01

    The target liquid waste to be solidified is generated from PWR primary loop spent resin treatment with sulphate acid, so, its main constituent is sodium sulphate and the activity of this liquid is relatively high. Waste form of this liquid waste is considered to be a candidate for the subsurface disposal. The disposed waste including sulphate is anticipated to rise a concentration of sulphate ion in the ground water around the disposal facility and it may cause degradation of materials such as cement and bentonite layer and comprise the disposal facility. There could be two approaches to avoid this problem, the strong design of the disposal facility and the minimization of sulphaste ion migration from the solidified waste. In this study, the latter approach was examined. In order to keep the low concentration of sulphate ion in the ground water, it is effective to make barium sulphate by adding barium compound into the liquid waste in solidification. However, adding equivalent amount of barium compound with sulphate ion causes difficulty of mixing, because production of barium sulphate causes high viscosity. In this study, mixing condition after and before adding cement into the liquid waste was estimated. The mixing condition was set with consideration to keep anion concentration low in the ground water and of mixing easily enough in practical operation. Long term leaching behavior of the simulated solidified waste was also analyzed by PHREEQC. And the concentration of the constitution affected to the disposal facility was estimated be low enough in the ground water. (author)

  3. Will vitrified high active reactor waste be deposited on the seabed

    International Nuclear Information System (INIS)

    Koksvik, R.M.

    1978-01-01

    The NRPB in the UK has carried out a preliminary study on the consequences of the disposal of high active waste on the ocean bed at great depths. This is reported in 'Assessment of the radiological protection aspects of disposal of high level waste on the ocean floor', by P.D. Grimswood and G.A.M. Webb (NRPB-R 48). This report is here briefly summarised and commented on. It deals with the radiological consequences to man of the erosion and corrosion of glassified (vitrified) radioactive waste, in the deep (>4000m) ocean, radionuclide migration and path ways to man. The conclusion is that considerations of radiological protection do not preclude the disposal of specially treated radioactive waste on the ocean bed. (JIW)

  4. Californium-252 neutron activation analysis of high-level processed nuclear tank waste

    International Nuclear Information System (INIS)

    Troyer, G.L.; Purcell, M.A.

    2000-01-01

    The basis for production assessment of the vitrification of Hanford nuclear fuel reprocessing wastes will be high-precision measurements of the elemental sodium content. However, the chemical analysis of both radioactive and nonradioactive components in nuclear waste can be challenged by high radiation dose rates. The dose rates compromise many analytical techniques as well as pose personnel dosimetry risks. In many cases, reduction of dose rates through dilution compromises the precision and sensitivity for certain key components. The use of neutron activation analysis (NAA) provides a method of analysis that avoids the need for dilutions or extensive sample preparation. These waste materials also contain trace quantities of fissionable isotopes, which, through neutron activation, can be estimated by delayed neutron counting of fissioned fragments

  5. Experimental Design for Hanford Low-Activity Waste Glasses with High Waste Loading

    Energy Technology Data Exchange (ETDEWEB)

    Piepel, Gregory F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cooley, Scott K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vienna, John D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-07-24

    This report discusses the development of an experimental design for the initial phase of the Hanford low-activity waste (LAW) enhanced glass study. This report is based on a manuscript written for an applied statistics journal. Appendices A, B, and E include additional information relevant to the LAW enhanced glass experimental design that is not included in the journal manuscript. The glass composition experimental region is defined by single-component constraints (SCCs), linear multiple-component constraints (MCCs), and a nonlinear MCC involving 15 LAW glass components. Traditional methods and software for designing constrained mixture experiments with SCCs and linear MCCs are not directly applicable because of the nonlinear MCC. A modification of existing methodology to account for the nonlinear MCC was developed and is described in this report. One of the glass components, SO3, has a solubility limit in glass that depends on the composition of the balance of the glass. A goal was to design the experiment so that SO3 would not exceed its predicted solubility limit for any of the experimental glasses. The SO3 solubility limit had previously been modeled by a partial quadratic mixture model expressed in the relative proportions of the 14 other components. The partial quadratic mixture model was used to construct a nonlinear MCC in terms of all 15 components. In addition, there were SCCs and linear MCCs. This report describes how a layered design was generated to (i) account for the SCCs, linear MCCs, and nonlinear MCC and (ii) meet the goals of the study. A layered design consists of points on an outer layer, and inner layer, and a center point. There were 18 outer-layer glasses chosen using optimal experimental design software to augment 147 existing glass compositions that were within the LAW glass composition experimental region. Then 13 inner-layer glasses were chosen with the software to augment the existing and outer

  6. Experimental Design for Hanford Low-Activity Waste Glasses with High Waste Loading

    International Nuclear Information System (INIS)

    Piepel, Gregory F.; Cooley, Scott K.; Vienna, John D.; Crum, Jarrod V.

    2015-01-01

    This report discusses the development of an experimental design for the initial phase of the Hanford low-activity waste (LAW) enhanced glass study. This report is based on a manuscript written for an applied statistics journal. Appendices A, B, and E include additional information relevant to the LAW enhanced glass experimental design that is not included in the journal manuscript. The glass composition experimental region is defined by single-component constraints (SCCs), linear multiple-component constraints (MCCs), and a nonlinear MCC involving 15 LAW glass components. Traditional methods and software for designing constrained mixture experiments with SCCs and linear MCCs are not directly applicable because of the nonlinear MCC. A modification of existing methodology to account for the nonlinear MCC was developed and is described in this report. One of the glass components, SO 3 , has a solubility limit in glass that depends on the composition of the balance of the glass. A goal was to design the experiment so that SO 3 would not exceed its predicted solubility limit for any of the experimental glasses. The SO 3 solubility limit had previously been modeled by a partial quadratic mixture model expressed in the relative proportions of the 14 other components. The partial quadratic mixture model was used to construct a nonlinear MCC in terms of all 15 components. In addition, there were SCCs and linear MCCs. This report describes how a layered design was generated to (i) account for the SCCs, linear MCCs, and nonlinear MCC and (ii) meet the goals of the study. A layered design consists of points on an outer layer, and inner layer, and a center point. There were 18 outer-layer glasses chosen using optimal experimental design software to augment 147 existing glass compositions that were within the LAW glass composition experimental region. Then 13 inner-layer glasses were chosen with the software to augment the existing and outer-layer glasses. The experimental

  7. Development plan. High activity-long living wastes project. Abstract; Plan de developpement. Projet HAVL. Resume

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    This brochure presents the actions that the ANDRA (the French national agency of radioactive wastes) has to implement in the framework of the project of high activity-long living (HALL) radioactive wastes (HAVL project) conformably to the requirements of the program defined in the law from June 28, 2006 (law no 2006-739). This law precises the three, complementary, research paths to explore for the management of this type of wastes: separation and transmutation of long-living radioactive elements, reversible disposal in deep geologic underground, and long duration storage. The ANDRA's action concerns the geologic disposal aspect. The following points are presented: the HALL wastes and their containers, the reversible disposal procedure, the HAVL project: financing of researches, storage concepts, development plan of the project (dynamics, information and dialogue approach, input data, main steps, schedule); the nine programs of the HAVL project (laboratory experiments and demonstration tests, surface survey, scientific program, simulation program, surface engineering studies and technological tests, information and communication program, program of environment and facilities surface observation and monitoring, waste packages management, monitoring and transport program, disposal program); the five transverse technical and scientific activities (safety, reversibility, cost, health and occupational safety, impact study). (J.S.)

  8. The opalinus clay project - disposal of medium and highly-active nuclear wastes

    International Nuclear Information System (INIS)

    Mueller, U.

    2003-01-01

    This article describes the project to demonstrate the feasibility of disposing of long-living medium-active and highly-radioactive nuclear wastes in sedimentary rock in Switzerland. The disposal tasks to be carried out are reviewed and the solutions proposed are described, including short-term handling, intermediate storage and final disposal of low, medium and highly-active wastes. The present state of affairs is described and, in particular, the feasibility of implementing a final storage facility in the opalinus clay beds to be found in northern Switzerland. The project for such a facility in the wine-growing area of the canton of Zurich is described in detail, including the storage concept, the technology to be used and operational aspects as well as questions of safety

  9. Preparation of Highly Porous Binderless Active Carbon Monoliths from Waste Aspen Sawdust

    Directory of Open Access Journals (Sweden)

    Dawei Li

    2014-01-01

    Full Text Available Waste aspen sawdust was used as a precursor to prepare binderless active carbon monoliths (ACMs with high porosities. The ACMs were prepared by activation with H3PO4 at different activation temperatures (500 to 700 °C and retention times (1 to 3 h. Their morphologies, yields, textural properties, and microcrystalline structures were investigated using scanning electron microscopy (SEM, an analytical balance, N2 adsorption/desorption techniques, and X-ray diffraction (XRD. The results indicated that waste aspen sawdust could be successfully converted into highly porous binderless ACMs. The apparent specific surface area (SSA and yield of ACMs were in the range of 688 to 951 m2/g and 26.6 to 36.2%, respectively. Highly microporous ACMs with a micropore percentage of 91.1%, apparent specific surface area of 951 m2/g, pore volume of 0.481 mL/g, and bulk density of 0.56 g/mL could be produced by activation at 700 °C for 1 h. Increasing the activation temperature or retention time increased the specific surface area, pore volume, and turbostratic degree, but decreased the yield.

  10. Final disposal of spent fuels and high activity waste: status and trends in the world

    International Nuclear Information System (INIS)

    Herscovich de Pahissa, Marta

    2007-01-01

    Geological disposal of spent nuclear fuel and high level waste from reprocessing, properly conditioned, is described. This issue is a major challenge related to radioactive waste management. Several options are analyzed, such as application of separation and transmutation to high level waste before final disposal; need of multinational repositories; a phased approach to deep geological disposal and long term surface storage. Bearing in mind this information, a future article will report the state of art in the world. (author) [es

  11. High capacity cask (TN28V) and International Transport System for the return shipment of vitrified high activity wastes

    International Nuclear Information System (INIS)

    Sert, G.; Savornin, B.; Rouquette, Y.

    1989-01-01

    The reprocessing of spent fuel generates different kinds of wastes. Among them fission products and non fissile actinides represent 98% of the radioactivity; these wastes are separated, concentrated, mixed with molten glass and poured into stainless steel containers. For political reasons, it is necessary to return these vitrified high activity wastes to the foreign countries which have decided to have their spent fuel reprocessed in France. So the transport of vitrified waste is vital for both the reprocessor and the utilities that have trusted the reprocessor and this operation has to be securely performed to give satisfaction to all concerned particles. For that reason Cogema will control the whole transport activity from La Hague plants to the receiving facilities of the customers. Therefore cogema will be responsible of the transport whatever the cask type (transport or storage) and will subcontract the transport operation to experienced companies such as Transnucleaire, PNTL or NTL, who will act on behalf of Cogema. Cogema will be the owner of the transport casks while the storage casks will normally be owned by the customers. Both cask types will of course have to comply with the requirements of La Hague, as published by Cogema

  12. UP2 400 High Activity Oxide Legacy Waste Retrieval Project Scope and Progress-13048

    Energy Technology Data Exchange (ETDEWEB)

    Chabeuf, Jean-Michel; Varet, Thierry [AREVA Site Value Development Business Unit, La Hague Site (France)

    2013-07-01

    The High Activity Oxide facility (HAO) reprocessed sheared and dissolved 4500 metric tons of light water reactor fuel the fuel of the emerging light water reactor spent fuel between 1976 and 1998. Over the period, approximately 2200 tons of process waste, composed primarily of sheared hulls, was produced and stored in a vast silo in the first place, and in canisters stored in pools in subsequent years. Upon shutdown of the facility, AREVA D and D Division in La Hague launched a thorough investigation and characterization of the silos and pools content, which then served as input data for the definition of a legacy waste retrieval and reconditioning program. Basic design was conducted between 2005 and 2007, and was followed by an optimization phase which lead to the definition of a final scenario and budget, 12% under the initial estimates. The scenario planned for the construction of a retrieval and reconditioning cell to be built on top of the storage silo. The retrieved waste would then be rinsed and sorted, so that hulls could subsequently be sent to La Hague high activity compacting facility, while resins and sludge would be cemented within the retrieval cell. Detailed design was conducted successfully from 2008 until 2011, while a thorough research and development program was conducted in order to qualify each stage of the retrieval and reconditioning process, and assist in the elaboration of the final waste package specification. This R and D program was defined and conducted as a response and mitigation of the major project risks identified during the basic design process. Procurement and site preparatory works were then launched in 2011. By the end of 2012, R and D is nearly completed, the retrieval and reconditioning process have been secured, the final waste package specification is being completed, the first equipment for the retrieval cell is being delivered on site, while preparation works are allowing to free up space above and around the silo, to

  13. Demonstration of the TRUEX process for the treatment of actual high activity tank waste at the INEEL using centrifugal contactors

    International Nuclear Information System (INIS)

    Law, J.D.; Brewer, K.N.; Todd, T.A.; Olson, L.G.

    1997-01-01

    The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Engineering and Environmental Laboratory (INEEL), formerly reprocessed spent nuclear fuel to recover fissionable uranium. The radioactive raffinates from the solvent extraction uranium recovery processes were converted to granular solids (calcine) in a high temperature fluidized bed. A secondary liquid waste stream was generated during the course of reprocessing, primarily from equipment decontamination between campaigns and solvent wash activities. This acidic tank waste cannot be directly calcined due to the high sodium content and has historically been blended with reprocessing raffinates or non-radioactive aluminum nitrate prior to calcination. Fuel reprocessing activities are no longer being performed at the ICPP, thereby eliminating the option of waste blending to deplete the waste inventory. Currently, approximately 5.7 million liters of high-activity waste are temporarily stored at the ICPP in large underground stainless-steel tanks. The United States Environmental Protection Agency and the Idaho Department of Health and Welfare filed a Notice of Noncompliance in 1992 contending some of the underground waste storage tanks do not meet secondary containment. As part of a 1995 agreement between the State of Idaho, the Department of Energy, and the Department of Navy, the waste must be removed from the tanks by 2012. Treatment of the tank waste inventories by partitioning the radionuclides and immobilizing the resulting high-activity and low-activity waste streams is currently under evaluation. A recent peer review identified the most promising radionuclide separation technologies for evaluation. The Transuranic Extraction-(TRUEX) process was identified as a primary candidate for separation of the actinides from ICPP tank waste

  14. R and D Activities on high-level nuclear waste management

    International Nuclear Information System (INIS)

    Watanabe, Shosuke

    1985-01-01

    High-level liquid waste (HLLW) at Tokai Reprocessing Plant has been generated from reprocessing of spent fuels from the light water reactors, and successfully managed since 1977. At the time of 1984, about 154m 3 of HLLW from 170 tons of spent fuels were stored in three high-integrity stainless steel tanks (90m 3 for each) as a nitric acid aqueous solution. The HLLW arises mainly from the first cycle solvent extraction phase. Alkaline solution to scrub the extraction solvent is another source of HLLW. The Advisory Committee on Radioactive Waste Management reported the concept on disposal of high-level waste (HLW) in Japan in 1980 report, that the waste be solidified into borosilicate glass and then be disposed in deep geologic formation so as to minimize the influence of the waste on human environment, with the aid of multibarrier system which is the combination of natural barrier and engineered barrier

  15. Source segregation and food waste prevention activities in high-density households in a deprived urban area

    International Nuclear Information System (INIS)

    Rispo, A.; Williams, I.D.; Shaw, P.J.

    2015-01-01

    Highlights: • Study of waste management in economically and socially deprived high-density housing. • Food waste segregation, prevention and recycling activities investigated. • Study involved a waste audit and household survey of 1034 households. • Populations in such areas are “hard-to-reach”. • Exceptional efforts and additional resources are required to improve performance. - Abstract: A waste audit and a household questionnaire survey were conducted in high-density housing estates in one of the most economically and socially deprived areas of England (Haringey, London). Such areas are under-represented in published research. The study examined source segregation, potential participation in a food waste segregation scheme, and food waste prevention activities in five estates (1034 households). The results showed that: contamination of recyclables containers was low; ca. 28% of the mixed residual waste’s weight was recyclable; food waste comprised a small proportion of the waste from these residents, probably because of their relatively disadvantaged economic circumstances; and the recycling profile reflected an intermittent pattern of behaviour. Although the majority of respondents reported that they would participate in a food waste separation scheme, the response rate was low and many responses of “don’t know” were recorded. Municipalities committed to foster improved diversion from landfill need to recognise that there is no “quick and easy fix”, regardless of local or national aspirations. Lasting and sustained behaviour change requires time and the quality of service provision and associated infrastructure play a fundamental role in facilitating residents to participate effectively in waste management activities that maximise capture of source-segregated materials. Populations in deprived areas that reside in high-rise, high-density dwellings are “hard-to-reach” in terms of participation in recycling schemes and exceptional

  16. Source segregation and food waste prevention activities in high-density households in a deprived urban area

    Energy Technology Data Exchange (ETDEWEB)

    Rispo, A.; Williams, I.D., E-mail: idw@soton.ac.uk; Shaw, P.J.

    2015-10-15

    Highlights: • Study of waste management in economically and socially deprived high-density housing. • Food waste segregation, prevention and recycling activities investigated. • Study involved a waste audit and household survey of 1034 households. • Populations in such areas are “hard-to-reach”. • Exceptional efforts and additional resources are required to improve performance. - Abstract: A waste audit and a household questionnaire survey were conducted in high-density housing estates in one of the most economically and socially deprived areas of England (Haringey, London). Such areas are under-represented in published research. The study examined source segregation, potential participation in a food waste segregation scheme, and food waste prevention activities in five estates (1034 households). The results showed that: contamination of recyclables containers was low; ca. 28% of the mixed residual waste’s weight was recyclable; food waste comprised a small proportion of the waste from these residents, probably because of their relatively disadvantaged economic circumstances; and the recycling profile reflected an intermittent pattern of behaviour. Although the majority of respondents reported that they would participate in a food waste separation scheme, the response rate was low and many responses of “don’t know” were recorded. Municipalities committed to foster improved diversion from landfill need to recognise that there is no “quick and easy fix”, regardless of local or national aspirations. Lasting and sustained behaviour change requires time and the quality of service provision and associated infrastructure play a fundamental role in facilitating residents to participate effectively in waste management activities that maximise capture of source-segregated materials. Populations in deprived areas that reside in high-rise, high-density dwellings are “hard-to-reach” in terms of participation in recycling schemes and exceptional

  17. High performance supercapacitor from activated carbon derived from waste orange skin

    Science.gov (United States)

    Ahmed, Sultan; Hussain, S.; Ahmed, Ahsan; Rafat, M.

    2018-05-01

    Activated carbon due to its inherent properties such as large surface area and low cost is most frequently used electrode material for supercapacitor. Activated carbon has been previously derived from various biomass such as coconut shell, coffee bean etc. Herein, we report the synthesis of activated carbon from waste orange skin. The material was synthesized employing chemical activation method and the success of synthesis was confirmed by its physical and electrochemical properties. The physical properties of the as-prepared sample were studied using the techniques of XRD, SEM, Raman spectroscopy and N2 adsorption/desorption analysis while its electrochemical properties were studied in two-electrode assembly using liquid electrolyte (consisting of 1 M solution of LiTFSI dispersed in ionic liquid EMITFSI) and employing the techniques of cyclic voltammetry, electrochemical impedance spectroscopy and galvanostatic charge- discharge. The synthesized sample of activated carbon exhibits high specific capacitance of 115 F g-1 at 10 mV s-1. Also, the activated carbon electrode shows the retention of ˜75% in initial capacitance value for more than 2000 initial cycles, indicating the as-prepared activated carbon can be profitably used as electrode material for energy storage devices.

  18. Testing and evaluation of the properties of various potential materials for immobilizing high activity waste

    International Nuclear Information System (INIS)

    Malow, G.; Beran, V.; Lutze, W.

    1980-01-01

    Second joint annual report of the work performed on the testing and evaluation of materials for immobilizing high activity waste under Community contracts. At Marcoule, active block samples containing HAW from the Marcoule reprocessing plant were cast to the specification of five of the six original reference samples and leach tested at ambient temperature. Phosphate glass -bead samples produced by the Gelsenberg/DWK PAMELA process- were included in the test programme at HMI-Berlin and UKAEA Harwell. Leaching tests of inactive samples with leachants of various pH-values, with ionized water and with natural water compositions representative of certain repository conditions (salt, clay and granite) were added to the Harwell programme. The studies of radiation and thermal effects and the investigation of devitrification phenomena, started in 1977, continued, as samples reached annealing times of 2400 h and doses 4 x 10 17 dpg. The results achieved have so far confirmed most of the favourable preliminary assessments of glass based solidification products. At this stage the programme aims primarily at the understanding of physical and chemical phenomena rather that at verification under realistic waste storage and disposal conditions

  19. High Level Radioactive Waste Management

    International Nuclear Information System (INIS)

    1991-01-01

    The proceedings of the second annual international conference on High Level Radioactive Waste Management, held on April 28--May 3, 1991, Las Vegas, Nevada, provides information on the current technical issue related to international high level radioactive waste management activities and how they relate to society as a whole. Besides discussing such technical topics as the best form of the waste, the integrity of storage containers, design and construction of a repository, the broader social aspects of these issues are explored in papers on such subjects as conformance to regulations, transportation safety, and public education. By providing this wider perspective of high level radioactive waste management, it becomes apparent that the various disciplines involved in this field are interrelated and that they should work to integrate their waste management activities. Individual records are processed separately for the data bases

  20. Activation/waste management

    International Nuclear Information System (INIS)

    Maninger, C.

    1984-10-01

    The selection of materials and the design of the blankets for fusion reactors have significant effects upon the radioactivity generated by neutron activation in the materials. This section considers some aspects of materials selection with respect to waste management. The activation of the materials is key to remote handling requirements for waste, to processing and disposal methods for waste, and to accident severity in waste management operations. In order to realize the desirable evnironmental potentials of fusion power systems, there are at least three major goals for waste management. These are: (a) near-surface burial; (b) disposal on-site of the fusion reactor; (c) acceptable radiation doses at least cost during and after waste management operations

  1. Loading and transport of high-active waste (HAW) with the TN85 flask in 2008

    International Nuclear Information System (INIS)

    Rys, Michael; Horn, Thomas; Graf, Wilhelm; Bonface, Jean-Michael

    2009-01-01

    As a part of the operation of nuclear power plants, it is essential to safely manage the radioactive waste. With new developments in science and technology, it is a dynamic process to adapt procedures, equipment and flasks to be used in the future. This is a task for specialists - a task for GNS Gesellschaft fuer Nuklear-Service mbH and for TN International. Until 1994 reprocessing of spent fuel from German nuclear power plants was mandatory for the Utilities (EVU) in Germany. Basis for the reprocessing was the German Atomic Act. The German Utilities concluded contracts on reprocessing with Compagnie Generale des Matieres Nucleaires (COGEMA, now AREVA NC) in France and British Nuclear Fuels plc (BNFL, now INS) in England. The total amount to be reprocessed comes to 5309 t HM contracted to AREVA NC and 768 t HM contracted to INS. The waste generated from reprocessing - or an equivalent amount of radioactive material - has to be returned to the country of origin. In 1979 already an exchange of notes took place between the German and the French government with the obligation of both sides to enable and support the return of reprocessing residues or equivalents. The return of high-active waste (HAW) from France has started in 1996 with the first attribution of 28 glass canisters (one flask) to German Utilities by AREVA NC. Until 2007, 75 flasks loaded with vitrified residue (VR) canisters have been transported to Gorleben. For these transports CASTOR registered HAW 20/28 CG flasks have been used. This presentation will give some background information about the last HAW transport in 2008 with the new flask generation of the type TN85. It will also describe the assembly of the new flask, the preparation of the flask for the loading campaign as well as the loading procedure. (orig.)

  2. Loading and transport of high-active waste (HAW) with the TN85 flask in 2008

    Energy Technology Data Exchange (ETDEWEB)

    Rys, Michael; Horn, Thomas; Graf, Wilhelm [GNS Gesellschaft fuer Nuklear-Service mbH (Germany); Bonface, Jean-Michael [TN International, Montigny-le-Bretonneux (France)

    2009-07-01

    As a part of the operation of nuclear power plants, it is essential to safely manage the radioactive waste. With new developments in science and technology, it is a dynamic process to adapt procedures, equipment and flasks to be used in the future. This is a task for specialists - a task for GNS Gesellschaft fuer Nuklear-Service mbH and for TN International. Until 1994 reprocessing of spent fuel from German nuclear power plants was mandatory for the Utilities (EVU) in Germany. Basis for the reprocessing was the German Atomic Act. The German Utilities concluded contracts on reprocessing with Compagnie Generale des Matieres Nucleaires (COGEMA, now AREVA NC) in France and British Nuclear Fuels plc (BNFL, now INS) in England. The total amount to be reprocessed comes to 5309 t HM contracted to AREVA NC and 768 t HM contracted to INS. The waste generated from reprocessing - or an equivalent amount of radioactive material - has to be returned to the country of origin. In 1979 already an exchange of notes took place between the German and the French government with the obligation of both sides to enable and support the return of reprocessing residues or equivalents. The return of high-active waste (HAW) from France has started in 1996 with the first attribution of 28 glass canisters (one flask) to German Utilities by AREVA NC. Until 2007, 75 flasks loaded with vitrified residue (VR) canisters have been transported to Gorleben. For these transports CASTOR {sup registered} HAW 20/28 CG flasks have been used. This presentation will give some background information about the last HAW transport in 2008 with the new flask generation of the type TN85. It will also describe the assembly of the new flask, the preparation of the flask for the loading campaign as well as the loading procedure. (orig.)

  3. Deep disposal of high activity radioactive wastes: the study of engineered and geological barriers behaviour

    International Nuclear Information System (INIS)

    Yu Jun; Cui; Delage, P.; Laure, E. de; Behrouz, Gatmiri; Sulem, J.; Anh Minh, Tang

    2008-09-01

    One option for the isolation of high activity and long lived radioactive wastes is the disposal of the vitrified waste containers in galleries dug inside impermeable rocks of the deep underground (granite, argillite, salt). The multi-barrier isolation concept is based on the use of successive barriers to avoid the migration of radionuclides towards the biosphere (container envelope, engineered barrier made of compacted swelling clay, and host rock). In parallel to the works carried out in underground laboratories, experiments and simulation works are performed in order to understand the behaviour of storage facilities and barriers under the effects of constraints, water fluxes and temperature changes. In this context, the UR Navier geotechnical team (CERMES), a joint research unit of Ecole des Ponts ParisTech and LCPC, has been working for more than 15 years on this topic for various contractors. These works are based on original experimental devices allowing to identify the thermo-hydro-mechanical phenomena and thereafter to model them. This dossier presents a summary of these works. (J.S.)

  4. Researches on the management of high activity and long-lived radioactive wastes. Axis 1 - separation-transmutation

    International Nuclear Information System (INIS)

    2005-11-01

    This document gathers the transparencies of seven presentations given at a technical workshop of the French nuclear energy society (SFEN) about the researches on separation-transmutation of high activity and long-lived radioactive wastes. The presentations deal with: inventory and radiotoxicity of the rad-wastes in concern; industrial experience; experience on chemical separation: molecules and processes; reactors physics and transmutation - reactors for transmutation; fuels and targets; scenarios that include transmutation; environmental impacts of these different scenarios. (J.S.)

  5. Enhanced high-solids anaerobic digestion of waste activated sludge by the addition of scrap iron.

    Science.gov (United States)

    Zhang, Yaobin; Feng, Yinghong; Yu, Qilin; Xu, Zibin; Quan, Xie

    2014-05-01

    Anaerobic digestion of waste activated sludge usually requires pretreatment procedure to improve the bioavailability of sludge, which involves considerable energy and high expenditures. This study proposes a cost-effective method for enhanced anaerobic digestion of sludge without a pretreatment by directly adding iron into the digester. The results showed that addition of Fe(0) powder could enhance 14.46% methane yield, and Fe scrap (clean scrap) could further enhance methane yield (improving rate 21.28%) because the scrap has better mass transfer efficiency with sludge and liquid than Fe(0) powder. The scrap of Fe with rust (rusty scrap) could induce microbial Fe(III) reduction, which resulted in achieving the highest methane yield (improving rate 29.51%), and the reduction rate of volatile suspended solids (VSS) was also highest (48.27%) among Fe powder, clean scrap and rusty scrap. PCR-DGGE proved that the addition of rusty scrap could enhance diversity of acetobacteria and enrich iron-reducing bacteria to enhance degradation of complex substrates. Copyright © 2014 Elsevier Ltd. All rights reserved.

  6. Demonstration of innovative partitioning processes for minor actinide recycling from high active waste solutions

    International Nuclear Information System (INIS)

    Modolo, G.; Wilden, A.; Geist, A.; Malmbeck, R.; Taylor, R.

    2014-01-01

    The recycling of the minor actinides (MA) using the Partitioning and Transmutation strategy (P and T) could contribute significantly to reducing the volume of high level waste in a geological repository and to decreasing the waste's longterm hazards originating from the long half-life of the actinides. Several extraction processes have been developed worldwide for the separation and recovery of MA from highly active raffinates (HAR, e.g. the PUREX raffinate). A multi-cycle separation strategy has been developed within the framework of European collaborative projects. The multi-cycle processes, on the one hand, make use of different extractants for every single process. Within the recent FP7 European research project ACSEPT (Actinide reCycling by SEParation and Transmutation), the development of new innovative separation processes with a reduced number of cycles was envisaged. In the so-called 'innovative SANEX' concept, the trivalent actinides and lanthanides are co-extracted from the PUREX raffinate by a DIAMEX like process (e.g. TODGA). Then, the loaded solvent is subjected to several stripping steps. The first one concerns selectively stripping the actinides(III) with selective water-soluble ligands (SO3-Ph-BTB), followed by the subsequent stripping of trivalent lanthanides. A more challenging route studied also within our laboratories is the direct actinide(III) separation from a PUREX-type raffinate using a mixture of CyMe 4 BTBP and TODGA as extractants, the so-called One cycle SANEX process. A new approach, which was also studied within the ACSEPT project, is the GANEX (Grouped ActiNide EXtraction) concept addressing the simultaneous partitioning of all transuranium (TRU) elements for their homogeneous recycling in advanced generation IV reactor systems. Bulk uranium is removed in the GANEX 1st cycle, e.g. using a monoamide extractant and the GANEX 2nd cycle then separates the TRU. A solvent composed of TODGA + DMDOHEMA in kerosene has been shown to

  7. Submission of the national commission of the public debate on the options concerning the long life high and medium activity radioactive wastes management

    International Nuclear Information System (INIS)

    2006-01-01

    This document deals with the presentation of a public debate on the radioactive wastes management and the opportunities of its organization. It presents successively the long life high and medium activity radioactive wastes, the today radioactive wastes management policy and some questions and topics which could be discussed during the debate. (A.L.B.)

  8. High-Temperature Corrosion Study for the RPP Low Activity Waste Melter

    International Nuclear Information System (INIS)

    Marshall, K.M.

    2003-01-01

    The River Protection Program (RPP) low activity waste (LAW) melter design incorporates a series of bubblers used to increase convection in the molten glass. Through runs of a pilot melter at Duratek, Inc. in Columbia, Maryland, the bubblers have been identified as the major component limiting LAW melter availability, requiring frequent replacement due to corrosive degradation, primarily at the melt line. Laboratory experiments were performed to evaluate the performance of several alloys and coatings in simulated RPP low activity waste melter vapor space and molten glass environments. The performance of the alloys and coatings was studied in order to advance our understanding of how these materials react at the melt/air interface inside the melter. The ultimate goal was to identify a material with superior performance compared to that of Inconel 693, and to deliver a bubbler sub-assembly made of that material to the RPP LAW melter pilot facility for further testing

  9. Leaching of vitrified high-level-active-waste in a near reality simulated repository system

    International Nuclear Information System (INIS)

    Froeschen, W.; Wolf, G.K.

    1987-01-01

    In the FRG it is planned to vitrify the high level waste from spent fuel reprocessing and to dispose of in a salt-mine. If water penetrates into the repository a highly corrosive brine (Q-brine) will be formed and radioactive material may be leached from the glasses and transported to human environment. The corrosion system of brine, corroded steel containers of the vitrified waste, and waste-glasses was investigated under near reality conditions. Experiments in hydrothermal environment were carried out including gamma radiation of the waste-glasses and ceramic In Can Lining between glasses and metallic containments. Screening experiments by application of external cobalt-gamma-radiation showed no principal changes in leaching behaviour of simulate glasses compared to leaching without radiation. Radiation effects result in pH changes mainly which are diminished by buffer capacity of Q-brine. Lining of steel containments with ceramic fleece does not reduce leaching but retards solution of Mo and Sr into brine. Decreasing of elements Sr, Cs and Mo in the near surface area of the glass and increasing of Zr and Ti has been found to be enhanced considerably in presence of canister corrosion products in Q-brine as well as in NaCl-leaching solution. (orig.) With 13 refs., 22 figs [de

  10. Greater confinement disposal of high activity and special case wastes at the Nevada Test Site: A unified migration assessment approach

    International Nuclear Information System (INIS)

    Davis, P.A.; Olague, N.E.; Johnson, V.L.; Dickman, P.T.; O'Neill, L.J.

    1993-01-01

    The Department of Energy's Nevada Field Office has disposed of a small quantity of high activity and special case wastes using Greater Confinement Disposal facilities in Area 5 of the Nevada Test Site. Because some of these wastes are transuranic radioactive wastes, the Environmental Protection Agency standards for their disposal under 40 CFR Part 191 which requires a compliance assessment. In conducting the 40 CFR Part 191 compliance assessment, review of the Greater Confinement Disposal inventory revealed potentially land disposal restricted hazardous wastes. The regulatory options for disposing of land disposal restricted wastes consist of (1) treatment and monitoring, or (2) developing a no-migration petition. Given that the waste is already buried without treatment, a no-migration petition becomes the primary option. Based on a desire to minimize costs associated with site characterization and performance assessment, a single approach has been developed for assessing compliance with 40 CFR Part 191, DOE Order 5820.2A (which regulates low-level radioactive wastes contained in Greater Confinement Disposal facilities) and developing a no-migration petition. The approach consists of common points of compliance, common time frame for analysis, and common treatment of uncertainty. The procedure calls for conservative bias of modeling assumptions, including model input parameter distributions and adverse processes and events that can occur over the regulatory time frame, coupled with a quantitative treatment of data and parameter uncertainty. This approach provides a basis for a defensible regulatory decision. In addition, the process is iterative between modeling and site characterization activities, where the need for site characterization activities is based on a quantitative definition of the most important and uncertain parameters or assumptions

  11. Deep-ocean disposal of high-activity nuclear wastes: a conservative assessment of the seafood critical pathway

    International Nuclear Information System (INIS)

    Baxter, M.S.; Economides, B.

    1984-01-01

    This paper applies conventional 'worst-case' assumptions to modelling the effects of possible future disposal of high-activity wastes in the oceans. It otherwise uses previously published and generally accepted data to assess the possible intakes of the waste nuclides via consumption of seaweeds, molluscs, crustaceans, plankton and fish. Model-predicted intakes for critical groups generally exceed ICRP-recommended limits, with 244 Cm, 241 Am and 137 Cs being the most potentially hazardous nuclides. The various seafood consumption pathways are found to rank, in decreasing order of potential hazard, as seaweeds > molluscs > plankton > fish > crustaceans. (Auth.)

  12. Nuclear energy - Waste-packages activity measurement - Part.1: high-resolution gamma spectrometry in integral mode with open geometry

    International Nuclear Information System (INIS)

    2004-01-01

    ISO 14850:2004 describes a procedure for measurements of gamma-emitting radionuclide activity in homogeneous objects such as unconditioned waste (including process waste, dismantling waste, etc.), waste conditioned in various matrices (bitumen, hydraulic binder, thermosetting resins, etc.), notably in the form of 100 L, 200 L, 400 L or 800 L drums, and test specimens or samples, (vitrified waste), and waste packaged in a container, notably technological waste. It also specifies the calibration of the gamma spectrometry chain. The gamma energies used generally range from 0,05 MeV to 3 MeV. (authors)

  13. Safety and radiation protection aspects of the management of radioactive wastes of high level activity

    International Nuclear Information System (INIS)

    Candes, P.; Pradel, J.

    1977-01-01

    Appropriate consideration is given in France to safety and protection problems to be solved from the production up to the final disposal of radioactive wastes of high level activity. The first stage of the work consisted in emphasizing the various technical options. Different strategies appear to be possible, taking into account technical, political, and psychological difficulties. This results in evaluating the safety problems to be solved in the framework of those strategies. In this field, the main safety and protection principles do not differ from those which apply to other nuclear facilities. Nevertheless, duration is in most cases a quite different factor (thousands or millions of years). The question is then raised of evaluating the importance to be given to very remote consequences, both at philosophical and scientific levels. As a first result of those considerations, the application of the ''barrier'' concept is recommended. This concept is familiar to safety specialists. Different barriers for which particular problems are listed and evaluated, are defined. Another results with regard to radiation protection principles is to consider that if safety provisions should lead to a containment of radioactive products as efficient as possible, it would not be realistic to consider such a containment as absolute, in particular for disposal durations arising to thousands of years. It is therefore assumed that a limited radioactivity transfer should be taken into account, and its consequences for environment and man be calculated. This is especially true in the study of an appropriate site for final storage, and the study should necessarily include a detailed investigation of the retention characteristics of soil layers, and the implementation of appropriate models giving a sufficiently accurate evaluation of the consequences of transfers, including those related to the effect of various elements after their arrival into the biosphere. The authors review the

  14. FY-97 operations of the pilot-scale glass melter to vitrify simulated ICPP high activity sodium-bearing waste

    International Nuclear Information System (INIS)

    Musick, C.A.

    1997-11-01

    A 3.5 liter refractory-lined joule-heated glass melter was built to test the applicability of electric melting to vitrify simulated high activity waste (HAW). The HAW streams result from dissolution and separation of Idaho Chemical Processing Plant (ICPP) calcines and/or radioactive liquid waste. Pilot scale melter operations will establish selection criteria needed to evaluate the application of joule heating to immobilize ICPP high activity waste streams. The melter was fabricated with K-3 refractory walls and Inconel 690 electrodes. It is designed to be continuously operated at 1,150 C with a maximum glass output rate of 10 lbs/hr. The first set of tests were completed using surrogate HAW-sodium bearing waste (SBW). The melter operated for 57 hours and was shut down due to excessive melt temperatures resulting in low glass viscosity (< 30 Poise). Due to the high melt temperature and low viscosity the molten glass breached the melt chamber. The melter has been dismantled and examined to identify required process improvement areas and successes of the first melter run. The melter has been redesigned and is currently being fabricated for the second run, which is scheduled to begin in December 1997

  15. Migration of gases around a cell containing high-activity vitrified wastes during the operational phase

    International Nuclear Information System (INIS)

    Hoch, A.; Wendling, J.

    2010-01-01

    Document available in extended abstract form only. Andra is studying the Callovo-Oxfordian mud-stones, located at a depth of approximately 500 m beneath the borders of the Meuse and the Haute-Marne Departements, in order to assess the feasibility of constructing a repository for radioactive waste in this low-permeability geological formation. Several categories of waste are being considered for disposal in the geological repository. Of these, the high-activity, long-lived vitrified wastes (known as 'Waste C') are strongly exothermic and will be stored in cells that are about 40 m long, with a single access aperture from an access gallery. The cell wall will be supported by a steel lining. During the reversible operational phase (lasting at least 100 years), and as long as the access galleries are ventilated, the corrosion of the steel lining by oxygen in air will be important. It is necessary to determine the thickness of the lining corroded during this phase in order to be able to design the lining with an appropriate initial thickness to maintain integrity. In addition, at the end of the cell farthest from the ventilated access gallery, it is possible that the oxygen concentration will be so low that corrosion occurs under reducing (anaerobic) conditions. This chemical reaction produces hydrogen gas, which will migrate towards the access gallery. To maintain operational safety, it is necessary to define the hydrogen concentration within the cell and within the access gallery in order to evaluate the fire and/or explosion hazards (the hydrogen-oxygen mixture is highly flammable and/or explosive at certain concentrations). The objectives of this study were to estimate for the reversible operational phase (i.e. for 100 years): - The thickness of the steel lining required to ensure mechanical integrity. - Transfer rates (advection and / or diffusion) and gas concentrations (oxygen, hydrogen and water vapour) at all points in the annular void between the

  16. High level nuclear wastes

    International Nuclear Information System (INIS)

    Lopez Perez, B.

    1987-01-01

    The transformations involved in the nuclear fuels during the burn-up at the power nuclear reactors for burn-up levels of 33.000 MWd/th are considered. Graphs and data on the radioactivity variation with the cooling time and heat power of the irradiated fuel are presented. Likewise, the cycle of the fuel in light water reactors is presented and the alternatives for the nuclear waste management are discussed. A brief description of the management of the spent fuel as a high level nuclear waste is shown, explaining the reprocessing and giving data about the fission products and their radioactivities, which must be considered on the vitrification processes. On the final storage of the nuclear waste into depth geological burials, both alternatives are coincident. The countries supporting the reprocessing are indicated and the Spanish programm defined in the Plan Energetico Nacional (PEN) is shortly reviewed. (author) 8 figs., 4 tabs

  17. Disposal of high active nuclear fuel waste. A critical review of the Nuclear Fuel Safety (KBS) project on final disposal of vitrified high active nuclear fuel waste

    International Nuclear Information System (INIS)

    1978-01-01

    This report has been prepared by the Swedish Energy Commission's working group for Safety and Environment. The main contributions are by profs. Jan Rydberg of Chalmers University of Technology, Sweden and John W Winchester of Florida State University, USA. The aim of the report is to discuss weather the KBS-project fullfills the Swedish ''Stipulations Act'', that a absolutely safe way of disposing of the nuclear waste must have been demonstrated before any new reactors are allowed to be taken inot use. Rydberg and Winchester do not arrive at similar conclusions. (L.E.)

  18. CRNL active waste incinerator

    International Nuclear Information System (INIS)

    McQuade, D.W.

    1965-02-01

    At CRNL the daily collection of 1200 pounds of active combustible waste is burned in a refractory lined multi-chamber incinerator. Capacity is 500-550 pounds per hour; volume reduction 96%. Combustion gases are cooled by air dilution and decontaminated by filtration through glass bags in a baghouse dust collector. This report includes a description of the incinerator plant, its operation, construction and operating costs, and recommendations for future designs. (author)

  19. High-Level Radioactive Waste.

    Science.gov (United States)

    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…

  20. Low-Activity Radioactive Wastes

    Science.gov (United States)

    In 2003 EPA published an Advance Notice of Proposed Rulemaking (ANPR) to collect public comment on alternatives for disposal of waste containing low concentrations of radioactive material ('low-activity' waste).

  1. A model problem for restricted-data gamma ray emission tomography of highly active nuclear waste

    International Nuclear Information System (INIS)

    Cattle, Brian A.

    2007-01-01

    This paper develops the work of Cattle et al. [Cattle, B.A., Fellerman, A.S., West, R.M., 2004. On the detection of solid deposits using gamma ray emission tomography with limited data. Measurement Science and Technology 15, 1429-1439] by considering a generalization of the model employed therein. The focus of the work is the gamma ray tomographic analysis of high-level waste processing. The work in this paper considers a two-dimensional model for the measurement of gamma ray photon flux, as opposed to the previous one-dimensional analysis via the integrated Beer-Lambert law. The mathematical inverse problem that arises in determining physical quantities from the photon count measurements is tackled using Bayesian statistical methods that are implemented computationally using a Markov chain Monte Carlo (MCMC) approach. In a further new development, the effect of the degree of collimation of the detector on the reliability of the solutions is also considered

  2. Partitioning of actinides from high active waste solution of Purex origin counter-current extraction studies using TBP and CMPO

    International Nuclear Information System (INIS)

    Chitnis, R.R.; Dhami, P.S.; Gopalkrishnan, V.; Wattal, P.K.; Ramanujam, A.; Murali, M.S.; Mathur, J.N.; Bauri, A.K.; Chattopadhyay, S.

    2000-10-01

    A solvent extraction scheme has been formulated for the partitioning of actinides from Purex high level waste (HLW). The scheme is based on the results of earlier studies carried out with simulated waste solutions. In the present studies, the scheme was tested with high active waste (HAW) solution generated during the reprocessing of spent fuel from research reactors using laboratory scale mixer-settlers. The proposed process involved two-step extraction using tri-n-butyl phosphate (TBP) and octyl (phenyl)-N,N-diisobutylcarbamolylmethylphosphine oxide (CMPO). In the first step, uranium, neptunium and plutonium were removed from the waste using TBP as extractant. The minor actinides left in the raffinate were extracted using a mixture of CMPO and TBP in the second step. The results showed complete extraction of actinides from the waste solution. Plutonium and neptunium extracted in TBP, were stripped together using a mixture of hydrogen peroxide and ascorbic acid in 2 M nitric acid medium, leaving uranium in the organic phase. Uranium can later be stripped using dilute nitric acid. Actinides extracted in CMPO-TBP phase were stripped using a mixture of formic acid, hydrazine, hydrate and citric acid. The stripping was quantitative in both the stripping runs. An additional extraction step for the preferential recovery of uranium, neptunium and plutonium from the waste solution using TBP is a modification over the conventional Truex process. Selective stripping of neptunium and plutonium from large quantities of uranium. The extraction of uranium using TBP eliminates the possibility of third phase and undesired loading of CMPO-TBP in the following step. Use of citrate-containing strippant allows the recovery of actinides from loaded CMPO-TBP mixture without causing any reflux of the actinides during stripping. The process has been developed with due consideration to minimising the generation of secondary wastes. The proposed strippants are effective even in presence of

  3. High-level-waste immobilization

    International Nuclear Information System (INIS)

    Crandall, J.L.

    1982-01-01

    Analysis of risks, environmental effects, process feasibility, and costs for disposal of immobilized high-level wastes in geologic repositories indicates that the disposal system safety has a low sensitivity to the choice of the waste disposal form

  4. Technical position on items and activities in the high-level waste geologic repository program subject to quality assurance requirements

    International Nuclear Information System (INIS)

    Duncan, A.B.; Bilhorn, S.G.; Kennedy, J.E.

    1988-04-01

    This document provides guidance on how to identify items and activities subject to Quality Assurance in the high-level nuclear waste repository program for pre-closure and post-closure phases of the repository. In the pre-closure phase, structures, systems and components essential to the prevention or mitigation of an accident that could result in an off-site radiation dose of 0.5rem or greater are termed ''important to safety''. In the post-closure phase, the barriers which are relied on to meet the containment and isolation requirements are defined as ''important to waste isolation''. These structures, systems, components, and barriers, and the activities related to their characterization, design, construction, and operation are required to meet quality assurance (QA) criteria to provide confidence in the performance of the geologic repository. The list of structures, systems, and components important to safety and engineered barriers important to waste isolation is referred to as the ''Q-List'' and lies within the scope of the QA program. 10 refs

  5. Performance testing of high specific activity waste forms per 10 CFR Part 61

    International Nuclear Information System (INIS)

    Neilson, R.M. Jr.; McConnell, J.W. Jr.; Rogers, R.D.

    1987-03-01

    The Low-Level Waste Data Base Development - EPICOR-II Resin/Liner Investigation Program of the US Nuclear Regulatory Commission (NRC) is obtaining information on radioactive waste by means of NRC-prescribed tests in a disposal environment. This paper describes the resin solidification task of that program and includes the current test status and results. 28 refs., 5 figs., 6 tabs

  6. Activated Carbon Fibers with Hierarchical Nanostructure Derived from Waste Cotton Gloves as High-Performance Electrodes for Supercapacitors.

    Science.gov (United States)

    Wei, Chao; Yu, Jianlin; Yang, Xiaoqing; Zhang, Guoqing

    2017-12-01

    One of the most challenging issues that restrict the biomass/waste-based nanocarbons in supercapacitor application is the poor structural inheritability during the activating process. Herein, we prepare a class of activated carbon fibers by carefully selecting waste cotton glove (CG) as the precursor, which mainly consists of cellulose fibers that can be transformed to carbon along with good inheritability of their fiber morphology upon activation. As prepared, the CG-based activated carbon fiber (CGACF) demonstrates a surface area of 1435 m 2  g -1 contributed by micropores of 1.3 nm and small mesopores of 2.7 nm, while the fiber morphology can be well inherited from the CG with 3D interconnected frameworks created on the fiber surface. This hierarchically porous structure and well-retained fiber-like skeleton can simultaneously minimize the diffusion/transfer resistance of the electrolyte and electron, respectively, and maximize the surface area utilization for charge accumulation. Consequently, CGACF presents a higher specific capacitance of 218 F g -1 and an excellent high-rate performance as compared to commercial activated carbon.

  7. Activated Carbon Fibers with Hierarchical Nanostructure Derived from Waste Cotton Gloves as High-Performance Electrodes for Supercapacitors

    Science.gov (United States)

    Wei, Chao; Yu, Jianlin; Yang, Xiaoqing; Zhang, Guoqing

    2017-06-01

    One of the most challenging issues that restrict the biomass/waste-based nanocarbons in supercapacitor application is the poor structural inheritability during the activating process. Herein, we prepare a class of activated carbon fibers by carefully selecting waste cotton glove (CG) as the precursor, which mainly consists of cellulose fibers that can be transformed to carbon along with good inheritability of their fiber morphology upon activation. As prepared, the CG-based activated carbon fiber (CGACF) demonstrates a surface area of 1435 m2 g-1 contributed by micropores of 1.3 nm and small mesopores of 2.7 nm, while the fiber morphology can be well inherited from the CG with 3D interconnected frameworks created on the fiber surface. This hierarchically porous structure and well-retained fiber-like skeleton can simultaneously minimize the diffusion/transfer resistance of the electrolyte and electron, respectively, and maximize the surface area utilization for charge accumulation. Consequently, CGACF presents a higher specific capacitance of 218 F g-1 and an excellent high-rate performance as compared to commercial activated carbon.

  8. Reducing the likelihood of future human activities that could affect geologic high-level waste repositories

    International Nuclear Information System (INIS)

    1984-05-01

    The disposal of radioactive wastes in deep geologic formations provides a means of isolating the waste from people until the radioactivity has decayed to safe levels. However, isolating people from the wastes is a different problem, since we do not know what the future condition of society will be. The Human Interference Task Force was convened by the US Department of Energy to determine whether reasonable means exist (or could be developed) to reduce the likelihood of future human unintentionally intruding on radioactive waste isolation systems. The task force concluded that significant reductions in the likelihood of human interference could be achieved, for perhaps thousands of years into the future, if appropriate steps are taken to communicate the existence of the repository. Consequently, for two years the task force directed most of its study toward the area of long-term communication. Methods are discussed for achieving long-term communication by using permanent markers and widely disseminated records, with various steps taken to provide multiple levels of protection against loss, destruction, and major language/societal changes. Also developed is the concept of a universal symbol to denote Caution - Biohazardous Waste Buried Here. If used for the thousands of non-radioactive biohazardous waste sites in this country alone, a symbol could transcend generations and language changes, thereby vastly improving the likelihood of successful isolation of all buried biohazardous wastes

  9. Reducing the likelihood of future human activities that could affect geologic high-level waste repositories

    Energy Technology Data Exchange (ETDEWEB)

    1984-05-01

    The disposal of radioactive wastes in deep geologic formations provides a means of isolating the waste from people until the radioactivity has decayed to safe levels. However, isolating people from the wastes is a different problem, since we do not know what the future condition of society will be. The Human Interference Task Force was convened by the US Department of Energy to determine whether reasonable means exist (or could be developed) to reduce the likelihood of future human unintentionally intruding on radioactive waste isolation systems. The task force concluded that significant reductions in the likelihood of human interference could be achieved, for perhaps thousands of years into the future, if appropriate steps are taken to communicate the existence of the repository. Consequently, for two years the task force directed most of its study toward the area of long-term communication. Methods are discussed for achieving long-term communication by using permanent markers and widely disseminated records, with various steps taken to provide multiple levels of protection against loss, destruction, and major language/societal changes. Also developed is the concept of a universal symbol to denote Caution - Biohazardous Waste Buried Here. If used for the thousands of non-radioactive biohazardous waste sites in this country alone, a symbol could transcend generations and language changes, thereby vastly improving the likelihood of successful isolation of all buried biohazardous wastes.

  10. Galvanic corrosion evaluation of high activity nuclear waste container metals components

    International Nuclear Information System (INIS)

    Semino, C.J.

    1990-04-01

    The final disposal container for vitrified high-level waste is assumed to have three metallic layers: a stainless steel inner layer, and external one of a metal to be selected and a thick lead layer (10 cm) in the middle. As design limit, the container shall act as an engineering barrier, granting the isolation of the radionuclides for approximately 1000 years. Preliminary titanium-lead galvanic couple tests showed that titanium behaved always as a cathode in the galvanic couple, promoting the galvanic corrosion of lead. This corrosion study focused on the behaviour of lead-AISI 304 stainless steel and lead-carbon steel (SAE 1010 and 1020) galvanic couples with different area relationships, temperature and media composition. High purity lead (99,999%) and commercial lead (99,9%) were used for galvanic couples tests. Tests were performed at 75, 50, 45 and 40 deg. C. Test solution was either synthetic groundwater, a suspension of 10% bentonite in groundwater, or synthetic sea water. The synthetic sea water was used at 100, 50 and 25% concentration by dilution with distilled water. Tests with lead-304 stainless steel galvanic couples showed that lead always behaves as an anode, corroding preferentially. Very low lead corrosion rates were found in lead-carbon steel galvanic couple in 10% bentonite suspension in synthetic groundwater test at 75 deg. C. An increase of carbon content in steel has very little influence on steel corrosion rate. Commercial lead has a higher corrosion rate and presented a more pronounced attack than high purity lead. Its corrosion rate is at least twice when lead-carbon steel area relationship increases from 1:10 to 1:40. There are higher steel corrosion rates in sea water than in groundwater. Lead behaves as a cathode to the end of the test. 8 refs, 85 figs, 10 tabs

  11. Conditioning of high activity solid waste: fuel claddings and dissolution residues

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    This chapter reports on experimental studies of embedding into matrix material, the melting and conversion of zircaloy, and waste properties and characterization. Methods are developed for embedding the waste scrap into a solid and resistant matrix material in order to confine the radioactivity and to prevent it from dispersion. The matrix materials investigated are lead alloys, ceramics and compacted graphite or aluminium powder. The treatment of fuel hulls by melting or chemical conversion is described. Cemented hulls are characterized and the pyrophoricity of zircaloy fines is investigated. Topics considered include the embedding of hulls into graphite and aluminium, the embedding of hulls and dissolution residues into alumino-ceramics, the solidification of alpha-bearing wastes into a ceramic matrix, and the conditioning of cladding waste by eutectoidic melting and by embedding in glass

  12. Radiological impact of high activity wastes disposal in a granitic rock

    International Nuclear Information System (INIS)

    Oliveira, A.A.; Palacios, E.

    1987-01-01

    This work analyzes, by a simplified model, the radiological impact due to radioactive wastes release when engineering and geologic barriers individually fail. Doses are calculated resulting from the individuals of a hypothetical critical group and the potential collective dosis for each one of the considered situations. (Author) [es

  13. Waste treatment activities incineration

    International Nuclear Information System (INIS)

    Weber, D.A.

    1985-01-01

    The waste management policy at SRP is to minimize waste generation as much as possible and detoxify and/or volume reduce waste materials prior to disposal. Incineration is a process being proposed for detoxification and volume reduction of combustion nonradioactive hazardous, low-level mixed and low-level beta-gamma waste. Present operation of the Solvent Burner Demonstration reduces the amount of solid combustible low-level beta-gamma boxed waste disposed of by shallow land burial by approximately 99,000 ft 3 per year producing 1000 ft 3 per year of ash and, by 1988, will detoxify and volume reduce 150,000 gallons or organic Purex solvent producing approximately 250 ft 3 of ash per year

  14. Combining high-rate aerobic wastewater treatment with anaerobic digestion of waste activated sludge at a pulp and paper mill.

    Science.gov (United States)

    Magnusson, Björn; Ekstrand, Eva-Maria; Karlsson, Anna; Ejlertsson, Jörgen

    2018-05-01

    The activated sludge process within the pulp and paper industry is generally run to minimize the production of waste activated sludge (WAS), leading to high electricity costs from aeration and relatively large basin volumes. In this study, a pilot-scale activated sludge process was run to evaluate the concept of treating the wastewater at high rate with a low sludge age. Two 150 L containers were used, one for aeration and one for sedimentation and sludge return. The hydraulic retention time was decreased from 24 hours to 7 hours, and the sludge age was lowered from 12 days to 2-4 days. The methane potential of the WAS was evaluated using batch tests, as well as continuous anaerobic digestion (AD) in 4 L reactors in mesophilic and thermophilic conditions. Wastewater treatment capacity was increased almost four-fold at maintained degradation efficiency. The lower sludge age greatly improved the methane potential of the WAS in batch tests, reaching 170 NmL CH 4 /g VS at a sludge age of 2 days. In addition, the continuous AD showed a higher methane production at thermophilic conditions. Thus, the combination of high-rate wastewater treatment and AD of WAS is a promising option for the pulp and paper industry.

  15. Reference concepts for the final disposal of LWR spent fuel and other high activity wastes in Spain

    International Nuclear Information System (INIS)

    Huertas, F.; Ulibarri, A.

    1993-01-01

    Studies over the last three years have been recently concluded with the selection of a reference repository concept for the final disposal of spent fuel and other high activity wastes in deep geological formations. Two non-site specific preliminary designs, at a conceptual level, have been developed; one considers granite as the host rock and the other rock salt formations. The Spanish General Radioactive Waste Program also considers clay as a potential host rock for HLW deep disposal; conceptualization for a deep repository in clay is in the initial phase of development. The salt repository concept contemplates the disposal of the HLW in self-shielding casks emplaced in the drifts of an underground facility, excavated at a depth of 850 m in a bedded salt formation. The Custos Type I(7) cask admits up to seven intact PWR fuel assemblies or 21 of BWR type. The final repository facilities are planned to accept a total of 20,000 fuel assemblies (PWR and BWR) and 50 vitrified waste canisters over a period of 25 years. The total space needed for the surface facilities amounts to 322,000 m 2 , including the rock salt dump. The space required for the underground facilities amounts to 1.2 km 2 , approximately. The granite repository concept contemplates the disposal of the HLW in carbon steel canisters, embedded in a 0.75 m thick buffer of swelling smectite clay, in the drifts of an underground facility, excavated at a depth of 55 m in granite. Each canister can host 3 PWR or 9 BWR fuel assemblies. For this concept the total number of canisters needed amounts to 4,860. The space required for the surface and underground facilities is similar to that of the salt concept. The technical principles and criteria used for the design are discussed, and a description of the repository concept is presented

  16. Criticality Safety Problems Related to Storage of Highly Active Liquid Waste

    International Nuclear Information System (INIS)

    Amin, E.

    1999-01-01

    The geometries of liquid waste storage tanks are not generally safe against criticality. Normally, this does not cause problems as fissile materials exist in nitric acid solution only as depleted uranium or in insignificant concentration of the originally reprocessed inventory of plutonium. However, if sedimentation of solid particles would occur, the deposited material would cause criticality safety problems. Particularly, non-horizontal installation of the storage tanks would increase the Eigen value. The effect of the storage tank inclination and the presence of transplutonium elements on the criticality safety are investigated using the NCNSRC code packages. The results are compared well with a similar German published results

  17. Status of defense radioactive waste disposal activities

    International Nuclear Information System (INIS)

    Wade, T.W.

    1988-01-01

    The Office of Defense Programs, U.S. Department of Energy, is responsible for the production of nuclear weapons and materials for national defense. As a byproduct to their activities, nuclear production facilities have generated, and will continue to generate, certain radioactive, hazardous, or mixed wastes that must be managed and disposed of in a safe and cost-effective manner. Compliance with all applicable Federal and State regulations is required. This paper describes the principal elements that comprise Defense Programs' approach to waste management and disposal. The status of high-level, transuranic, and low-level radioactive waste disposal is set forth. Defense Programs' activities in connection with the environmental restoration of inactive facilities and with the safe transport of waste materials are summarized. Finally, the principal challenges to realizing the goals set for the defense waste program are discussed in terms of regulatory, public acceptance, technical, and budget issues

  18. Preparation and characterization of high-surface-area activated carbon fibers from silkworm cocoon waste for congo red adsorption

    International Nuclear Information System (INIS)

    Li, Jia; Ng, Dickon H.L.; Song, Peng; Kong, Chao; Song, Yi; Yang, Ping

    2015-01-01

    Herein, we report the preparation of activated carbon fibers from silkworm cocoon waste via the combination of (NH 4 ) 2 HPO 4 -pretreatment and KOH activation. The morphology, phase structure and surface chemistry constitute of the obtained ACFs were characterized by X-ray diffraction, IR spectroscopy, Micro Raman spectroscopy, X-ray photoelectron spectroscopy, scanning electron microscopy, thermal analysis and N 2 adsorption–desorption isotherm. The effects of various factors such as the concentration of (NH 4 ) 2 HPO 4 and the activation time of KOH were also evaluated. These results demonstrated that the synthesized ACFs retained the fibrous morphology of silkworm cocoon waste, and exhibited highly defective graphite layer structure. A large amount of surface oxygen-containing functional groups were found on the ACFs surface. The obtained samples exhibited high BET surface areas ranging from 1153 to 2797 m 2  g −1 , total pore volumes of 0.64–1.74 cm 3  g −1 with micropore volume fractions between 75.2 and 93.6%. In addition, we also evaluated the congo red (CR) adsorption performance of the obtained ACFs. The CR adsorption fitted well to the pseudo-second-order kinetic model. Adsorption isotherm data indicated that the adsorption of CR onto ACFs was monolayer adsorption which followed well the Langmuir isotherm model. The maximum adsorption capacity of CR was 512 g kg −1 . The mechanism of the adsorption process was also described from the intraparticle diffusion model. - Highlights: • A new biomass fibroin precursor for activated carbon fibers (ACFs) was proposed. • High specific surface area (2797 m 2  g −1 ) and total pore volume (1.74 cm 3  g −1 ) were obtained. • The original fibrous structure of raw silkworm cocoons was retained in the ACF product. • Congo red maximum monolayer adsorption capacity of our ACF product was up to 1100 g kg −1

  19. CASTOR(r) and CONSTOR(r) type transport and storage casks for spent fuel and high active waste

    International Nuclear Information System (INIS)

    Kuehne, B.; Sowa, W.

    2002-01-01

    The German company GNB has developed, tested, licensed, fabricated, loaded, transported and stored a large number of casks for spent fuel and high-level waste. CASTOR(r) casks are used at 18 sites on three continents. Spent fuel assemblies of the types PWR, BWR, VVER, RBMK, MTR and THTR as well as vitrified high active waste (HAW) containers are stored in these kinds of casks. More than 600 CASTOR(r) casks have been loaded for long-term storage. The two decades of storage have shown that the basic requirements, which are safe confinement, criticality safety, sufficient shielding and appropriate heat transfer have been fulfilled in each case. There is no indication that problems will arise in the future. Of course, the experience of 20 years has resulted in improvements of the cask design. One basic improvement is GNB's development since the mid 1990s of a sandwich cask design using heavy concrete and steel as basic materials, for economical and technical reasons. This CONSTOR(r) cask concept also fulfils all design criteria for transport and storage given by the IAEA recommendations and national authorities. By May 2002 40 CONSTOR(r) casks had been delivered and 15 had been successfully loaded and stored. In this paper the different types of casks are presented. Experiences gained during the large number of cask loadings and more than 4000 cask-years of storage will be summarised. The presentation of recent and future development shows the optimisation potential of the CASTOR(r) and CONSTOR(r) cask families for safe and economical management of spent fuel. (author)

  20. Treatment of solid non-active wastes

    International Nuclear Information System (INIS)

    Chmielewska, E.

    2008-01-01

    In this part of the text-book treatment of solid non-active wastes is described. This part consist of following chapters: (1) Law on wastes; (2) Present situation in waste management; (3) Strategic tendencies of waste management; (4) Incineration (disposal of solid wastes); (5) Disposal; (6) Composting; (7) Treatment of sludge from sewage clarification plant; (8) Biodegradation; (9) Recycling of wastes (assessing of secondary raw materials). Legal aspects of treatment of solid non-active wastes is presented

  1. Implications of the use of low-pH cementitious materials in high activity radioactive waste repositories

    International Nuclear Information System (INIS)

    Garcia Calvo, J.L.; Alonso, M.C.; Fernandez Luco, L.; Hidalgo, A.; Sanchez, M.

    2008-01-01

    One of the most accepted engineering construction concepts for high radioactive nuclear waste of underground repositories considers the use of low pH cementitious materials, in order to avoid the formation of an alkaline plume fluid which perturbs one of the engineered barriers of the repository, the bentonite. The accepted solution to maintain the bentonite stability, which is function of the pH, is to develop cementitious materials that generate pore waters with pH ≤ 11, because the corrosion velocity of the clay is significantly reduced below this value. The IETcc-CSIC has focused the research activity on low-pH cementitious materials using two cements: Ordinary Portland Cements (OPC) and Calcium Aluminates Cements (CAC). In both cases, the achievement of a low-pH environment implies the use of high content of mineral admixtures to prepare the binder. Obviously, the inclusion of high contents of mineral admixtures in the cement formulation modifies most of the concrete 'standard' properties and the microstructure of the obtained cement products. When designing a concrete based on low-pH binders, not only the functional requirements have to be reached but also the modifications of the basic properties of the concrete must be taken into account. Besides, due to the location and the long service life of this type of products, their durability properties must be also guaranteed. This paper deals with the procedure followed in the design of a specific application of low pH cements; for instance, the shotcrete plug fabrication. The challenge of this type of use (shotcreting) is more complex taking into account that requires the employment of additives that must be compatible with the concrete mixture. Furthermore, their effectiveness must be assured without increase the pH above the admissible levels. Therefore, their compatibility with admixtures is tested in the present work. The compliance of the requirements for a shotcrete plug was evaluated at laboratory scale

  2. Prediction for the high-level alpha-active waste to be generated by nuclear power stations in the Member States of the European Communities

    International Nuclear Information System (INIS)

    Schmidt, E.

    1977-04-01

    Starting with a forecast for the nuclear power generating capacity to be installed in the Member States of the European Communities before the end of this century, a prediction is made of the annual production of high-level alpha-active waste from reprocessing plants and the corresponding accumulation up to the year 2000. The isotopic composition of the alpha-active waste from individual reactor types was calculated and an estimation of the influence of recycling plutonium through light water reactors on the produced quantity of higher actinides is made

  3. Optimizing High Level Waste Disposal

    International Nuclear Information System (INIS)

    Dirk Gombert

    2005-01-01

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

  4. Waste Away--Information and Activities for Investigating Trash Problems and Solutions for Upper Elementary and Junior High School Students.

    Science.gov (United States)

    Vermont Inst. of Natural Science, Woodstock.

    The solid waste problem is so pervasive that it can no longer be ignored. The purpose of this set of materials is to encourage a lifestyle that includes the use of the three R's: Reduce, Reuse, Recycle. It was designed to educate upper elementary and junior high school students who may educate their classmates, families, and community about the…

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

    Science.gov (United States)

    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…

  6. Radioprotection and physical surveillance during activities of liquid wastes of high and low activity in italian ITREC plant; Sorveglianza fisica di radioprotezione durante la prima campagna di rifiuti liquidi radioattivi nell`Impianto SIRTE

    Energy Technology Data Exchange (ETDEWEB)

    Petagna, Edoardo; Tortorelli, Pietro [ENEA, Centro Richerche Trisaia, Rotondella, Matera (Italy). Dipt. Ambiente

    1997-03-01

    Many studies were made in ITREC Plant, located in ENEA - Trisaia Research Center, in the field of the nuclear fuel reprocessing, in the past years. During these activities liquid wastes of high and low activity were yielded and stored in the special area of tanks named Waste-1. In order to condition the low activity liquid wastes, essentially fission products, beta and gamma emitters, was built the SIRTE Plant (Integrate System for the Raise and Effluents Treatment) based on cementation process. In the present work, the radiological monitoring performed within the plant during the first campaign of cementation, is showed.

  7. Final disposal of spent fuels and high activity waste: the European model for a shared regional repository. Part 3

    International Nuclear Information System (INIS)

    Herscovich de Pahissa, Marta

    2009-01-01

    Geological disposal is a essential element and the only available approach to the management strategy for spent nuclear fuel and high level radioactive waste from reprocessing and also for other long-lived waste from nuclear technology applications. It is technically feasible and offers the required long term safety. The growth of existing nuclear programmes and the expansion of nuclear technology to new countries will have effects on the fuel cycle because of the increased concern on proliferation and waste management. The crucial task is to ensure that all countries that use nuclear energy now or will do it in the future, have defined and agreed safety and security standards for all facilities and a credible waste disposal strategy , accepted by the community, when this become necessary. Multinational cooperation on essential aspects of fuel cycle, particularly the geological disposal, is required for several countries with relatively small nuclear energy programmes or small quantities of radioactive waste. For these countries, that can be in different stages of development, the possibility to share a deep geological repository could be convenient. The European Union SAPIERR project is described in this paper as an example of a regional multinational cooperation. (author) [es

  8. Simultaneous production of l-lactic acid with high optical activity and a soil amendment with food waste that demonstrates plant growth promoting activity.

    Science.gov (United States)

    Kitpreechavanich, Vichien; Hayami, Arisa; Talek, Anfal; Chin, Clament Fui Seung; Tashiro, Yukihiro; Sakai, Kenji

    2016-07-01

    A unique method to produce highly optically-active l-lactic acid and soil amendments that promote plant growth from food waste was proposed. Three Bacillus strains Bacillus subtilis KBKU21, B. subtilis N3-9 and Bacillus coagulans T27, were used. Strain KBKU21 accumulated 36.9 g/L l-lactic acid with 95.7% optical activity and 98.2% l-lactic acid selectivity when fermented at 43°C for 84 h in a model kitchen refuse (MKR) medium. Residual precipitate fraction (anaerobically-fermented MKR (AFM) compost) analysis revealed 4.60%, 0.70% and 0.75% of nitrogen (as N), phosphorous (as P2O5), and potassium (as K2O), respectively. Additionally, the carbon to nitrogen ratio decreased from 13.3 to 10.6. AFM compost with KBKU21 promoted plant growth parameters, including leaf length, plant height and fresh weight of Brassica rapa (Komatsuna), than that by chemical fertilizers or commercial compost. The concept provides an incentive for the complete recycling of food waste, contributing towards a sustainable production system. Copyright © 2016 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

  9. Quantitative determination of total cesium in highly active liquid waste by using liquid electrode plasma optical emission spectrometry.

    Science.gov (United States)

    Do, Van-Khoai; Yamamoto, Masahiko; Taguchi, Shigeo; Takamura, Yuzuru; Surugaya, Naoki; Kuno, Takehiko

    2018-06-01

    A sensitive analytical method for determination of total cesium (Cs) in highly active liquid waste (HALW) by using modified liquid electrode plasma optical emission spectrometry (LEP-OES) is developed in this study. The instrument is modified to measure radioactive samples in a glove box. The effects of important factors, including pulsed voltage sequence and nitric acid concentration, on the emission of Cs are investigated. The limit of detection (LOD) and limit of quantification (LOQ) are 0.005 mg/L and 0.02 mg/L, respectively. The achieved LOD is one order lower than that of recently developed spectroscopic methods using liquid discharge plasma. The developed method is validated by subjecting a simulated HALW sample to inductively coupled plasma mass spectrometry (ICP-MS). The recoveries obtained from a spike-and-recovery test are 96-102%, implying good accuracy. The method is successfully applied to the quantification of Cs in a real HALW sample at the Tokai reprocessing plant in Japan. Apart from dilution and filtration of the HALW sample, no other pre-treatment process is required. The results agree well with the values obtained using gamma spectrometry. The developed method offers a reliable technique for rapid analysis of total Cs in HALW samples. Copyright © 2018 Elsevier B.V. All rights reserved.

  10. Ramifications of defining high-level waste

    International Nuclear Information System (INIS)

    Wood, D.E.; Campbell, M.H.; Shupe, M.W.

    1987-01-01

    The Nuclear Regulatory Commission (NRC) is considering rule making to provide a concentration-based definition of high-level waste (HLW) under authority derived from the Nuclear Waste Policy Act (NWPA) of 1982 and the Low Level Waste Policy Amendments Act of 1985. The Department of Energy (DOE), which has the responsibility to dispose of certain kinds of commercial waste, is supporting development of a risk-based classification system by the Oak Ridge National Laboratory to assist in developing and implementing the NRC rule. The system is two dimensional, with the axes based on the phrases highly radioactive and requires permanent isolation in the definition of HLW in the NWPA. Defining HLW will reduce the ambiguity in the present source-based definition by providing concentration limits to establish which materials are to be called HLW. The system allows the possibility of greater-confinement disposal for some wastes which do not require the degree of isolation provided by a repository. The definition of HLW will provide a firm basis for waste processing options which involve partitioning of waste into a high-activity stream for repository disposal, and a low-activity stream for disposal elsewhere. Several possible classification systems have been derived and the characteristics of each are discussed. The Defense High Level Waste Technology Lead Office at DOE - Richland Operations Office, supported by Rockwell Hanford Operations, has coordinated reviews of the ORNL work by a technical peer review group and other DOE offices. The reviews produced several recommendations and identified several issues to be addressed in the NRC rule making. 10 references, 3 figures

  11. Deep reversible storage. Design options for the storage in deep geological formation - High-medium activity, long living wastes 2009 milestone

    International Nuclear Information System (INIS)

    2010-09-01

    This report aims at presenting a synthesis of currently studied solutions for the different components of the project of deep geological radioactive waste storage centre. For each of these elements, the report indicates the main operational objectives to be taken into account in relationship with safety functions or with reversibility. It identifies the currently proposed design options, presents the technical solutions (with sometime several possibilities), indicates industrial references (in the nuclear sector, in underground works) and comments results of technological tests performed by the ANDRA. After a description of functionalities and of the overall organisation of storage components, the different following elements and aspects are addressed: surface installations, underground architecture, parcel transfer between the surface and storage cells, storage container for medium-activity long-life (MAVL) waste, storage cell for medium-activity long-life waste, handling of MAVL parcels in storage cells, storage container for high-activity (HA) waste, storage cell for HA waste, handling of HA parcels in storage cells, and works for site closing

  12. SECONDARY WASTE MANAGEMENT STRATEGY FOR EARLY LOW ACTIVITY WASTE TREATMENT

    Energy Technology Data Exchange (ETDEWEB)

    TW, CRAWFORD

    2008-07-17

    This study evaluates parameters relevant to River Protection Project secondary waste streams generated during Early Low Activity Waste operations and recommends a strategy for secondary waste management that considers groundwater impact, cost, and programmatic risk. The recommended strategy for managing River Protection Project secondary waste is focused on improvements in the Effiuent Treatment Facility. Baseline plans to build a Solidification Treatment Unit adjacent to Effluent Treatment Facility should be enhanced to improve solid waste performance and mitigate corrosion of tanks and piping supporting the Effiuent Treatment Facility evaporator. This approach provides a life-cycle benefit to solid waste performance and reduction of groundwater contaminants.

  13. Technology development activities supporting tank waste remediation

    International Nuclear Information System (INIS)

    Bonner, W.F.; Beeman, G.H.

    1994-06-01

    This document summarizes work being conducted under the U.S. Department of Energy's Office of Technology Development (EM-50) in support of the Tank Waste Remediation System (TWRS) Program. The specific work activities are organized by the following categories: safety, characterization, retrieval, barriers, pretreatment, low-level waste, and high-level waste. In most cases, the activities presented here were identified as supporting tank remediation by EM-50 integrated program or integrated demonstration lead staff and the selections were further refined by contractor staff. Data sheets were prepared from DOE-HQ guidance to the field issued in September 1993. Activities were included if a significant portion of the work described provides technology potentially needed by TWRS; consequently, not all parts of each description necessarily support tank remediation

  14. The management of high-level radioactive wastes

    International Nuclear Information System (INIS)

    Lennemann, Wm.L.

    1979-01-01

    The definition of high-level radioactive wastes is given. The following aspects of high-level radioactive wastes' management are discussed: fuel reprocessing and high-level waste; storage of high-level liquid waste; solidification of high-level waste; interim storage of solidified high-level waste; disposal of high-level waste; disposal of irradiated fuel elements as a waste

  15. High-level radioactive wastes

    International Nuclear Information System (INIS)

    Grissom, M.C.

    1982-10-01

    This bibliography contains 812 citations on high-level radioactive wastes included in the Department of Energy's Energy Data Base from January 1981 through July 1982. These citations are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number

  16. Disposal of high level radioactive wastes in geological formations

    International Nuclear Information System (INIS)

    Martins, L.A.M.; Carvalho Bastos, J.P. de

    1978-01-01

    The disposal of high-activity radioactive wastes is the most serious problem for the nuclear industry. Among the solutions, the disposal of wastes in approriated geological formations is the most realistic and feasible. In this work the methods used for geological disposal, as well as, the criteria, programs and analysis for selecting a bite for waste disposal are presented [pt

  17. Other-than-high-level waste

    International Nuclear Information System (INIS)

    Bray, G.R.

    1976-01-01

    The main emphasis of the work in the area of partitioning transuranic elements from waste has been in the area of high-level liquid waste. But there are ''other-than-high-level wastes'' generated by the back end of the nuclear fuel cycle that are both large in volume and contaminated with significant quantities of transuranic elements. The combined volume of these other wastes is approximately 50 times that of the solidified high-level waste. These other wastes also contain up to 75% of the transuranic elements associated with waste generated by the back end of the fuel cycle. Therefore, any detailed evaluation of partitioning as a viable waste management option must address both high-level wastes and ''other-than-high-level wastes.''

  18. Immobilization of high activity nuclear wastes in sintered glass. Fabrication of blocks at semi-industrial scale by hot pressing technique

    International Nuclear Information System (INIS)

    Russo, D.O.; Messi, N.B.; Riquelme, R.; Sterba, M.E.; Audero, M.A.

    1990-01-01

    The sintering process under glass pressure has been studied as an alternative of melting with the aim of obtaining a monolytic material apt to preserve the high activity nuclear wastes. Different properties of the products obtained have been evaluated where the material is selected on the basis of the results attained. The purpose of this work is the equipment development and the process adjusting for the blocks obtainment. (Author) [es

  19. Final disposal of spent fuels and high activity waste: status and trends in the world. Part 2

    International Nuclear Information System (INIS)

    Herscovich de Pahissa, Marta

    2008-01-01

    The proper management of spent fuel arising from nuclear power production is a key issue for the sustainable development of nuclear energy. Some countries have adopted reprocessing of spent fuel and part of them has continued to develop and improve closed fuel cycle technologies; some other countries have adopted a direct final disposal. The objective in this article is to provide an update on the latest development in the world related with the geological disposal of spent nuclear fuel and high level wastes. (author) [es

  20. Enzyme Activities in Waste Water and Activated Sludge

    DEFF Research Database (Denmark)

    Nybroe, Ole; Jørgensen, Per Elberg; Henze, Mogens

    1992-01-01

    The purpose of the present study was to evaluate the potential of selected enzyme activity assays to determine microbial abundance and heterotrophic activity in waste water and activated sludge. In waste water, esterase and dehydrogenase activities were found to correlate with microbial abundance...... measured as colony forming units of heterotrophic bacteria. A panel of four enzyme activity assays, α-glucosidase, alanine-aminopeptidase, esterase and dehydrogenase were used to characterize activated sludge and anaerobic hydrolysis sludge from a pilot scale plant. The enzymatic activity profiles were...... distinctly different, suggesting that microbial populations were different, or had different physiological properties, in the two types of sludge. Enzyme activity profiles in activated sludge from four full-scale plants seemed to be highly influenced by the composition of the inlet. Addition of hydrolysed...

  1. Oxygen- and nitrogen-co-doped activated carbon from waste particleboard for potential application in high-performance capacitance

    International Nuclear Information System (INIS)

    Shang, Tong-Xin; Ren, Ru-Quan; Zhu, Yue-Mei; Jin, Xiao-Juan

    2015-01-01

    Graphical abstract: All electrodes showed excellent capacitance and retention versus discharge current density from 0.05 to 5 A/g. - Abstract: Oxygen- and nitrogen-co-doped activated carbons were obtained from phosphoric acid treated nitrogen-doped activated carbons which were prepared from waste particleboard bonded with urea-formaldehyde resin adhesives. The activated carbon samples obtained were tested as supercapacitors in two-electrode cell and extensive wetting 7 M KOH electrolytes. Their structural properties and surface chemistry, before the electrical testing, were investigated using elemental analysis, X-ray photoelectron spectroscopy, scanning electron microscopy, X-ray diffraction, Raman spectra, and adsorption of nitrogen. Activated carbon treated by 4 M phosphoric acid of the highest capacitance (235 F/g) was measured in spite of a relatively lower surface (1360 m 2 /g) than that of the activated carbon treated by 2 M phosphoric acid (1433 m 2 /g). The surface chemistry, and especially oxygen- and nitrogen-containing functional groups, was found of paramount importance for the capacitive behavior and for the effective pore space utilization by the electrolyte ions

  2. Solid Waste Activity Packet for Teachers.

    Science.gov (United States)

    Illinois Univ., Urbana. Cooperative Extension Service.

    This solid waste activity packet introduces students to the solid waste problem in Illinois. Topics explore consumer practices in the market place, packaging, individual and community garbage generation, and disposal practices. The activities provide an integrated approach to incorporating solid waste management issues into subject areas. The…

  3. Long-term high-level waste technology program

    International Nuclear Information System (INIS)

    1980-04-01

    The Department of Energy (DOE) is conducting a comprehensive program to isolate all US nuclear wastes from the human environment. The DOE Office of Nuclear Energy - Waste (NEW) has full responsibility for managing the high-level wastes resulting from defense activities and additional responsiblity for providing the technology to manage existing commercial high-level wastes and any that may be generated in one of several alternative fuel cycles. Responsibilities of the Three Divisions of DOE-NEW are shown. This strategy document presents the research and development plan of the Division of Waste Products for long-term immobilization of the high-level radioactive wastes resulting from chemical processing of nuclear reactor fuels and targets. These high-level wastes contain more than 99% of the residual radionuclides produced in the fuels and targets during reactor operations. They include essentially all the fission products and most of the actinides that were not recovered for use

  4. Supplemental Performance Analyses for Igneous Activity and Human Intrusion at the Potential High-Level Nuclear Waste Repository at Yucca Mountain

    International Nuclear Information System (INIS)

    Swift, P.; Gaither, K.; Freeze, G.; McCord, J.; Kalinich, D.; Saulnier, G.; Statham, W.

    2002-01-01

    The U.S. Department of Energy (DOE) is considering the possible recommendation of a site at Yucca Mountain, Nevada, for the potential development of a geologic repository for the disposal of high-level radioactive waste and spent nuclear fuel. Consequences of hypothetical disruption of the Yucca Mountain site by igneous activity or human intrusion have been evaluated in the Yucca Mountain Science and Engineering Report (S and ER) (1), which presents technical information supporting the consideration of the possible site recommendation. Since completion of the S and ER, supplemental analyses have examined possible impacts of new information and alternative assumptions on the estimates of the consequences of these events. Specifically, analyses of the consequences of igneous disruption address uncertainty regarding: (1) the impacts of changes in the repository footprint and waste package spacing on the probability of disruption; (2) impacts of alternative assumptions about the appropriate distribution of future wind speeds to use in the analysis; (3) effects of alternative assumptions about waste particle sizes; and (4) alternative assumptions about the number of waste packages damaged by igneous intrusion; and (5) alternative assumptions about the exposure pathways and the biosphere dose conversion factors used in the analysis. Additional supplemental analyses, supporting the Final Environmental Impact Statement (FEIS), have examined the results for both igneous disruption and human intrusion, recalculated for a receptor group located 18 kilometers (km) from the repository (the location specified in 40 CFR 197), rather than at the 20 km distance used in the S and ER analyses

  5. Alternatives evaluation of high activity radioactive wastes disposal; Evaluacion de alternativas de eliminacion de residuos radiactivos de alta actividad

    Energy Technology Data Exchange (ETDEWEB)

    Ciallella, N R; Petraitis, E J

    1990-12-31

    Different alternatives considered in the world to be used as barriers to isolate the high level radioactive from the environment wastes produced during the electric energy generation of nuclear origin are presented. Engineering and geologic barriers, are analyzed, considering nuclear fuel cycles with or without plutonium recycling; to that purpose the consideration of elements such as durability and resistance of the various engineering, availability of the fabrication processes, associated radiological impact, geological media apt to be used as geological barrier. Finally, the scopes of the Feasibility Study and Engineering draft are presented for the construction of a repository for high-level radioactive wastes, for the Argentine Nuclear Program needs, which contemplates the construction of six nuclear power plants with a potential installed towards the year 2000 GW({sub e}), with natural and/or lowly enriched uranium power plants and recycling of plutonium generated in the cycle. (Author). [Espanol] Se presentan las distintas alternativas consideradas en el mundo para ser utilizadas como barreras para aislar del ambiente los residuos radiactivos de alta actividad producidos durante la generacion de energia electrica de origen nuclear. Se analizan barreras de ingenieria y geologicas considerando tanto ciclos de combustible nuclear con y sin reciclado de plutonio, realizandose a tal fin la consideracion de elementos tales como durabilidad y resistencia de las distintas barreras de ingenieria, disponibilidad de los procesos de fabricacion, impacto radiologico asociado, medios geologicos aptos para ser utilizados como barrera geologica. Por ultimo, se presentan los alcances del Estudio de Factibilidad y Anteproyecto de Ingenieria para la construccion de un Repositorio para Residuos Radiactivos de Alta Actividad, para las necesidades del Plan Nuclear Argentino, que contempla la construccion de seis centrales nucleares, con una potencia instalada hacia el ano 2000 GW

  6. Current high-level waste solidification technology

    International Nuclear Information System (INIS)

    Bonner, W.F.; Ross, W.A.

    1976-01-01

    Technology has been developed in the U.S. and abroad for solidification of high-level waste from nuclear power production. Several processes have been demonstrated with actual radioactive waste and are now being prepared for use in the commercial nuclear industry. Conversion of the waste to a glass form is favored because of its high degree of nondispersibility and safety

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

    International Nuclear Information System (INIS)

    Ramsey, William Gene

    2013-01-01

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

  8. High-level waste processing and disposal

    International Nuclear Information System (INIS)

    Crandall, J.L.; Krause, H.; Sombret, C.; Uematsu, K.

    1984-11-01

    Without reprocessing, spent LWR fuel itself is generally considered an acceptable waste form. With reprocessing, borosilicate glass canisters, have now gained general acceptance for waste immobilization. The current first choice for disposal is emplacement in an engineered structure in a mined cavern at a depth of 500-1000 meters. A variety of rock types are being investigated including basalt, clay, granite, salt, shale, and volcanic tuff. This paper gives specific coverage to the national high level waste disposal plans for France, the Federal Republic of Germany, Japan and the United States. The French nuclear program assumes prompt reprocessing of its spent fuels, and France has already constructed the AVM. Two larger borosilicate glass plants are planned for a new French reprocessing plant at La Hague. France plans to hold the glass canisters in near-surface storage for a forty to sixty year cooling period and then to place them into a mined repository. The FRG and Japan also plan reprocessing for their LWR fuels. Both are currently having some fuel reprocessed by France, but both are also planning reprocessing plants which will include waste vitrification facilities. West Germany is now constructing the PAMELA Plant at Mol, Belgium to vitrify high level reprocessing wastes at the shutdown Eurochemic Plant. Japan is now operating a vitrification mockup test facility and plans a pilot plant facility at the Tokai reprocessing plant by 1990. Both countries have active geologic repository programs. The United State program assumes little LWR fuel reprocessing and is thus primarily aimed at direct disposal of spent fuel into mined repositories. However, the US have two borosilicate glass plants under construction to vitrify existing reprocessing wastes

  9. Studies on the thermal decomposition of nitrates found in highly active waste and of chemicals used to convert the waste to glass

    International Nuclear Information System (INIS)

    Chun, K.S.

    1977-05-01

    The decomposition of all the individual chemicals used in the Harwell inactive vitrification pilot plant has been studied by means of a thermal balance. Weight loss curves to 1100 0 C have been obtained. The four materials sodium nitrate, cesium nitrate, lithium nitrate and ruthenium nitroso-nitrate (solution) showed a greater weight loss than that based on an oxide yield, and hence these compounds or their products of decomposition are volatile below 1100 0 C. The remaining materials suffered a weight loss no more than that corresponding to a full yield of the oxide, and hence they were not volatile below 1100 0 C. Most of the chemicals begin to decompose at less than 75 0 C but the nitrates of cesium, strontium, barium and sodium not until 295 0 to 590 0 C. The results obtained can be used in the analysis of process conditions in the vitrification and calcination of highly radioactive wastes and also of the thermal decomposition behaviour of mixtures containing those materials. The materials tested were: Al(NO 3 ) 3 .9H 2 O, Ba(NO 3 ) 2 , CaNo 3 , Cr(NO 3 ) 3 .9H 2 O, Fe(NO 3 ) 3 .9H 2 O. Mg(NO 3 ) 2 .6H 2 O, Ni(NO 3 ) 2 .6H 2 O, R.E. Nitrates, Ruthenium Solution, Sr(NO 3 ) 2 , UO 2 (NO 3 ) 2 .6H 2 O, Zn(NO 3 ) 2 .6H 2 O. Zirconium Solution, 'Gasil WP' Silica, 'Neosyl' Silica, LiOH.H 2 O. LiNO 3 .3H 2 O, Na 2 CO 3 , NaNO 3 , Na 2 B 4 O 7 .10H 2 O. (author)

  10. Overview: Defense high-level waste technology program

    International Nuclear Information System (INIS)

    Shupe, M.W.; Turner, D.A.

    1987-01-01

    Defense high-level waste generated by atomic energy defense activities is stored on an interim basis at three U.S. Department of Energy (DOE) operating locations; the Savannah River Plant in South Carolina, the Hanford Site in Washington, and the Idaho National Engineering Laboratory in Idaho. Responsibility for the permanent disposal of this waste resides with DOE's Office of Defense Waste and Transportation Management. The objective of the Defense High-Level Wast Technology Program is to develop the technology for ending interim storage and achieving permanent disposal of all U.S. defense high-level waste. New and readily retrievable high-level waste are immobilized for disposal in a geologic repository. Other high-level waste will be stabilized in-place if, after completion of the National Environmental Policy Act (NEPA) process, it is determined, on a site-specific basis, that this option is safe, cost effective and environmentally sound. The immediate program focus is on implementing the waste disposal strategy selected in compliance with the NEPA process at Savannah River, while continuing progress toward development of final waste disposal strategies at Hanford and Idaho. This paper presents an overview of the technology development program which supports these waste management activities and an assessment of the impact that recent and anticipated legal and institutional developments are expected to have on the program

  11. Pump Jet Mixing and Pipeline Transfer Assessment for High-Activity Radioactive Wastes in Hanford Tank 241-AZ-102

    Energy Technology Data Exchange (ETDEWEB)

    Y Onishi; KP Recknagle; BE Wells

    2000-08-09

    The authors evaluated how well two 300-hp mixer pumps would mix solid and liquid radioactive wastes stored in Hanford double-shell Tank 241-AZ-102 (AZ-102) and confirmed the adequacy of a three-inch (7.6-cm) pipeline system to transfer the resulting mixed waste slurry to the AP Tank Farm and a planned waste treatment (vitrification) plant on the Hanford Site. Tank AZ-102 contains 854,000 gallons (3,230 m{sup 3}) of supernatant liquid and 95,000 gallons (360 m{sup 3}) of sludge made up of aging waste (or neutralized current acid waste). The study comprises three assessments: waste chemistry, pump jet mixing, and pipeline transfer. The waste chemical modeling assessment indicates that the sludge, consisting of the solids and interstitial solution, and the supernatant liquid are basically in an equilibrium condition. Thus, pump jet mixing would not cause much solids precipitation and dissolution, only 1.5% or less of the total AZ-102 sludge. The pump jet mixing modeling indicates that two 300-hp mixer pumps would mobilize up to about 23 ft (7.0 m) of the sludge nearest the pump but would not erode the waste within seven inches (0.18 m) of the tank bottom. This results in about half of the sludge being uniformly mixed in the tank and the other half being unmixed (not eroded) at the tank bottom.

  12. Pump Jet Mixing and Pipeline Transfer Assessment for High-Activity Radioactive Wastes in Hanford Tank 241-AZ-102

    International Nuclear Information System (INIS)

    Onishi, Y.; Recknagle, K.P.; Wells, B.E.

    2000-01-01

    The authors evaluated how well two 300-hp mixer pumps would mix solid and liquid radioactive wastes stored in Hanford double-shell Tank 241-AZ-102 (AZ-102) and confirmed the adequacy of a three-inch (7.6-cm) pipeline system to transfer the resulting mixed waste slurry to the AP Tank Farm and a planned waste treatment (vitrification) plant on the Hanford Site. Tank AZ-102 contains 854,000 gallons (3,230 m 3 ) of supernatant liquid and 95,000 gallons (360 m 3 ) of sludge made up of aging waste (or neutralized current acid waste). The study comprises three assessments: waste chemistry, pump jet mixing, and pipeline transfer. The waste chemical modeling assessment indicates that the sludge, consisting of the solids and interstitial solution, and the supernatant liquid are basically in an equilibrium condition. Thus, pump jet mixing would not cause much solids precipitation and dissolution, only 1.5% or less of the total AZ-102 sludge. The pump jet mixing modeling indicates that two 300-hp mixer pumps would mobilize up to about 23 ft (7.0 m) of the sludge nearest the pump but would not erode the waste within seven inches (0.18 m) of the tank bottom. This results in about half of the sludge being uniformly mixed in the tank and the other half being unmixed (not eroded) at the tank bottom

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

    International Nuclear Information System (INIS)

    Deffenbaugh, M.L.

    1997-01-01

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

  14. High level waste at Hanford: Potential for waste loading maximization

    International Nuclear Information System (INIS)

    Hrma, P.R.; Bailey, A.W.

    1995-09-01

    The loading of Hanford nuclear waste in borosilicate glass is limited by phase-related phenomena, such as crystallization or formation of immiscible liquids, and by breakdown of the glass structure because of an excessive concentration of modifiers. The phase-related phenomena cause both processing and product quality problems. The deterioration of product durability determines the ultimate waste loading limit if all processing problems are resolved. Concrete examples and mass-balance based calculations show that a substantial potential exists for increasing waste loading of high-level wastes that contain a large fraction of refractory components

  15. Cermets for high level waste containment

    International Nuclear Information System (INIS)

    Aaron, W.S.; Quinby, T.C.; Kobisk, E.H.

    1978-01-01

    Cermet materials are currently under investigation as an alternate for the primary containment of high level wastes. The cermet in this study is an iron--nickel base metal matrix containing uniformly dispersed, micron-size fission product oxides, aluminosilicates, and titanates. Cermets possess high thermal conductivity, and typical waste loading of 70 wt % with volume reduction factors of 2 to 200 and low processing volatility losses have been realized. Preliminary leach studies indicate a leach resistance comparable to other candidate waste forms; however, more quantitative data are required. Actual waste studies have begun on NFS Acid Thorex, SRP dried sludge and fresh, unneutralized SRP process wastes

  16. Radioactive waste characterisation by neutron activation

    International Nuclear Information System (INIS)

    Nicol, Tangi

    2016-01-01

    Nuclear activities produce radioactive wastes classified following their radioactive level and decay time. an accurate characterization is necessary for efficient classification and management. Medium and high level wastes containing long lived radioactive isotopes will be stored in deep geological storage for hundreds of thousands years. at the end of this period, it is essential to ensure that the wastes do not represent any risk for humans and environment, not only from radioactive point of view, but also from stable toxic chemicals. This PhD thesis concerns the characterization of toxic chemicals and nuclear material in radioactive waste, by using neutron activation analysis, in the frame of collaboration between the Nuclear Measurement Laboratory of CEA Cadarache, France, and the Institute of Nuclear Waste Management and Reactor Safety of the research center, FZJ (Forschungszentrum Juelich GmbH), Germany. The first study is about the validation of the numerical model of the neutron activation cell MEDINA (FZJ), using MCNP Monte Carlo transport code. Simulations and measurements of prompt capture gamma rays from small samples measured in MEDINA have been compared for a number of elements of interest (beryllium, aluminum, chlorine, copper, selenium, strontium, and tantalum). The comparison was performed using different nuclear databases, resulting in satisfactory agreement and validating simulation in view of following studies. Then, the feasibility of fission delayed gamma-ray measurements of "2"3"9Pu and "2"3"5U in 225 L waste drums has been studied, considering bituminized or concrete matrixes representative of wastes produced in France and Germany. The delayed gamma emission yields were first determined from uranium and plutonium metallic samples measurements in REGAIN, the neutron activation cell of LMN, showing satisfactory consistency with published data. The useful delayed gamma signals of "2"3"9Pu and "2"3"5U, homogeneously distributed in the 225 L

  17. High Fat Diet-Induced Skeletal Muscle Wasting Is Decreased by Mesenchymal Stem Cells Administration: Implications on Oxidative Stress, Ubiquitin Proteasome Pathway Activation, and Myonuclear Apoptosis

    Directory of Open Access Journals (Sweden)

    Johanna Abrigo

    2016-01-01

    Full Text Available Obesity can lead to skeletal muscle atrophy, a pathological condition characterized by the loss of strength and muscle mass. A feature of muscle atrophy is a decrease of myofibrillar proteins as a result of ubiquitin proteasome pathway overactivation, as evidenced by increased expression of the muscle-specific ubiquitin ligases atrogin-1 and MuRF-1. Additionally, other mechanisms are related to muscle wasting, including oxidative stress, myonuclear apoptosis, and autophagy. Stem cells are an emerging therapy in the treatment of chronic diseases such as high fat diet-induced obesity. Mesenchymal stem cells (MSCs are a population of self-renewable and undifferentiated cells present in the bone marrow and other mesenchymal tissues of adult individuals. The present study is the first to analyze the effects of systemic MSC administration on high fat diet-induced skeletal muscle atrophy in the tibialis anterior of mice. Treatment with MSCs reduced losses of muscle strength and mass, decreases of fiber diameter and myosin heavy chain protein levels, and fiber type transitions. Underlying these antiatrophic effects, MSC administration also decreased ubiquitin proteasome pathway activation, oxidative stress, and myonuclear apoptosis. These results are the first to indicate that systemically administered MSCs could prevent muscle wasting associated with high fat diet-induced obesity and diabetes.

  18. Effects of loading rate and temperature on anaerobic co-digestion of food waste and waste activated sludge in a high frequency feeding system, looking in particular at stability and efficiency.

    Science.gov (United States)

    Li, Qian; Li, Hao; Wang, Gaojun; Wang, Xiaochang

    2017-08-01

    A continuously stirred tank reactor (CSTR) with a high feeding frequency (HFF) of once every 15min was employed in order to ease the loading shock frequently occurred in digester with a low feeding frequency. The effects of the organic loading rate (OLR) and temperature on the co-digestion of food waste and waste activated sludge was evaluated in a 302-day long-term experiment. Due to the high hydrolysis rate, the maximum CH 4 yield in a thermophilic reactor was 407mL CH 4 /gVS added , a value that was significantly higher than the 350mL CH 4 /gVS added that occurred in a mesophilic reactor. Although the alkalinity declined when HRT was shorted than 10d, caused by the decrease of conversion ratio from protein to ammonium, the increase of specific methanogenic activity helped HFF system to achieve stable performance at an OLR of 11.2 (HRT 7.5d) and 30.2gVS/L/d (HRT 3d) under mesophilic and thermophilic conditions, respectively. Copyright © 2017 Elsevier Ltd. All rights reserved.

  19. High integrity container evaluation for solid waste disposal burial containers

    International Nuclear Information System (INIS)

    Josephson, W.S.

    1996-01-01

    In order to provide radioactive waste disposal practices with the greatest measure of public protection, Solid Waste Disposal (SWD) adopted the Nuclear Regulatory Commission (NRC) requirement to stabilize high specific activity radioactive waste prior to disposal. Under NRC guidelines, stability may be provided by several mechanisms, one of which is by placing the waste in a high integrity container (HIC). During the implementation process, SWD found that commercially-available HICs could not accommodate the varied nature of weapons complex waste, and in response developed a number of disposal containers to function as HICs. This document summarizes the evaluation of various containers that can be used for the disposal of Category 3 waste in the Low Level Burial Grounds. These containers include the VECTRA reinforced concrete HIC, reinforced concrete culvert, and the reinforced concrete vault. This evaluation provides justification for the use of these containers and identifies the conditions for use of each

  20. Recent activity on disposal of uranium waste

    International Nuclear Information System (INIS)

    Fujiwara, Noboru

    1999-01-01

    The concept on the disposal of uranium waste has not been discussed in the Atomic Energy Commission of Japan, but the research and development of it are carried out in the company and agency which are related to uranium waste. In this paper, the present condition and problems on disposal of uranium waste were shown in aspect of the nuclear fuel manufacturing companies' activity. As main contents, the past circumstances on the disposal of uranium waste, the past activity of nuclear fuel manufacturing companies, outline and properties of uranium waste were shown, and ideas of nuclear fuel manufacturing companies on the disposal of uranium waste were reported with disposal idea in the long-term program for development and utilization of nuclear energy. (author)

  1. Hanford Immobilized Low-Activity Waste Product Acceptance Test Plan

    International Nuclear Information System (INIS)

    Peeler, D.

    1999-01-01

    'The Hanford Site has been used to produce nuclear materials for the U.S. Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during Pu production, exists in 177 underground single- and double-shell tanks. These wastes are to be retrieved and separated into low-activity waste (LAW) and high-level waste (HLW) fractions. The DOE is proceeding with an approach to privatize the treatment and immobilization of Handord''s LAW and HLW.'

  2. Hanford Immobilized Low-Activity Waste Product Acceptance Test Plan

    Energy Technology Data Exchange (ETDEWEB)

    Peeler, D.

    1999-06-22

    'The Hanford Site has been used to produce nuclear materials for the U.S. Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during Pu production, exists in 177 underground single- and double-shell tanks. These wastes are to be retrieved and separated into low-activity waste (LAW) and high-level waste (HLW) fractions. The DOE is proceeding with an approach to privatize the treatment and immobilization of Handord''s LAW and HLW.'

  3. The IAEA's high level radioactive waste management programme

    International Nuclear Information System (INIS)

    Saire, D.E.

    1994-01-01

    This paper presents the different activities that are performed under the International Atomic Energy Agency's (IAEA) high level radioactive waste management programme. The Agency's programme is composed of five main activities (information exchange, international safety standards, R ampersand D activities, advisory services and special projects) which are described in the paper. Special emphasis is placed on the RADioactive WAste Safety Standards (RADWASS) programme which was implemented in 1991 to document international consensus that exists on the safe management of radioactive waste. The paper also raises the question about the need for regional repositories to serve certain countries that do not have the resources or infrastructure to construct a national repository

  4. High-level radioactive waste management

    International Nuclear Information System (INIS)

    Schneider, K.J.; Liikala, R.C.

    1974-01-01

    High-level radioactive waste in the U.S. will be converted to an encapsulated solid and shipped to a Federal repository for retrievable storage for extended periods. Meanwhile the development of concepts for ultimate disposal of the waste which the Federal Government would manage is being actively pursued. A number of promising concepts have been proposed, for which there is high confidence that one or more will be suitable for long-term, ultimate disposal. Initial evaluations of technical (or theoretical) feasibility for the various waste disposal concepts show that in the broad category, (i.e., geologic, seabed, ice sheet, extraterrestrial, and transmutation) all meet the criteria for judging feasibility, though a few alternatives within these categories do not. Preliminary cost estimates show that, although many millions of dollars may be required, the cost for even the most exotic concepts is small relative to the total cost of electric power generation. For example, the cost estimates for terrestrial disposal concepts are less than 1 percent of the total generating costs. The cost for actinide transmutation is estimated at around 1 percent of generation costs, while actinide element disposal in space is less than 5 percent of generating costs. Thus neither technical feasibility nor cost seems to be a no-go factor in selecting a waste management system. The seabed, ice sheet, and space disposal concepts face international policy constraints. The information being developed currently in safety, environmental concern, and public response will be important factors in determining which concepts appear most promising for further development

  5. Nuclear waste isolation activities report

    International Nuclear Information System (INIS)

    1980-12-01

    Included are: a report from the Deputy Assistant Secretary, a summary of recent events, new literature, a list of upcoming waste management meetings, and background information on DOE's radwaste management programs

  6. Desactivation of liquid radioactive wastes of low and medium activity

    International Nuclear Information System (INIS)

    Golinski, M.; Charomska, K.

    1978-01-01

    The results of research made according to the prodranm of scientific and technical cooperation of the CMEA countries are discussed. The main direction of these research works is on future improvement of installations for purification of liquid radioactive wastes by chemical methods of coprecipitation and coagulation, ion exchange, sorption, distillation and electrolysis. It was shown that methods of coprecipitation and coagulation have low efficiency and the activity reduction factor seldom was more than 10. In sorption processes different sorbents, both organic and nonorganic were used. The modified bentonite used as a sorbent agent has shown high selectivity towards zesium ions. Waste concentration by means of distillation is an universal but rather expensive method and is applied mainly in the cases of high salts concentration and high specific activity of liquid wastes. Electrolysis, as a method of the liquid wastes purification is used in the USSR and has high efficiency with low energy consumption. (I.T.) [ru

  7. 1. round table - Nuclear wastes and radioactive materials. 2. round table - risks linked with nuclear wastes and materials. 3. round table - the problem of long-term management of medium-high activity and long lived wastes. The process defined by the 1991 law

    International Nuclear Information System (INIS)

    2005-01-01

    The law from December 30, 1991, precisely defines 3 axes of researches for the management of high level and long-lived radioactive wastes: separation/transmutation, surface storage and underground disposal. A global evaluation report about these researches is to be supplied in 2006 by the French government to the Parliament. A first synthesis of the knowledge gained after 14 years of research has led the national commission of the public debate (CNDP) to organize a national debate about the general options of management of high-level and long-lived radioactive wastes before the 2006 date line. The debate comprises 4 public hearings (September 2005: Bar-le-Duc, Saint-Dizier, Pont-du-Gard, Cherbourg), 12 round-tables (October and November 2005: Paris, Joinville, Caen, Nancy, Marseille), a synthesis meeting (December 2005, Dunkerque) and a closing meeting (January 2006, Lyon). This document is the synthesis of the debates of the first round table of Paris about the problems raised by nuclear wastes in the case of the geologic disposal option. Four families of questions have been tackled: 1 - the exhaustiveness of ANDRA's inventory, the solutions foreseen for the different types of wastes; 2 - the high-medium activity wastes and their processing; 3 - the management of non-reprocessed spent MOX fuels; 4 - the safety and security standards used and their establishment. Four presentations are attached to these proceedings and deal with: the measured and estimated inventory of all radioactive wastes; the inventory and management of radioactive wastes and the place of citizens; the point of view of the nuclear safety authority; conditioning and storage. (J.S.)

  8. Researches on the management of high activity and long-lived radioactive wastes. Axis 1 - separation-transmutation; Recherches sur la gestion des dechets radioactifs a haute activite et a vie longue. Axe 1 - separation-transmutation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-11-15

    This document gathers the transparencies of seven presentations given at a technical workshop of the French nuclear energy society (SFEN) about the researches on separation-transmutation of high activity and long-lived radioactive wastes. The presentations deal with: inventory and radiotoxicity of the rad-wastes in concern; industrial experience; experience on chemical separation: molecules and processes; reactors physics and transmutation - reactors for transmutation; fuels and targets; scenarios that include transmutation; environmental impacts of these different scenarios. (J.S.)

  9. Bacteriological studies on dairy waste activated sludge

    NARCIS (Netherlands)

    Adamse, A.D.

    1966-01-01

    Dairy-waste activated sludge was examined for bacterial composition and response to different conditions. Strains isolated were classified mainly into three groups: predominantly coryneform bacteria (largely Arthrobacter), some Achromobacteraceae and a small groups of Pseudomonadaceae.

  10. PAIRWISE BLENDING OF HIGH LEVEL WASTE

    International Nuclear Information System (INIS)

    CERTA, P.J.

    2006-01-01

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

  11. Disposal of high-level radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Costello, J M [Australian Atomic Energy Commission Research Establishment, Lucas Heights

    1982-03-01

    The aims and options for the management and disposal of highly radioactive wastes contained in spent fuel from the generation of nuclear power are outlined. The status of developments in reprocessing, waste solidification and geologic burial in major countries is reviewed. Some generic assessments of the potential radiological impacts from geologic repositories are discussed, and a perspective is suggested on risks from radiation.

  12. Characterization plan for the immobilized low-activity waste borehole

    International Nuclear Information System (INIS)

    Reidel, S.P.; Reynolds, K.D.

    1998-03-01

    The US Department of Energy's (DOE's) Hanford Site has the most diverse and largest amounts of radioactive tank waste in the US. High-level radioactive waste has been stored at Hanford in large underground tanks since 1944. Approximately 209,000 m 3 (54 Mgal) of waste are currently stored in 177 tanks. Vitrification and onsite disposal of low activity tank waste (LAW) are embodied in the strategy described in the Tri-Party Agreement. The tank waste is to be retrieved, separated into low- and high-level fractions, and then immobilized by private vendors. The DOE will receive the vitrified waste from private vendors and dispose of the low-activity fraction in the Hanford Site 200 East Area. The Immobilized Low-Activity Waste Disposal Complex (ILAWDC) is part of the disposal complex. This report is a plan to drill the first characterization borehole and collect data at the ILAWDC. This plan updates and revises the deep borehole portion of the characterization plan for the ILAWDC by Reidel and others (1995). It describes data collection activities for determining the physical and chemical properties of the vadose zone and the saturated zone at and in the immediate vicinity of the proposed ILAWDC. These properties then will be used to develop a conceptual geohydrologic model of the ILAWDC site in support of the Hanford ILAW Performance Assessment

  13. Recovery of Am-Cm from high-activity waste concentrate by in-canyon-tank precipitation as oxalates

    International Nuclear Information System (INIS)

    Gray, L.W.; Burney, G.A.; Wilson, T.W.; McKibben, J.M.

    1980-01-01

    Savannah River Laboratory and Savannah River Plant have been separating actinides for more than 25 years. Work continues to upgrade processes and to initiate new processes. This report summarizes work on a precipitation process to separate kg amounts of Am and Cm from hundreds of kilograms of NaNO 3 and Al(NO 3 ) 3 . The developed process includes formic acid denitration of the Am-Cm bearing streams for acid adjustment; oxalate precipitation of the Am-Cm; and Mn +2 catalyzed oxidation of oxalate in both the decanted supernate and the precipitated actinides. The new process generates one fourth the radioactive waste as the solvent extraction process which it replaced, and produces a cleaner feed solution for downstream processing to separate the Am and Cm before conversion to their respective oxides

  14. The development of a high level radioactive waste management strategy

    International Nuclear Information System (INIS)

    Beale, H.

    1979-11-01

    The management of high level radioactive waste, from the removal of spent fuel from reactors to final disposal of vitrified waste, involves a complex choice of operational variables which interact one with another. If the various operations are designed and developed in isolation it will almost certainly lead to suboptimal choice. Management of highly active waste should therefore be viewed as a complete system and analysed in such a way that account is taken of the interactions between the various operations. This system must have clearly defined and agreed objectives as well as criteria against which performance can be judged. A thorough analysis of the system will provide a framework within which the necessary research and development can be carried out in a co-ordinated fashion and lead to an optimised strategy for managing highly active wastes. (author)

  15. Fission products determination in high activity waste solution by wavelength dispersive X-ray fluorescence spectral interference correction by intensity ratio

    International Nuclear Information System (INIS)

    Sato, I.M.

    1988-01-01

    Fission products Se, Rb, Y, Zr, Mo, Ru, Rh, Pd, Te, Cd, Cs, Ba, La, Ce, Pr, Nd, Sm, Eu and Gd were determined in simulated high activity radioactive waste solution by wavelength dispersive X-ray fluorescence spectrometry without chemical separation. Thin layer technique was employed for the sample preparation. For the L spectral lines, the absorption effect was verified by Rasberry-Heinrich, Lucas Tooth-Pyne and Lachance-Trail relations. This effect was quantified and corrected accordingly. The spectral interferences of Kα and/or Lα lines of Y, Zr, Mo, La, Ce, Pr, Nd, Sm, Eu and Gd elements were eliminated by the intensity ratio method. The overlapping of up to three analytical lines was corrected by applying this method. The concentration influence of the interfering element on the intensity ratio values as well the efficiency of the correction method were investigated in order to assure that no systematic or residual error, resulting from the correction, affect the actual fluorescent intensity determination. The results is compared with the data obtained from measurements of free lines of spectral interference and also with those obtained by the linear equation system. Fission products determination presented a precision in the range of 0.1 to 5.0% and an accuracy of up to ± 7.0% the results are compared with those obtained by neutron activation analysis and inductively coupled plasma - atomic emission spectrometry. Leaching data, when radioactive waste is incorporated in cement matrix, were attempted by X-ray fluorescence technique. For two years leaching period, leaching rate and diffusion coefficient data of cesium were determined. The results obtained agree with those obtained by γ-spectromety. (author) [pt

  16. Performance Enhancements to the Hanford Waste Treatment and Immobilization Plant Low-Activity Waste Vitrification System

    International Nuclear Information System (INIS)

    Hamel, W. F.; Gerdes, K.; Holton, L. K.; Pegg, I.L.; Bowan, B.W.

    2006-01-01

    The U.S Department of Energy Office of River Protection (DOE-ORP) is constructing a Waste Treatment and Immobilization Plant (WTP) for the treatment and vitrification of underground tank wastes stored at the Hanford Site in Washington State. The WTP comprises four major facilities: a pretreatment facility to separate the tank waste into high level waste (HLW) and low-activity waste (LAW) process streams, a HLW vitrification facility to immobilize the HLW fraction; a LAW vitrification facility to immobilize the LAW fraction, and an analytical laboratory to support the operations of all four treatment facilities. DOE has established strategic objectives to optimize the performance of the WTP facilities and the LAW and HLW waste forms to reduce the overall schedule and cost for treatment and vitrification of the Hanford tank wastes. This strategy has been implemented by establishing performance expectations in the WTP contract for the facilities and waste forms. In addition, DOE, as owner-operator of the WTP facilities, continues to evaluate 1) the design, to determine the potential for performance above the requirements specified in the WTP contract; and 2) improvements in production of the LAW and HLW waste forms. This paper reports recent progress directed at improving production of the LAW waste form. DOE's initial assessment, which is based on the work reported in this paper, is that the treatment rate of the WTP LAW vitrification facility can be increased by a factor of 2 to 4 with a combination of revised glass formulations, modest increases in melter glass operating temperatures, and a second-generation LAW melter with a larger surface area. Implementing these improvements in the LAW waste immobilization capability can benefit the LAW treatment mission by reducing the cost of waste treatment. (authors)

  17. High Level Waste (HLW) Processing Experience with Increased Waste Loading

    International Nuclear Information System (INIS)

    JANTZEN, CAROL

    2004-01-01

    The Defense Waste Processing Facility (DWPF) Engineering requested characterization of glass samples that were taken after the second melter had been operational for about 5 months. After the new melter had been installed, the waste loading had been increased to about 38 weight percentage after a new quasicrystalline liquidus model had been implemented. The DWPF had also switched from processing with refractory Frit 200 to a more fluid Frit 320. The samples were taken after DWPF observed very rapid buildup of deposits in the upper pour spout bore and on the pour spout insert while processing the high waste loading feedstock. These samples were evaluated using various analytical techniques to determine the cause of the crystallization. The pour stream sample was homogeneous, amorphous, and representative of the feed batch from which it was derived. Chemical analysis of the pour stream sample indicated that a waste loading of 38.5 weight per cent had been achieved. The data analysis indicated that surface crystallization, induced by temperature and oxygen fugacity gradients in the pour spout, caused surface crystallization to occur in the spout and on the insert at the higher waste loadings even though there was no crystallization in the pour stream

  18. Engineering properties for high kitchen waste content municipal solid waste

    Directory of Open Access Journals (Sweden)

    Wu Gao

    2015-12-01

    Full Text Available Engineering properties of municipal solid waste (MSW depend largely on the waste's initial composition and degree of degradation. MSWs in developing countries usually have a high kitchen waste content (called HKWC MSW. After comparing and analyzing the laboratory and field test results of physical composition, hydraulic properties, gas generation and gas permeability, and mechanical properties for HKWC MSW and low kitchen waste content MSW (called LKWC MSW, the following findings were obtained: (1 HKWC MSW has a higher initial water content (IWC than LKWC MSW, but the field capacities of decomposed HKWC and LKWC MSWs are similar; (2 the hydraulic conductivity and gas permeability for HKWC MSW are both an order of magnitude smaller than those for LKWC MSW; (3 compared with LKWC MSW, HKWC MSW has a higher landfill gas (LFG generation rate but a shorter duration and a lower potential capacity; (4 the primary compression feature for decomposed HKWC MSW is similar to that of decomposed LKWC MSW, but the compression induced by degradation of HKWC MSW is greater than that of LKWC MSW; and (5 the shear strength of HKWC MSW changes significantly with time and strain. Based on the differences of engineering properties between these two kinds of MSWs, the geo-environmental issues in HKWC MSW landfills were analyzed, including high leachate production, high leachate mounds, low LFG collection efficiency, large settlement and slope stability problem, and corresponding advice for the management and design of HKWC MSW landfills was recommended.

  19. Heat transfer in high-level waste management

    International Nuclear Information System (INIS)

    Dickey, B.R.; Hogg, G.W.

    1979-01-01

    Heat transfer in the storage of high-level liquid wastes, calcining of radioactive wastes, and storage of solidified wastes are discussed. Processing and storage experience at the Idaho Chemical Processing Plant are summarized for defense high-level wastes; heat transfer in power reactor high-level waste processing and storage is also discussed

  20. High temperature slagging incineration of hazardous waste

    International Nuclear Information System (INIS)

    Vanbrabant, R.; Van de Voorde, N.

    1987-01-01

    The SCK/CEN, as the treatment center for the low level radioactive waste in Belgium, develops appropriate treatment systems for different kinds of wastes. The technical concept of the high temperature slagging incineration system has been developed and improved. The construction of the first demonstration plant was initiated in 1974. Since then the system has been operated regularly and further developed with the view to industrial operations. Now it handles about 5 tons of waste in a week. The waste which is treated consists of low level beta/gamma and alpha-contaminated radioactive waste. Because of the special characteristics the system is thought to be an excellent incineration system for industrial hazardous waste as well. Recently the SCK/CEN has received the authorization to treat industrial hazardous waste in the same installation. Preliminary tests have been executed on special waste products, such as PCB-contaminated liquids, with excellent incineration results. Incineration efficiency up to 99.9999% could be obtained. The paper presents the state of the art of this original The SCK/CEN-technology and gives the results of the tests done with special hazard

  1. Waste Treatment Technology Process Development Plan For Hanford Waste Treatment Plant Low Activity Waste Recycle

    International Nuclear Information System (INIS)

    McCabe, Daniel J.; Wilmarth, William R.; Nash, Charles A.

    2013-01-01

    The purpose of this Process Development Plan is to summarize the objectives and plans for the technology development activities for an alternative path for disposition of the recycle stream that will be generated in the Hanford Waste Treatment Plant Low Activity Waste (LAW) vitrification facility (LAW Recycle). This plan covers the first phase of the development activities. The baseline plan for disposition of this stream is to recycle it to the WTP Pretreatment Facility, where it will be concentrated by evaporation and returned to the LAW vitrification facility. Because this stream contains components that are volatile at melter temperatures and are also problematic for the glass waste form, they accumulate in the Recycle stream, exacerbating their impact on the number of LAW glass containers. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and reducing the halides in the Recycle is a key component of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, this stream does not have a proven disposition path, and resolving this gap becomes vitally important. This task seeks to examine the impact of potential future disposition of this stream in the Hanford tank farms, and to develop a process that will remove radionuclides from this stream and allow its diversion to another disposition path, greatly decreasing the LAW vitrification mission duration and quantity of glass waste. The origin of this LAW Recycle stream will be from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover or precipitates of scrubbed components (e.g. carbonates). The soluble

  2. Study of the behavior of high activity waste produced during the regeneration of fast reactor fuel elements by the gaseous fluoride method

    International Nuclear Information System (INIS)

    Kiriolovich, A.P.; Dem'yanovich, M.A.; Lavrinovich, Yu.G.; Skiba, O.V.; Gazizov, R.K.; Vorobey, M.P.

    The composition and certain physicochemical and radiation properties of waste obtained from reprocessing BOR-60 fuel by the gaseous fluoride method were studied on an experimental stand. It was found that from 80 to 85 percent of the total activity is concentrated in the fluorination cakes. From thermographic analysis, measurement of radiation-induced gas evolution and composition of the gaseous phase, it is concluded that the waste must be canned. Compatibility of carbon steel, austenitic steel, and nickel with low-activity fluoride cakes was studied. It is shown that the corrosion of these materials increases with temperature and moisture content of the medium. In the first stage of waste disposal, hermetic storage in special containers of stainless steel or nickel is recommended. From their corrosion resistance in low-activity fluoride cakes, the wall temperatures of containers made of St. 3 should not exceed 100 0 C, Kh18N10T 230 0 C, and nickel 300 0 C. Later, after cooling and a decrease in the γ activity of the waste (10 to 15 years), they can be reprocessed to recover valuable components (Am, Cm, etc.) and then buried. 5 tables

  3. Vitrification of high level wastes in France

    International Nuclear Information System (INIS)

    Sombret, C.

    1984-02-01

    A brief historical background of the research and development work conducted in France over 25 years is first presented. Then, the papers deals with the vitrification at (1) the UP1 reprocessing plant (Marcoule) and (2) the UP2 and UP3 reprocessing plants (La Hague). 1) The properties of glass required for high-level radioactive waste vitrification are recalled. The vitrification process and facility of Marcoule are presented. (2) The average characteristics (chemical composition, activity) of LWR fission product solution are given. The glass formulations developed to solidify LWR waste solution must meet the same requirements as those used in the UP1 facility at Marcoule. Three important aspects must be considered with respect to the glass fabrication process: corrosiveness of the molten glass with regard to metals, viscosity of the molten glass, and, volatization during glass fabrication. The glass properties required in view of interim storage and long-term disposal are then largely developed. Two identical vitrification facilities are planned for the site: T7, to process the UP2 throughput, and T7 for the UP3 plant. A prototype unit was built and operated at Marcoule

  4. Proceedings of the public debate on the general options of management of high and intermediate activity and long-lived radioactive wastes

    International Nuclear Information System (INIS)

    2006-01-01

    The law from December 30, 1991, precisely defines 3 axes of researches for the management of high level and long-lived radioactive wastes: separation/transmutation, surface storage and underground disposal. A global evaluation report about these researches is to be supplied in 2006 by the French government to the Parliament. A first synthesis of the knowledge gained after 14 years of research has led the national commission of the public debate (CNDP) to organize a national debate about the general options of management of high-level and long-lived radioactive wastes before the 2006 date line. The debate comprised 4 public hearings (September 2005: Bar-le-Duc, Saint-Dizier, Pont-du-Gard, Cherbourg), 12 round-tables (October and November 2005: Paris, Joinville, Caen, Nancy, Marseille), a synthesis meeting (December 2005, Dunkerque) and a closing meeting (January 2006, Lyon). This document is a summary of the main questions tackled during this debate from its preparation to its closing meeting and dealing with: the acceptation of the debate, the progress of the debate, the socio-technical problem of gathering the different points to be debated with respect to the different areas in concern, the general questions about the overall nuclear wastes and materials (radioactivity, health, radioprotection, management, reprocessing, control, actors organization, knowledge sharing, perenniality of the financing), the specific questions about long-lived wastes (inventory, separation-transmutation feasibility, nuclear energy and energy policy, management solutions, storage and geologic disposal feasibility, impact of debates on the 2006 law, long-lived waste territories), the conclusions for the 2006 law: mastering the overall nuclear wastes and materials and step-by-step building up of a solution for long-lived wastes (difficulties, lessons learnt from foreign experience, first draft of the 2006 law). Some opinions expressed by some participants about these proposals conclude

  5. Development of high-level waste solidification technology 1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Joon Hyung; Kim, Hwan Young; Kim, In Tae [and others

    1999-02-01

    Spent nuclear fuel contains useful nuclides as valuable resource materials for energy, heat and catalyst. High-level wastes (HLW) are expected to be generated from the R and D activities and reuse processes. It is necessary to develop vitrification or advanced solidification technologies for the safe long-term management of high level wastes. As a first step to establish HLW vitrification technology, characterization of HLWs that would arise at KAERI site, glass melting experiments with a lab-scale high frequency induction melter, and fabrication and property evaluation of base-glass made of used HEPA filter media and additives were performed. Basic study on the fabrication and characterization of candidate ceramic waste form (Synroc) was also carried out. These HLW solidification technologies would be directly useful for carrying out the R and Ds on the nuclear fuel cycle and waste management. (author). 70 refs., 29 tabs., 35 figs.

  6. Technical dialogue on High-level and Intermediate-level wastes. Risks during exploitation: co-activity

    International Nuclear Information System (INIS)

    Sene, Monique; Lheureux, Yves; Demet, Michel; Jaquet, Benoit; Gueritte, Michel; Rollinger, Francois; Le Bars, Igor; Bauduin, Eloi; Flachet, Margot; Besnus, Francois; Espivent, Camille; Tichauer, Michael; Serres, Christophe; Rocher, Muriel; Michel, Maurice; Perroud, Geraldine; Sauvage, Marite; Stupien, Maurice; Delamare, Thierry; Klajman, Irene; Castel, Cecile; Makhijani, Arjun; Pellegrini, Delphine; Dessaint, Anthony; Proust, Christophe; Gratton, Laura; Vinot, Thierry; Michel, Maurice; Demet, Michel

    2015-04-01

    This publication contains contributions and the content of debates on these contributions. After introduction speeches, a contribution on safety challenges during exploitation, and an overview of questions asked about these risks by the Citizen Conference, the following issues have been addressed: the accident of the American Isolation Pilot Plant with a fire and a radioactive release in 2014 (facts and analysis, other points of view, debate), fire risks in Cigeo as a major challenge which combines constraints related to mining activities and to the exploitation of a nuclear installation (IRSN point of view, debate), risks related to transportation, risks related to parallel surface and underground activities such as storage filling, digging, road traffic and so on (points of view of SNCF and ANDRA, debate), risk of radioactivity dispersion within Cigeo (problem description, IRSN point of view, debate), and exploitation risks within Cigeo (risks of explosion due to heat and hydrogen releases, IRSN point of view, debate)

  7. Activity monitoring of alpha-bearing wastes

    International Nuclear Information System (INIS)

    Birkhoff, G.; Bondar, L.

    1980-01-01

    The paper aims at the survey on the actual situation in activity monitoring of alpha-bearing wastes. Homogeneous materials such as liquid-, gaseous- and homogeneous solid wastes are amenable to destructive analyses of representative samples. Available destructive analyses methods are sensitive and precise enough to cope with all requirements in alpha-waste monitoring. The more difficult problems are encountered with alpha-contaminated solids, when representative sampling is not practicable. Non-destructive analysis techniques are applied for monitoring this category of solid wastes. The techniques for nondestructive analysis of alpha-bearing wastes are based on the detection of gamma and/or neutron-emission of actinides. Principles and a theory of non-destructive radiometric assay of plutonium contaminated solid waste streams are explained. Guidelines for the calibration of instruments and interpretation of experimental data are given. Current theoretical and experimental development work in this problem area is reviewed. Evaluations concerning capabilities and limitations of monitoring systems for alpha-bearing solid wastes are very complex and out of the scope of this paper

  8. High-level waste processing and disposal

    International Nuclear Information System (INIS)

    Crandall, J.L.; Krause, H.; Sombret, C.; Uematsu, K.

    1984-01-01

    The national high-level waste disposal plans for France, the Federal Republic of Germany, Japan, and the United States are covered. Three conclusions are reached. The first conclusion is that an excellent technology already exists for high-level waste disposal. With appropriate packaging, spent fuel seems to be an acceptable waste form. Borosilicate glass reprocessing waste forms are well understood, in production in France, and scheduled for production in the next few years in a number of other countries. For final disposal, a number of candidate geological repository sites have been identified and several demonstration sites opened. The second conclusion is that adequate financing and a legal basis for waste disposal are in place in most countries. Costs of high-level waste disposal will probably add about 5 to 10% to the costs of nuclear electric power. The third conclusion is less optimistic. Political problems remain formidable in highly conservative regulations, in qualifying a final disposal site, and in securing acceptable transport routes

  9. High-Level Waste Melter Study Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-13

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

  10. 40 CFR 227.30 - High-level radioactive waste.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false High-level radioactive waste. 227.30...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from the operation of the first cycle solvent extraction system, or equivalent, and the concentrated waste from...

  11. Microbial Insight into a Pilot-Scale Enhanced Two-Stage High-Solid Anaerobic Digestion System Treating Waste Activated Sludge.

    Science.gov (United States)

    Wu, Jing; Cao, Zhiping; Hu, Yuying; Wang, Xiaolu; Wang, Guangqi; Zuo, Jiane; Wang, Kaijun; Qian, Yi

    2017-11-30

    High solid anaerobic digestion (HSAD) is a rapidly developed anaerobic digestion technique for treating municipal sludge, and has been widely used in Europe and Asia. Recently, the enhanced HSAD process with thermal treatment showed its advantages in both methane production and VS reduction. However, the understanding of the microbial community is still poor. This study investigated microbial communities in a pilot enhanced two-stage HSAD system that degraded waste activated sludge at 9% solid content. The system employed process "thermal pre-treatment (TPT) at 70 °C, thermophilic anaerobic digestion (TAD), and mesophilic anaerobic digestion (MAD)". Hydrogenotrophic methanogens Methanothermobacter spp. dominated the system with relative abundance up to about 100% in both TAD and MAD. Syntrophic acetate oxidation (SAO) bacteria were discovered in TAD, and they converted acetate into H₂ and CO₂ to support hydrogenotrophic methanogenesis. The microbial composition and conversion route of this system are derived from the high solid content and protein content in raw sludge, as well as the operational conditions. This study could facilitate the understanding of the enhanced HSAD process, and is of academic and industrial importance.

  12. Treatment of rod shaped intermediate active waste

    International Nuclear Information System (INIS)

    Graf, A.; Blase, F.; Dirks, F.; Valencia, L.

    2002-01-01

    The Central Decontamination Operation Department (HDB) of the Research Center Karlsruhe operates facilities for the disposal of radioactive waste. In general, their objective is to reduce the volume of the radioactive waste and to obtain waste products suitable for repository storage. One of the central facilities of the HDB is the intermediate level waste (ILW) scrapping facility which processes intermediate level waste. Since the ILW scrapping facility was not large enough to handle radioactive waste coming from the dismantling and operating of nuclear facilities, HDB expanded and built a larger hot cell. It contains a hydraulically driven metal cutter with a guiding channel and a high pressure compactor. A major task in the hot cell of the ILW scrapping facility is disposing of fuel boxes. These are cut in pieces and scrapped, which is a unique technique in Germany for fuel box disposal. HDB's experiences in disposing of radioactive waste in the ILW scrapping facility will described in detail, with special emphasis on the handling of rod shaped components. (author)

  13. Evaluation of conditioned high-level waste forms

    International Nuclear Information System (INIS)

    Mendel, J.E.; Turcotte, R.P.; Chikalla, T.D.; Hench, L.L.

    1983-01-01

    The evaluation of conditioned high-level waste forms requires an understanding of radiation and thermal effects, mechanical properties, volatility, and chemical durability. As a result of nuclear waste research and development programs in many countries, a good understanding of these factors is available for borosilicate glass containing high-level waste. The IAEA through its coordinated research program has contributed to this understanding. Methods used in the evaluation of conditioned high-level waste forms are reviewed. In the US, this evaluation has been facilitated by the definition of standard test methods by the Materials Characterization Center (MCC), which was established by the Department of Energy (DOE) in 1979. The DOE has also established a 20-member Materials Review Board to peer-review the activities of the MCC. In addition to comparing waste forms, testing must be done to evaluate the behavior of waste forms in geologic repositories. Such testing is complex; accelerated tests are required to predict expected behavior for thousands of years. The tests must be multicomponent tests to ensure that all potential interactions between waste form, canister/overpack and corrosion products, backfill, intruding ground water and the repository rock, are accounted for. An overview of the status of such multicomponent testing is presented

  14. High Solids Consolidated Incinerator Facility (CIF) Wastes Stabilization with Ceramicrete and Super Cement

    International Nuclear Information System (INIS)

    Walker, B.W.

    1999-01-01

    High Solids ash and scrubber solution waste streams were generated at the incinerator facility at SRS by burning radioactive diatomaceous filter rolls which contained small amounts of uranium, and listed solvents (F and U). This report details solidification activities using selected Mixed Waste Focus Area (MWFA) technologies with the High Solids waste streams

  15. Evaluation of radionuclide concentrations in high-level radioactive wastes

    International Nuclear Information System (INIS)

    Fehringer, D.J.

    1985-10-01

    This report describes a possible approach for development of a numerical definition of the term ''high-level radioactive waste.'' Five wastes are identified which are recognized as being high-level wastes under current, non-numerical definitions. The constituents of these wastes are examined and the most hazardous component radionuclides are identified. This report suggests that other wastes with similar concentrations of these radionuclides could also be defined as high-level wastes. 15 refs., 9 figs., 4 tabs

  16. High level waste fixation in cermet form

    International Nuclear Information System (INIS)

    Kobisk, E.H.; Aaron, W.S.; Quinby, T.C.; Ramey, D.W.

    1981-01-01

    Commercial and defense high level waste fixation in cermet form is being studied by personnel of the Isotopes Research Materials Laboratory, Solid State Division (ORNL). As a corollary to earlier research and development in forming high density ceramic and cermet rods, disks, and other shapes using separated isotopes, similar chemical and physical processing methods have been applied to synthetic and real waste fixation. Generally, experimental products resulting from this approach have shown physical and chemical characteristics which are deemed suitable for long-term storage, shipping, corrosive environments, high temperature environments, high waste loading, decay heat dissipation, and radiation damage. Although leach tests are not conclusive, what little comparative data are available show cermet to withstand hydrothermal conditions in water and brine solutions. The Soxhlet leach test, using radioactive cesium as a tracer, showed that leaching of cermet was about X100 less than that of 78 to 68 glass. Using essentially uncooled, untreated waste, cermet fixation was found to accommodate up to 75% waste loading and yet, because of its high thermal conductivity, a monolith of 0.6 m diameter and 3.3 m-length would have only a maximum centerline temperature of 29 K above the ambient value

  17. International high-level radioactive waste repositories

    International Nuclear Information System (INIS)

    Lin, W.

    1996-01-01

    Although nuclear technologies benefit everyone, the associated nuclear wastes are a widespread and rapidly growing problem. Nuclear power plants are in operation in 25 countries, and are under construction in others. Developing countries are hungry for electricity to promote economic growth; industrialized countries are eager to export nuclear technologies and equipment. These two ingredients, combined with the rapid shrinkage of worldwide fossil fuel reserves, will increase the utilization of nuclear power. All countries utilizing nuclear power produce at least a few tens of tons of spent fuel per year. That spent fuel (and reprocessing products, if any) constitutes high-level nuclear waste. Toxicity, long half-life, and immunity to chemical degradation make such waste an almost permanent threat to human beings. This report discusses the advantages of utilizing repositories for disposal of nuclear wastes

  18. Solidification of high-level radioactive wastes. Final report

    International Nuclear Information System (INIS)

    1979-06-01

    A panel on waste solidification was formed at the request of the Nuclear Regulatory Commission to study the scientific and technological problems associated with the conversion of liquid and semiliquid high-level radioactive wastes into a stable form suitable for transportation and disposition. Conclusions reached and recommendations made are as follows. Many solid forms described in this report could meet standards as stringent as those currently applied to the handling, storage, and transportation of spent fuel assemblies. Solid waste forms should be selected only in the context of the total radioactive waste management system. Many solid forms are likely to be satisfactory for use in an appropriately designed system, The current United States policy of deferring the reprocessing of commercial reactor fuel provides additional time for R and D solidification technology for this class of wastes. Defense wastes which are relatively low in radioactivity and thermal power density can best be solidified by low-temperature processes. For solidification of fresh commercial wastes that are high in specific activity and thermal power density, the Panel recommends that, in addition to glass, the use of fully-crystalline ceramics and metal-matrix forms be actively considered. Preliminary analysis of the characteristics of spent fuel pins indicates that they may be eligible for consideration as a waste form. Because the differences in potential health hazards to the public resulting from the use of various solid form and disposal options are likely to be small, the Panel concludes that cost, reliability, and health hazards to operating personnel will be major considerations in choosing among the options that can meet safety requiremens. The Panel recommends that responsibility for all radioactive waste management operations (including solidification R and D) should be centralized

  19. Use of polyethers in the treatment of acidic high activity nuclear wastes. First progress report, June 1, 1975--May 31, 1977

    International Nuclear Information System (INIS)

    Davis, M.W. Jr.

    1977-01-01

    Objective is to develop a solvent extraction system for treating acidic high activity nuclear wastes in order to remove 99.9% of the cesium and strontium. A series of macrocyclic polyethers ''crown compounds'' which can form organic soluble complexes with the alkali metals (Cs +1 ) and the alkali earth metals (Sr +2 ) are being screened. Thus far, five ''crown'' compounds have been tested for solubility in both carbon tetrachloride and kerosene, each solvent containing a range of TBP concentrations. Also distribution coefficients of both cesium and strontium have been measured. Results have been negative in that sufficient extraction of either Cs +1 or Sr +2 into the organic phase was not achieved to make a solvent extraction process feasible. The highest distribution coefficient obtained (organic/aqueous)/sup (2)/ was 0.04 +- 0.02 for both Sr +2 and Cs +1 . The problem seems to lie in the low solubility of the ''crowns'' in the organic phases and the apparent ability of the ''crowns'' to distribute into the aqueous phase and complex the metal. The solubility data indicate that the addition of aliphatic character to the ''crowns'' improves their organic solubility whereas aromatic addition decreases their organic solubility. Accordingly, a new ''crown'', 4, 4'-didecanolybenzo 24-crown-8, was ordered from the Parish Chemical Company in Provo, Utah

  20. Submission of the national commission of the public debate on the options concerning the long life high and medium activity radioactive wastes management; Saisine de la commission nationale du debat public sur les options generales en matiere de gestion des dechets radioactifs de haute activite et de moyenne activite a vie longue

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    This document deals with the presentation of a public debate on the radioactive wastes management and the opportunities of its organization. It presents successively the long life high and medium activity radioactive wastes, the today radioactive wastes management policy and some questions and topics which could be discussed during the debate. (A.L.B.)

  1. Disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    Glasby, G.P.

    1977-01-01

    Although controversy surrounding the possible introduction of nuclear power into New Zealand has raised many points including radiation hazards, reactor safety, capital costs, sources of uranium and earthquake risks on the one hand versus energy conservation and alternative sources of energy on the other, one problem remains paramount and is of global significance - the storage and dumping of the high-level radioactive wastes of the reactor core. The generation of abundant supplies of energy now in return for the storage of these long-lived highly radioactive wastes has been dubbed the so-called Faustian bargain. This article discusses the growth of the nuclear industry and its implications to high-level waste disposal particularly in the deep-sea bed. (auth.)

  2. The Defense Waste Processing Facility: an innovative process for high-level waste immobilization

    International Nuclear Information System (INIS)

    Cowan, S.P.

    1985-01-01

    The Defense Waste Processing Facility (DWPF), under construction at the Department of Energy's Savannah River Plant (SRP), will process defense high-level radioactive waste so that it can be disposed of safely. The DWPF will immobilize the high activity fraction of the waste in borosilicate glass cast in stainless steel canisters which can be handled, stored, transported and disposed of in a geologic repository. The low-activity fraction of the waste, which represents about 90% of the high-level waste HLW volume, will be decontaminated and disposed of on the SRP site. After decontamination the canister will be welded shut by an upset resistance welding technique. In this process a slightly oversized plug is pressed into the canister opening. At the same time a large current is passed through the canister and plug. The higher resistance of the canister/plug interface causes the heat which welds the plug in place. This process provides a high quality, reliable weld by a process easily operated remotely

  3. Compaction and packaging of dry active municipal wastes

    International Nuclear Information System (INIS)

    Chen Zongming; Xi Xinmin

    1994-01-01

    The authors present the feature of a compaction system for active municipal wastes and the radiological monitoring results of workplace and environment. A variety of dry active municipal wastes could be compacted by this system. Volume reduction factor attained to 5 to 7 for soft wastes and 8 to 13 for hard wastes. No evident radiological impact was found on workplace and environment

  4. Nuclear waste: Status of DOE's nuclear waste site characterization activities

    International Nuclear Information System (INIS)

    1987-01-01

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

  5. Distinguishing method for contamination/radio-activation of radioactive wastes

    International Nuclear Information System (INIS)

    Fukazawa, Takuji; Kato, Keiichiro; Koda, Satoshi.

    1994-01-01

    The present invention concerns a method of distinguishing the contamination/radio-activation of radioactive wastes used in processing wastes generated upon dismantling of exhausted nuclear reactors. Especially, contaminated/radio-activation is distinguished for wastes having openings such as pipes and valves, by utilizing scattering of γ-rays or γ-ray to β-ray ratio. That is, ratio of scattered γ-rays and direct γ-rays or ratio of β-rays and γ-rays from radioactive wastes are measured and compared by a radiation detector, to distinguish whether the radioactive wastes contaminated materials or radio-activated materials. For example, when an object to be measured having an opening is contaminated at the inner side, the radiation detector facing to the opening mainly detects high direct γ-rays emitted from the object to be measured while a radiation detector not facing the opening mainly detects high scattered γ-rays relatively. On the other hand, when the object is a radio-activated material, any of the detectors detect scattered γ-rays, so that they can be distinguished by these ratios. (I.S.)

  6. Low Activity Waste Feed Process Control Strategy

    International Nuclear Information System (INIS)

    STAEHR, T.W.

    2000-01-01

    The primary purpose of this document is to describe the overall process control strategy for monitoring and controlling the functions associated with the Phase 1B high-level waste feed delivery. This document provides the basis for process monitoring and control functions and requirements needed throughput the double-shell tank system during Phase 1 high-level waste feed delivery. This document is intended to be used by (1) the developers of the future Process Control Plan and (2) the developers of the monitoring and control system

  7. Radioecological activity limits for radioactive waste disposal

    International Nuclear Information System (INIS)

    Ahmet, E. Osmanlioglu

    2006-01-01

    Full text: Near surface disposal is an option used by many countries for the disposal of radioactive waste containing mainly short lived radionuclides. Near surface disposal term includes broad range of facilities from simple trenches to concrete vaults. Principally, disposal of radioactive waste requires the implementation of measures that will provide safety for human health and environment now and in the future. For this reason preliminary activity limits should be determined to avoid radioecological problems. Radioactive waste has to be safely disposed in a regulated manner, consistent with internationally agreed principles and standards and with national legislations to avoid serious radioecological problems. The purpose of this study, presents a safety assessment approach to derive operational and post-closure radioecological activity limits for the disposal of radioactive waste. Disposal system has three components; the waste, the facility (incl. engineered barriers) and the site (natural barriers). Form of the waste (unconditioned or conditioned) is effective at the beginning of the migration scenerio. Existence of the engineered barriers in the facility will provide long term isolation of the waste from environment. The site characteristics (geology, groundwater, seismicity, climate etc.) are important for the safety of the system. Occupational exposure of a worker shall be controlled so that the following dose limits are not exceeded: an effective dose of 20mSv/y averaged over 5 consecutive years; and an effective dose of 50mSv in any single year. The effective dose limit for members of the public recommended by ICRP and IAEA is 1 mSv/y for exposures from all man-made sources [1,2]. Dose constraints are typically a fraction of the dose limit and ICRP recommendations (0.3 mSv/y) could be applied [3,4]. Radioecological activity concentration limits of each radionuclide in the waste (Bq/kg) were calculated. As a result of this study radioecological activity

  8. Timing of High-level Waste Disposal

    International Nuclear Information System (INIS)

    2008-01-01

    This study identifies key factors influencing the timing of high-level waste (HLW) disposal and examines how social acceptability, technical soundness, environmental responsibility and economic feasibility impact on national strategies for HLW management and disposal. Based on case study analyses, it also presents the strategic approaches adopted in a number of national policies to address public concerns and civil society requirements regarding long-term stewardship of high-level radioactive waste. The findings and conclusions of the study confirm the importance of informing all stakeholders and involving them in the decision-making process in order to implement HLW disposal strategies successfully. This study will be of considerable interest to nuclear energy policy makers and analysts as well as to experts in the area of radioactive waste management and disposal. (author)

  9. Final disposal of high levels waste and spent nuclear fuel

    International Nuclear Information System (INIS)

    Gelin, R.

    1984-05-01

    Foreign and international activities on the final disposal of high-level waste and spent nuclear fuel have been reviewed. A considerable research effort is devoted to development of acceptable disposal options. The different technical concepts presently under study are described in the report. Numerous studies have been made in many countries of the potential risks to future generations from radioactive wastes in underground disposal repositories. In the report the safety assessment studies and existing performance criteria for geological disposal are briefly discussed. The studies that are being made in Canada, the United States, France and Switzerland are the most interesting for Sweden as these countries also are considering disposal into crystalline rocks. The overall time-tables in different countries for realisation of the final disposal are rather similar. Normally actual large-scale disposal operations for high-level wastes are not foreseen until after year 2000. In the United States the Congress recently passed the important Nuclear Waste Policy Act. It gives a rather firm timetable for site-selection and construction of nuclear waste disposal facilities. According to this act the first repository for disposal of commercial high-level waste must be in operation not later than in January 1998. (Author)

  10. Materials Science of High-Level Nuclear Waste Immobilization

    International Nuclear Information System (INIS)

    Weber, William J.; Navrotsky, Alexandra; Stefanovsky, S. V.; Vance, E. R.; Vernaz, Etienne Y.

    2009-01-01

    With the increasing demand for the development of more nuclear power comes the responsibility to address the technical challenges of immobilizing high-level nuclear wastes in stable solid forms for interim storage or disposition in geologic repositories. The immobilization of high-level nuclear wastes has been an active area of research and development for over 50 years. Borosilicate glasses and complex ceramic composites have been developed to meet many technical challenges and current needs, although regulatory issues, which vary widely from country to country, have yet to be resolved. Cooperative international programs to develop advanced proliferation-resistant nuclear technologies to close the nuclear fuel cycle and increase the efficiency of nuclear energy production might create new separation waste streams that could demand new concepts and materials for nuclear waste immobilization. This article reviews the current state-of-the-art understanding regarding the materials science of glasses and ceramics for the immobilization of high-level nuclear waste and excess nuclear materials and discusses approaches to address new waste streams

  11. Strategic lessons in high-level waste management planning

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, Neil

    1999-07-01

    This presentation discusses some issues in the planning and execution of high-level waste (HLW) disposal. The topics are (1) Initial considerations, (2) Issues in structuring a programme, (3) Disposal concepts, (4) Geological environments, (5) Site selection and characterisation, (6) Waste transport, (7) Performance assessment methodology and application, (8) Some key issues. The options for spent fuel management can give rise to a variety of different wastes. The quantity of waste arising will affect the volume of rock required for deposition, both with respect to rock integrity and requirements for heat dissipation. A repository must not be considered in isolation from the rest of the waste management programme. The repository development plan should be supported by a schedule of activities and related funding mechanisms, implying a long-term commitment in policy terms, and should include a corresponding legal and regulatory framework. The idea that disposed waste might be retrieved by future generations for processing under new technology is discussed. Safeguards requirements on fissile material within spent fuel or any other wastes imply indefinite control. Disposal concepts include the geological environment and the engineered barrier system within it. Site selection involves several steps: regional-scale characterisation, local characterisation, hydrological studies, etc. Key issues are retrieval vs. safeguards, optimisation of repository design, reducing long programme timescales, international collaboration.

  12. Strategic lessons in high-level waste management planning

    International Nuclear Information System (INIS)

    Chapman, Neil

    1999-01-01

    This presentation discusses some issues in the planning and execution of high-level waste (HLW) disposal. The topics are (1) Initial considerations, (2) Issues in structuring a programme, (3) Disposal concepts, (4) Geological environments, (5) Site selection and characterisation, (6) Waste transport, (7) Performance assessment methodology and application, (8) Some key issues. The options for spent fuel management can give rise to a variety of different wastes. The quantity of waste arising will affect the volume of rock required for deposition, both with respect to rock integrity and requirements for heat dissipation. A repository must not be considered in isolation from the rest of the waste management programme. The repository development plan should be supported by a schedule of activities and related funding mechanisms, implying a long-term commitment in policy terms, and should include a corresponding legal and regulatory framework. The idea that disposed waste might be retrieved by future generations for processing under new technology is discussed. Safeguards requirements on fissile material within spent fuel or any other wastes imply indefinite control. Disposal concepts include the geological environment and the engineered barrier system within it. Site selection involves several steps: regional-scale characterisation, local characterisation, hydrological studies, etc. Key issues are retrieval vs. safeguards, optimisation of repository design, reducing long programme timescales, international collaboration

  13. Transportation research activities in support of nuclear waste management programs

    International Nuclear Information System (INIS)

    Allen, G.C. Jr.; Cashwell, J.W.; Jefferson, R.M.

    1983-01-01

    Transportation Technology Center has been conducting a wide range of technical research activities to assure the ability to transport radioactive materials in a safe, reliable manner. These activities include tasks in basic, analysis methodology and system research areas. Recently, the requirements of defense waste shipments have served as a focal point for development tasks with the expectation that they would serve as a precursor for commercial activities. The passage of the Nuclear Waste Policy Act has placed additional responsibility on the Department of Energy for concerns involving the shipments of civilian materials. The development of additional research responsibilities is expected to proceed concurrently with the evolution of the transportation mission plan for civilian spent fuel and high-level wastes

  14. Transportation research activities in support of nuclear waste management programs

    International Nuclear Information System (INIS)

    Allen, G.C.; Luna, R.E.; Jefferson, R.M.; Wowak, W.E.

    1983-01-01

    The Transportation Technology Center has been conducting a wide range of technical and non-technical research activities to assure the ability to transport radioactive materials in a safe, reliable, and publicly acceptable manner. These activities include tasks in Information and Intergovernmental issues, Safety Assessment and Environmental Analysis and Technology Development. Until recently, the requirements of defense waste shipments have served as a focal point for development tasks with the expectation that they would serve as a precursor for commercial activities. The passage of the Nuclear Waste Policy Act has placed additional responsibility on DOE for concerns involving the shipments of civilian materials. The development of additional research responsibilities is expected to proceed concurrently with the evolution of the transportation mission plan for civilian spent fuel and high-level wastes

  15. Public debate on the general options relative to the management of high-medium activity and long-lived radioactive wastes

    International Nuclear Information System (INIS)

    2005-09-01

    The law from December 30, 1991, precisely defines 3 axes of researches for the management of high level and long-lived radioactive wastes: separation/transmutation, surface storage and underground disposal. A global evaluation report about these researches is to be supplied in 2006 by the French government to the Parliament. A first synthesis of the knowledge gained after 14 years of research has led the national commission of the public debate (CNDP) to organize a national debate about the general options of management of high-level and long-lived radioactive wastes before the 2006 date line. This document presents the organizational aspects of the debate: origin, organization committee (mission, members, commitments), framework (the December 31, 1991 law, technical enlargement, society aspects), topics (summary of the debate in ten questions), organization modalities (4 main steps, schedule, venues), objectives and perspectives (sharing information, decision making processes to be implemented by 2006). (J.S.)

  16. Technetium Chemistry in High-Level Waste

    International Nuclear Information System (INIS)

    Hess, Nancy J.

    2006-01-01

    Tc contamination is found within the DOE complex at those sites whose mission involved extraction of plutonium from irradiated uranium fuel or isotopic enrichment of uranium. At the Hanford Site, chemical separations and extraction processes generated large amounts of high level and transuranic wastes that are currently stored in underground tanks. The waste from these extraction processes is currently stored in underground High Level Waste (HLW) tanks. However, the chemistry of the HLW in any given tank is greatly complicated by repeated efforts to reduce volume and recover isotopes. These processes ultimately resulted in mixing of waste streams from different processes. As a result, the chemistry and the fate of Tc in HLW tanks are not well understood. This lack of understanding has been made evident in the failed efforts to leach Tc from sludge and to remove Tc from supernatants prior to immobilization. Although recent interest in Tc chemistry has shifted from pretreatment chemistry to waste residuals, both needs are served by a fundamental understanding of Tc chemistry

  17. High-level waste melter alternatives assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Calmus, R.B.

    1995-02-01

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

  18. High-level waste melter alternatives assessment report

    International Nuclear Information System (INIS)

    Calmus, R.B.

    1995-02-01

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

  19. High-level waste management technology program plan

    International Nuclear Information System (INIS)

    Harmon, H.D.

    1995-01-01

    The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs

  20. High-level waste management technology program plan

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, H.D.

    1995-01-01

    The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs.

  1. Waste Treatment Technology Process Development Plan For Hanford Waste Treatment Plant Low Activity Waste Recycle

    Energy Technology Data Exchange (ETDEWEB)

    McCabe, Daniel J.; Wilmarth, William R.; Nash, Charles A.

    2013-08-29

    The purpose of this Process Development Plan is to summarize the objectives and plans for the technology development activities for an alternative path for disposition of the recycle stream that will be generated in the Hanford Waste Treatment Plant Low Activity Waste (LAW) vitrification facility (LAW Recycle). This plan covers the first phase of the development activities. The baseline plan for disposition of this stream is to recycle it to the WTP Pretreatment Facility, where it will be concentrated by evaporation and returned to the LAW vitrification facility. Because this stream contains components that are volatile at melter temperatures and are also problematic for the glass waste form, they accumulate in the Recycle stream, exacerbating their impact on the number of LAW glass containers. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and reducing the halides in the Recycle is a key component of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, this stream does not have a proven disposition path, and resolving this gap becomes vitally important. This task seeks to examine the impact of potential future disposition of this stream in the Hanford tank farms, and to develop a process that will remove radionuclides from this stream and allow its diversion to another disposition path, greatly decreasing the LAW vitrification mission duration and quantity of glass waste. The origin of this LAW Recycle stream will be from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover or precipitates of scrubbed components (e.g. carbonates). The soluble

  2. Site suitability criteria for solidified high level waste repositories

    International Nuclear Information System (INIS)

    Heckman, R.A.; Holdsworth, T.; Towse, D.F.

    1979-01-01

    Activities devoted to development of regulations, criteria, and standards for storage of solidified high-level radioactive wastes are reported. The work is summarized in sections on site suitability regulations, risk calculations, geological models, aquifer models, human usage model, climatology model, and repository characteristics. Proposed additional analytical work is also summarized

  3. Hanford immobilized low-activity tank waste performance assessment

    International Nuclear Information System (INIS)

    Mann, F.M.

    1998-01-01

    The Hanford Immobilized Low-Activity Tank Waste Performance Assessment examines the long-term environmental and human health effects associated with the planned disposal of the vitrified low-level fraction of waste presently contained in Hanford Site tanks. The tank waste is the by-product of separating special nuclear materials from irradiated nuclear fuels over the past 50 years. This waste has been stored in underground single and double-shell tanks. The tank waste is to be retrieved, separated into low and high-activity fractions, and then immobilized by private vendors. The US Department of Energy (DOE) will receive the vitrified waste from private vendors and plans to dispose of the low-activity fraction in the Hanford Site 200 East Area. The high-level fraction will be stored at Hanford until a national repository is approved. This report provides the site-specific long-term environmental information needed by the DOE to issue a Disposal Authorization Statement that would allow the modification of the four existing concrete disposal vaults to provide better access for emplacement of the immobilized low-activity waste (ILAW) containers; filling of the modified vaults with the approximately 5,000 ILAW containers and filler material with the intent to dispose of the containers; construction of the first set of next-generation disposal facilities. The performance assessment activity will continue beyond this assessment. The activity will collect additional data on the geotechnical features of the disposal sites, the disposal facility design and construction, and the long-term performance of the waste. Better estimates of long-term performance will be produced and reviewed on a regular basis. Performance assessments supporting closure of filled facilities will be issued seeking approval of those actions necessary to conclude active disposal facility operations. This report also analyzes the long-term performance of the currently planned disposal system as a basis

  4. Hanford immobilized low-activity tank waste performance assessment

    Energy Technology Data Exchange (ETDEWEB)

    Mann, F.M.

    1998-03-26

    The Hanford Immobilized Low-Activity Tank Waste Performance Assessment examines the long-term environmental and human health effects associated with the planned disposal of the vitrified low-level fraction of waste presently contained in Hanford Site tanks. The tank waste is the by-product of separating special nuclear materials from irradiated nuclear fuels over the past 50 years. This waste has been stored in underground single and double-shell tanks. The tank waste is to be retrieved, separated into low and high-activity fractions, and then immobilized by private vendors. The US Department of Energy (DOE) will receive the vitrified waste from private vendors and plans to dispose of the low-activity fraction in the Hanford Site 200 East Area. The high-level fraction will be stored at Hanford until a national repository is approved. This report provides the site-specific long-term environmental information needed by the DOE to issue a Disposal Authorization Statement that would allow the modification of the four existing concrete disposal vaults to provide better access for emplacement of the immobilized low-activity waste (ILAW) containers; filling of the modified vaults with the approximately 5,000 ILAW containers and filler material with the intent to dispose of the containers; construction of the first set of next-generation disposal facilities. The performance assessment activity will continue beyond this assessment. The activity will collect additional data on the geotechnical features of the disposal sites, the disposal facility design and construction, and the long-term performance of the waste. Better estimates of long-term performance will be produced and reviewed on a regular basis. Performance assessments supporting closure of filled facilities will be issued seeking approval of those actions necessary to conclude active disposal facility operations. This report also analyzes the long-term performance of the currently planned disposal system as a basis

  5. Multipurpose optimization models for high level waste vitrification

    International Nuclear Information System (INIS)

    Hoza, M.

    1994-08-01

    Optimal Waste Loading (OWL) models have been developed as multipurpose tools for high-level waste studies for the Tank Waste Remediation Program at Hanford. Using nonlinear programming techniques, these models maximize the waste loading of the vitrified waste and optimize the glass formers composition such that the glass produced has the appropriate properties within the melter, and the resultant vitrified waste form meets the requirements for disposal. The OWL model can be used for a single waste stream or for blended streams. The models can determine optimal continuous blends or optimal discrete blends of a number of different wastes. The OWL models have been used to identify the most restrictive constraints, to evaluate prospective waste pretreatment methods, to formulate and evaluate blending strategies, and to determine the impacts of variability in the wastes. The OWL models will be used to aid in the design of frits and the maximize the waste in the glass for High-Level Waste (HLW) vitrification

  6. High-level radioactive wastes. Supplement 1

    International Nuclear Information System (INIS)

    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. Materials for high-level waste containment

    International Nuclear Information System (INIS)

    Marsh, G.P.

    1982-01-01

    The function of the high-level radioactive waste container in storage and of a container/overpack combination in disposal is considered. The consequent properties required from potential fabrication materials are discussed. The strategy adopted in selecting containment materials and the experimental programme underway to evaluate them are described. (U.K.)

  8. Hanford Waste Vitrification Plant quality assurance program description for defense high-level waste form development and qualification

    International Nuclear Information System (INIS)

    Hand, R.L.

    1990-12-01

    The US Department of Energy-Office of Civilian Radioactive Waste Management has been designated the national high-level waste repository licensee and the recipient for the canistered waste forms. The Office of Waste Operations executes overall responsibility for producing the canistered waste form. The Hanford Waste Vitrification Plant Project, as part of the waste form producer organization, consists of a vertical relationship. Overall control is provided by the US Department of Energy-Environmental Restoration and Waste Management Headquarters; with the US Department of Energy-Office of Waste Operations; the US Department of Energy- Headquarters/Vitrification Project Branch; the US Department of Energy-Richland Operations Office/Vitrification Project Office; and the Westinghouse Hanford Company, operations and engineering contractor. This document has been prepared in response to direction from the US Department of Energy-Office of Civilian Radioactive Waste Management through the US Department of Energy-Richland Operations Office for a quality assurance program that meets the requirements of the US Department of Energy. This document provides guidance and direction for implementing a quality assurance program that applies to the Hanford Waste Vitrification Plant Project. The Hanford Waste Vitrification Plant Project management commits to implementing the quality assurance program activities; reviewing the program periodically, and revising it as necessary to keep it current and effective. 12 refs., 6 figs., 1 tab

  9. Status of the public debate on the general options of management of high and intermediate activity and long-lived radioactive wastes

    International Nuclear Information System (INIS)

    2006-01-01

    The law from December 30, 1991, precisely defines 3 axes of researches for the management of high level and long-lived radioactive wastes: separation/transmutation, surface storage and underground disposal. A global evaluation report about these researches is to be supplied in 2006 by the French government to the Parliament. A first synthesis of the knowledge gained after 14 years of research has led the national commission of the public debate (CNDP) to organize a national debate about the general options of management of high-level and long-lived radioactive wastes before the 2006 date line. The debate comprised 4 public hearings (September 2005: Bar-le-Duc, Saint-Dizier, Pont-du-Gard, Cherbourg), 12 round-tables (October and November 2005: Paris, Joinville, Caen, Nancy, Marseille), a synthesis meeting (December 2005, Dunkerque) and a closing meeting (January 2006, Lyon). This document is a status of the general organisation and progress of the debate from its preparing to its end in mid-January 2006. It shows the challenges the CNDP had to take up to treat the scientifical and technical questions in an accessible way, allowing the participation and the hearing of the large public. A status is made of the deception and satisfaction of the public. A lack of confidence in public authorities and scientists has been expressed several times. No pro-nuclear/anti-nuclear shock has occurred and the debate has revealed a remarkable richness in its content. One contribution of the debate to the future project of law is its enlargement to the overall nuclear wastes and valorizable materials and not only to the high/intermediate-level and long-lived wastes. (J.S.)

  10. Neutron Activation analysis of waste water

    International Nuclear Information System (INIS)

    Hernandez H, V.

    1997-01-01

    An instrumental neutron activation analysis for the simultaneous determination of chlorine, bromine, sodium, manganese, cobalt, copper, chromium, zinc, nickel, antimony and iron in waste water is described. They were determined in waste water samples under normal conditions by non-destructive neutron activation simultaneously using a suitable monostandard method. Standardized water samples were used and irradiated in polyethylene ampoules at a neutron flux of 10 13 cm -2 s -1 for periods of 1 minute, 1 and 10 hours. A Ge hyperpure detector was used for your activity determination, with count times of 60, 180, 300 and 600 seconds. The obtained results show than the method can be utilized for the determination of this elements without realize anything previous treatment of the samples. (Author)

  11. Gravitational sedimentation of flocculated waste activated sludge.

    Science.gov (United States)

    Chu, C P; Lee, D J; Tay, J H

    2003-01-01

    The sedimentation characteristics of flocculated wastewater sludge have not been satisfactorily explored using the non-destructive techniques, partially owing to the rather low solid content (ca. 1-2%) commonly noted in the biological sediments. This paper investigated, for the first time, the spatial-temporal gravitational settling characteristics of original and polyelectrolyte flocculated waste activated sludge using Computerized Axial Tomography Scanner. The waste activated sludge possessed a distinct settling characteristic from the kaolin slurries. The waste activated sludges settled more slowly and reached a lower solid fraction in the final sediment than the latter. Flocculation markedly enhanced the settleability of both sludges. Although the maximum achievable solid contents for the kaolin slurries were reduced, flocculation had little effects on the activated sludge. The purely plastic rheological model by Buscall and White (J Chem Soc Faraday Trans 1(83) (1987) 873) interpreted the consolidating sediment data, while the purely elastic model by Tiller and Leu (J. Chin. Inst. Chem. Eng. 11 (1980) 61) described the final equilibrated sediment. Flocculation produced lower yield stress during transient settling, thereby resulting in the more easily consolidated sludge than the original sample. Meanwhile, the flocculated activated sludge was stiffer in the final sediment than in the original sample. The data reported herein are valuable to the theories development for clarifier design and operation.

  12. High-level nuclear waste disposal

    International Nuclear Information System (INIS)

    Burkholder, H.C.

    1985-01-01

    The meeting was timely because many countries had begun their site selection processes and their engineering designs were becoming well-defined. The technology of nuclear waste disposal was maturing, and the institutional issues arising from the implementation of that technology were being confronted. Accordingly, the program was structured to consider both the technical and institutional aspects of the subject. The meeting started with a review of the status of the disposal programs in eight countries and three international nuclear waste management organizations. These invited presentations allowed listeners to understand the similarities and differences among the various national approaches to solving this very international problem. Then seven invited presentations describing nuclear waste disposal from different perspectives were made. These included: legal and judicial, electric utility, state governor, ethical, and technical perspectives. These invited presentations uncovered several issues that may need to be resolved before high-level nuclear wastes can be emplaced in a geologic repository in the United States. Finally, there were sixty-six contributed technical presentations organized in ten sessions around six general topics: site characterization and selection, repository design and in-situ testing, package design and testing, disposal system performance, disposal and storage system cost, and disposal in the overall waste management system context. These contributed presentations provided listeners with the results of recent applied RandD in each of the subject areas

  13. Activation of waste brewer's yeast Saccharomyces cerevisiae for bread production

    Directory of Open Access Journals (Sweden)

    Popov Stevan D.

    2005-01-01

    Full Text Available The waste brewer's yeast S. cerevisiae (activated and non-activated was compared with the commercial baker's yeast regarding the volume of developed gas in dough, volume and freshness stability of produced bread. The activation of waste brewer's yeast resulted in the increased volume of developed gas in dough by 100% compared to non-activated brewer's yeast, and the obtained bread is of more stable freshness compared to bread produced with baker's yeast. The activation of BY affects positively the quality of produced bread regarding bread volume. The volume of developed gas in dough prepared with the use of non-activated BY was not sufficient, therefore, it should not be used as fermentation agent, but only as an additive in bread production process for bread freshness preservation. Intense mixing of dough results in more compressible crumb 48 hrs after baking compared to high-speed mixing.

  14. Advanced High-Level Waste Glass Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Peeler, David K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vienna, John D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schweiger, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fox, Kevin M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-01

    The U.S. Department of Energy Office of River Protection (ORP) has implemented an integrated program to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. The integrated ORP program is focused on providing a technical, science-based foundation from which key decisions can be made regarding the successful operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) facilities. The fundamental data stemming from this program will support development of advanced glass formulations, key process control models, and tactical processing strategies to ensure safe and successful operations for both the low-activity waste (LAW) and high-level waste (HLW) vitrification facilities with an appreciation toward reducing overall mission life. The purpose of this advanced HLW glass research and development plan is to identify the near-, mid-, and longer-term research and development activities required to develop and validate advanced HLW glasses and their associated models to support facility operations at WTP, including both direct feed and full pretreatment flowsheets. This plan also integrates technical support of facility operations and waste qualification activities to show the interdependence of these activities with the advanced waste glass (AWG) program to support the full WTP mission. Figure ES-1 shows these key ORP programmatic activities and their interfaces with both WTP facility operations and qualification needs. The plan is a living document that will be updated to reflect key advancements and mission strategy changes. The research outlined here is motivated by the potential for substantial economic benefits (e.g., significant increases in waste throughput and reductions in glass volumes) that will be realized when advancements in glass formulation continue and models supporting facility operations are implemented. Developing and applying advanced

  15. Coupled processes in NRC high-level waste research

    International Nuclear Information System (INIS)

    Costanzi, F.A.

    1987-01-01

    The author discusses NRC research effort in support of evaluating license applications for disposal of nuclear waste and for promulgating regulations and issuing guidance documents on nuclear waste management. In order to do this they fund research activities at a number of laboratories, academic institutions, and commercial organizations. One of our research efforts is the coupled processes study. This paper discusses interest in coupled processes and describes the target areas of research efforts over the next few years. The specific research activities relate to the performance objectives of NRC's high-level waste (HLW) regulation and the U.S. Environmental Protection Agency (EPA) HLW standard. The general objective of the research program is to ensure the NRC has a sufficient independent technical base to make sound regulatory decisions

  16. New Waste Calciner High Temperature Operation

    International Nuclear Information System (INIS)

    Swenson, M.C.

    2000-01-01

    A new Calciner flowsheet has been developed to process the sodium-bearing waste (SBW) in the INTEC Tank Farm. The new flowsheet increases the normal Calciner operating temperature from 500 C to 600 C. At the elevated temperature, sodium in the waste forms stable aluminates, instead of nitrates that melt at calcining temperatures. From March through May 2000, the new high-temperature flowsheet was tested in the New Waste Calcining Facility (NWCF) Calciner. Specific test criteria for various Calciner systems (feed, fuel, quench, off-gas, etc.) were established to evaluate the long-term operability of the high-temperature flowsheet. This report compares in detail the Calciner process data with the test criteria. The Calciner systems met or exceeded all test criteria. The new flowsheet is a visible, long-term method of calcining SBW. Implementation of the flowsheet will significantly increase the calcining rate of SBW and reduce the amount of calcine produced by reducing the amount of chemical additives to the Calciner. This will help meet the future waste processing milestones and regulatory needs such as emptying the Tank Farm

  17. EPA's high-level waste standards and waste package performance

    International Nuclear Information System (INIS)

    Meyers, S.

    1985-01-01

    The seven assurance requirements EPA was considering were these: (1) Disposal systems shall not depend on active institutional controls for more than 100 years after disposal; (2) Long-term disposal system performance should be monitored for a reasonable time as a supplement to other types of protection; (3) Disposal systems shall be marked and their locations recorded in all appropriate government records; (4) Disposal systems shall be designed with several different types of barriers, both natural and engineered; (5) Sites should not be located where scarce or easily accessible resources are located; (6) Site selection should consider the relative isolation offered by potential alternatives; and (7) Wastes shall be recoverable for a reasonable time after disposal. (orig./PW)

  18. Waste acceptance product specifications for vitrified high-level waste forms

    International Nuclear Information System (INIS)

    Applewhite-Ramsey, A.; Sproull, J.F.

    1993-01-01

    The Nuclear Waste Policy Act of 1982 mandated that all high-level waste (HLW) be sent to a federal geologic repository for permanent disposal. DOE published the Environmental Assessment in 1982 which identified borosilicate glass as the chosen HLW form. 1 In 1985 the Department of Energy instituted a Waste Acceptance Process to assure that DWPF glass waste forms would be acceptable to such a repository. This assurance was important since production of waste forms will precede repository construction and licensing. As part of this Waste Acceptance Process, the DOE Office of Civilian Radioactive Waste Management (RW) formed the Waste Acceptance Committee (WAC). The WAC included representatives from the candidate repository sites, the waste producing sites and DOE. The WAC was responsible for developing the Waste Acceptance Preliminary Specifications (WAPS) which defined the requirements the waste forms must meet to be compatible with the candidate repository geologies

  19. BIOLOGICALLY ACTIVE SUBSTANCES OF SPIRIT PRODUCTION WASTE

    Directory of Open Access Journals (Sweden)

    A. S. Kayshev

    2014-01-01

    Full Text Available A content of biologically active compounds (BAC with signified pharmacological activity in distillers grains was proved. It is prospective for applications of these grains as a raw material resource of pharmaceuticals. A composition of BAC distillers grains received from wheat, corn, barley, millet at different spirit enterprises which use hydro fermentative grain processing. Considering polydispersity of distillers grains they were separated on solid and liquid phases preliminary. Physical and chemical characteristics of distillers grains' liquid base were identified. Elementary composition of distillers grains is signified by active accumulation of biogenic elements (phosphorus, potassium, magnesium, calcium, sodium, iron and low content of heavy metals. The solid phase of distillers grains accumulates carbon, hydrogen and nitrogen in high concentration. The liquid phase of distillers grains contains: proteins and amino acids (20-46%, reducing sugars (5,6%-17,5%, galacturonides (0,8-1,4%, ascorbic acid (6,2-11,4 mg%. The solid base of distillers grains contains: galacturonides (3,4-5,3%, fatty oil (8,4-11,1% with predomination of essential fatty acids, proteins and amino acids (2,1-2,5%, flavonoids (0,4-0,9%, tocopherols (3,4-7,7 mg%. A method of complex processing of distillers grains based on application of membrane filtering of liquid phase and liquid extraction by inorganic and organic solvents of solid phase, which allows almost full extraction of the sum of biologically active compounds (BAC from liquid phase (Biobardin BM and solid phase (Biobardin UL. Biobardin BM comprises the following elements: proteins and amino acids (41-69%, reducing sugars (3,5-15,6%, fatty oil (0,2-0,3%, flavonoids (0,2-0,7%, ascorbic acid (17-37 mg%. Biobardin UL includes: oligouronids (16,4-19,5%, proteins and amino acids (11-21%, fatty oil (3,2-4,9% which includes essential acids; flavonoids (0,6-1,5%, tocopherols (6,6-10,2 mg%, carotinoids (0,13-0,21 mg

  20. Effects of waste content of glass waste forms on Savannah River high-level waste disposal costs

    International Nuclear Information System (INIS)

    McDonell, W.R.; Jantzen, C.M.

    1985-01-01

    Effects of the waste content of glass waste forms of Savannah River high-level waste disposal costs are evaluated by their impact on the number of waste canisters produced. Changes in waste content affect onsite Defense Waste Processing Facility (DWPF) costs as well as offsite shipping and repository emplacement charges. A nominal 1% increase over the 28 wt % waste loading of DWPF glass would reduce disposal costs by about $50 million for Savannah River wastes generated to the year 2000. Waste form modifications under current study include adjustments of glass frit content to compensate for added salt decontamination residues and increased sludge loadings in the DWPF glass. Projected cost reductions demonstrate significant incentives for continued optimization of the glass waste loadings. 13 refs., 3 figs., 3 tabs

  1. Activity measurement algorithm in solid radioactive waste clearance procedure

    International Nuclear Information System (INIS)

    Hinca, R.; Skala, L.

    2014-01-01

    Most of European metrology experts prefer in radiation wastes characterization process to use scaling factor method in combination with high definition gamma spectrometry rather than radionuclide vector method in combination with gross gamma activity measurement. The second method is actually used in Slovakian nuclear facilities. The international (IAEA) and Slovak national authorities (UVZ) recognize both approach but there is a fact, that high definition gamma spectrometry radionuclides more properly. (authors)

  2. Measurement of total alpha activity of neptunium, plutonium, and americium in highly radioactive Hanford waste by iron hydroxide precipitation and 2-heptanone solvent extraction

    International Nuclear Information System (INIS)

    Maiti, T.C.; Kaye, J.H.

    1992-06-01

    An improved method has been developed to concentrate the major alpha-emitting actinide elements neptunium, plutonium, and americium from samples with high salt content such as those resulting from efforts to characterize Hanford storage tank waste. Actinide elements are concentrated by coprecipitation of their hydroxides using iron carrier. The iron is removed by extraction from 8M HCI with 2-heptanone. The actinide elements remain in the aqueous phase free from salts, iron, and long-lived fission products. Recoveries averaged 98 percent

  3. Licensing information needs for a high-level waste repository

    International Nuclear Information System (INIS)

    Wright, R.J.; Greeves, J.T.; Logsdon, M.J.

    1985-01-01

    The information needs for licensing findings during the development of a repository for high-level waste (HLW) are described. In particular, attention is given to the information and needs to demonstrate, for construction authorization purposes: repository constructibility, waste retrievability, waste containment, and waste isolation

  4. High-level waste-form-product performance evaluation

    International Nuclear Information System (INIS)

    Bernadzikowski, T.A.; Allender, J.S.; Stone, J.A.; Gordon, D.E.; Gould, T.H. Jr.; Westberry, C.F. III.

    1982-01-01

    Seven candidate waste forms were evaluated for immobilization and geologic disposal of high-level radioactive wastes. The waste forms were compared on the basis of leach resistance, mechanical stability, and waste loading. All forms performed well at leaching temperatures of 40, 90, and 150 0 C. Ceramic forms ranked highest, followed by glasses, a metal matrix form, and concrete. 11 tables

  5. Alternative High-Performance Ceramic Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Sundaram, S. K. [Alfred Univ., NY (United States)

    2017-02-01

    This final report (M5NU-12-NY-AU # 0202-0410) summarizes the results of the project titled “Alternative High-Performance Ceramic Waste Forms,” funded in FY12 by the Nuclear Energy University Program (NEUP Project # 12-3809) being led by Alfred University in collaboration with Savannah River National Laboratory (SRNL). The overall focus of the project is to advance fundamental understanding of crystalline ceramic waste forms and to demonstrate their viability as alternative waste forms to borosilicate glasses. We processed single- and multiphase hollandite waste forms based on simulated waste streams compositions provided by SRNL based on the advanced fuel cycle initiative (AFCI) aqueous separation process developed in the Fuel Cycle Research and Development (FCR&D). For multiphase simulated waste forms, oxide and carbonate precursors were mixed together via ball milling with deionized water using zirconia media in a polyethylene jar for 2 h. The slurry was dried overnight and then separated from the media. The blended powders were then subjected to melting or spark plasma sintering (SPS) processes. Microstructural evolution and phase assemblages of these samples were studied using x-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersion analysis of x-rays (EDAX), wavelength dispersive spectrometry (WDS), transmission electron spectroscopy (TEM), selective area x-ray diffraction (SAXD), and electron backscatter diffraction (EBSD). These results showed that the processing methods have significant effect on the microstructure and thus the performance of these waste forms. The Ce substitution into zirconolite and pyrochlore materials was investigated using a combination of experimental (in situ XRD and x-ray absorption near edge structure (XANES)) and modeling techniques to study these single phases independently. In zirconolite materials, a transition from the 2M to the 4M polymorph was observed with increasing Ce content. The resulting

  6. Radioactive waste repository of high ecological safety

    International Nuclear Information System (INIS)

    Sobolev, I.; Barinov, A.; Prozorov, L.

    2000-01-01

    With the purpose to construct a radioactive waste repository of high ecological safety and reliable containment, MosNPO 'Radon' specialists have developed an advanced type repository - large diameter well (LBD) one. A project is started for the development of a technology for LDW repository construction and pilot operation of the new repository for 25-30 years. The 2 LDW repositories constructed at the 'Radon' site and the developed monitoring system are described

  7. High-level radioactive wastes. Supplement 1

    Energy Technology Data Exchange (ETDEWEB)

    McLaren, L.H. (ed.)

    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.

  8. Decommissioning high-level waste surface facilities

    International Nuclear Information System (INIS)

    1978-04-01

    The protective storage, entombment and dismantlement options of decommissioning a High-Level Waste Surface Facility (HLWSF) was investigated. A reference conceptual design for the facility was developed based on the designs of similar facilities. State-of-the-art decommissioning technologies were identified. Program plans and cost estimates for decommissioning the reference conceptual designs were developed. Good engineering design concepts were on the basis of this work identified

  9. A truck cask design for shipping defense high-level waste

    International Nuclear Information System (INIS)

    Madsen, M.M.; Zimmer, A.

    1985-01-01

    The Defense High-Level Waste (DHLW) cask is a Type B packaging currently under development by the U.S. Department of Energy (DOE). This truck cask has been designed to initially transport borosilicate glass waste from the Defense Waste Processing Facility (DWPF) to the Waste Isolation Pilot Plant (WIPP). Specific program activities include designing, testing, certifying, and fabricating a prototype legal-weight truck cask system. The design includes such state-of-the-art features as integral impact limiters and remote handling features. A replaceable shielding liner provides the flexibility for shipping a wide range of waste types and activity levels

  10. Update of Nuclear Waste Policy Act transportation activities

    International Nuclear Information System (INIS)

    Callaghan, E.F.

    1987-01-01

    As directed by the Nuclear Waste Policy Act of 1982 (NWPA), the Department of Energy (DOE) is developing a nationwide system for transporting spent nuclear fuel and high-level radioactive waste from commercial power plants to deep geologic repositories for disposal. Plans for the transportation system will consider the following factors: the President's 1985 decision to co-locate some defense high-level waste with commercial waste in a repository, the NWPA requirement that the private sector be used to the fullest extent possible in developing and operating the system, and the possible approval by Congress of the DOE's proposal for a Monitored Retrievable Storage (MRS) facility, submitted in March 1987. (The MRS, if approved, would provide for the consolidation, packaging, and perhaps the temporary storage of spent fuel from reactors.) The ''Transportation Business Plan'', published in January 1986, reflects these considerations. The transportation system, when operational, will consist of two elements: (1) the cask system, which includes the transportation casks, the vehicular conveyances, tie-downs, and associated equipment for handling the casks; and (2) the transportation support system which is comprised of facilities, equipment, and services to support waste transportation. Development of the transportation system incorporates the following work elements: operational planning, support systems development, cash system development, systems analysis, and institutional activities. This paper focusses on the technical aspects of the system

  11. Verification of Vitrified High-Activity Waste Stored in a CASTOR HAW 20/28 CG Cask by Simulated Baseline Comparison

    International Nuclear Information System (INIS)

    Shephard, A.; Arenas-Carrasco, J.; Dratschmidt, H.; De-Baere, P.; Af Ekenstam, G.; Lebrun, A.

    2010-01-01

    The verification process for the vitrification of high-activity waste (HAW) focuses on maintaining the continuity-of-knowledge of special nuclear material (SNM) as it traverses a vitrification facility. However, the inaccessible nature of a vitrification facility presents an obstacle to the deployment of conventional safeguards, albeit the process area of a vitrification facility is effectively a hot cell. The employment of remotely operated NDA hardware/DA sample equipment inside the process area would be problematic-at-best and the alternative of continuous monitoring would draw heavily on the critical resource of inspector time. In response to the aforementioned constraints, the IAEA and Euratom opted to develop a new method which focuses on the verification of SNM after the vitrified HAW has been sealed in storage casks. The new method verifies the presence of the vitrified HAW through the comparison of total neutron count rates collected at points around a cask with those predicted by Monte Carlo simulation. The model includes a dual N50 neutron slab detector (custom design by Euratom) and a CASTOR HAW 20/28 CG storage cask configured with the operator declared contents. By comparison of the simulated neutron emission pattern and field measurements, the displacement of Pu and U is evident from a detectable neutron signal defect. Because the spontaneous fission of 244 Cm is the dominant neutron source in vitrified HAW, the 244 Cm/Pu and 244 Cm/U mass ratios must be known in order to relate the neutron signal outside the cask to the amounts of Pu and U stored inside. These mass ratios can be determined from HAW samples collected by the inspectorates from the accountability tanks and analyzed by DA. The absence of separation of SNM from the HAW is verified by other measures. To ensure the validity of the simulation, sources of uncertainty were systematically addressed and quantified. This new verification method effectively removes the need for NDA equipment

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

    International Nuclear Information System (INIS)

    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. Spent fuel and high-level radioactive waste transportation report

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  16. Permitting plan for the high-level waste interim storage

    International Nuclear Information System (INIS)

    Deffenbaugh, M.L.

    1997-01-01

    This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist

  17. Nuclear waste. DOE's program to prepare high-level radioactive waste for final disposal

    International Nuclear Information System (INIS)

    Bannerman, Carl J.; Owens, Ronald M.; Dowd, Leonard L.; Herndobler, Christopher S.; Purvine, Nancy R.; Stenersen, Stanley G.

    1989-11-01

    years later than the schedule established in early 1984, and the cost could be about $1.1 billion, more than double the 1984 cost estimate. DOE has plans for immobilization facilities at the other two Sites, but unresolved issues could affect the reliability of current cost and schedule estimates; the Hanford facility, currently in the design phase, has an estimated immobilization completion date of 2008, but this date assumes that Hanford's defense mission nuclear processing activities will end in the mid 1990s and only the waste stored in Hanford's double-shell tanks will be immobilized; the INEL facility is currently in such an early planning phase that DOE has not yet selected the waste immobilization technology that it will use. The waste may be transformed into a glass-ceramic or other material instead of being vitrified. DOE expects to make this decision in 1993. Section 1 contains an overview of DOE's high-level waste immobilization program. Sections 2 through 5 contain more detailed information about each of the four projects

  18. Evaluation of treatment alternatives for wastes from both spent fuel rod consolidation and miscellaneous commercial activities

    International Nuclear Information System (INIS)

    Ross, W.A.; Schneider, K.J.; Oma, K.H.; Smith, R.I.; Bunnell, L.R.

    1986-07-01

    Alternative treatments were considered for both existing commercial transuranic wastes and future wastes from spent fuel rod consolidation. Waste treatment was assumed to occur at a hypothetical central treatment facility (a Monitored Retrieval Storage [MRS] facility was used as a reference). Disposal of the waste in a geologic repository was also assumed. The waste form charcteristics, process characteristics, and costs were evaluated for each waste treatment alternative. The evaluation indicated that selection of a high volume reduction alternative can save almost $1 billion in life-cycle costs for the management of transuranic and high-activity wastes from 70,000 MTU of spent fuel compared to the reference MRS waste treatment processes. The supercompaction, arc pyrolysis and melting, and maximum volume reduction alternatives are recommended for further consideration; the latter two are recommended for further testing and demonstration

  19. Fluidized bed system for calcination of high level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Pande, D P; Prasad, T L; Yadgiri, N K; Theyyunni, T K [Process Engineering and Systems Development Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    During the operation of nuclear facilities significant quantities of radiochemical liquid effluents of different concentrations and varying chemical compositions are generated. These effluents contain activated radionuclides, corrosion products and fission products. The advantage of feeding the waste in solid form into the vitrifying equipment are multifold. Efforts are therefore made in many countries to calcine the high level waste, and obtain waste in the oxide form before the same is mixed with glass forming additives and fed into the melter unit. An experimental rig for fluidized bed calcination is constructed for carrying out the detailed investigation of this process, in order to adopt the same for plant scale application. To achieve better gas-solid contact and avoid raining down of solids, a distributor of bubble cap type was designed. A review of existing experience at various laboratories and design of new experimental facility for development of calciners are given. (author). 11 refs., 5 figs.

  20. A comparison of high-level waste form characteristics

    International Nuclear Information System (INIS)

    Salmon, R.; Notz, K.J.

    1991-01-01

    The US DOE is responsible for the eventual disposal in a repository of spent fuels, high-level waste (HLW) and other radioactive wastes that may require long-term isolation. This includes light-water reactor (LWR) spent fuel and immobilized HLW as the two major sources, plus other forms including non-LWR spent fuels and miscellaneous sources (such as activated metals in the Greater-Than-Class-C category). The Characteristics Data Base, sponsored by DOE's Office of Civilian Radioactive Waste Management (OCRWM), was created to systematically tabulate the technical characteristics of these materials. Data are presented here on the immobilized HLW forms that are expected to be produced between now and 2020

  1. Low and medium activity solid wastes processing and encapsulation

    International Nuclear Information System (INIS)

    Taillard, D.; Claes, J.; Hennart, D.

    1983-01-01

    This work, carried out under contract with the European Atomic Energy Community, describes the techniques in use for waste management. The activity of low and medium activity solid wastes is from few curies to few tens of curies per cubic meter, they are produced by nuclear facilities and are often complex mixtures. Radioactive wastes are characterized and processing and conditioning are described. Leaching, stability, mechanical resistance and radiolysis of encapsulated wastes are examined. Handling, storage and disposal are treated

  2. The very-low activity waste storage facility. A new waste management system

    International Nuclear Information System (INIS)

    2006-01-01

    Very-low activity wastes have a radioactivity level close to the natural one. This category of waste is taken into consideration by the French legislation and their storage is one of their point of achievement. This document gives a complete overview of the principles of storage implemented at the storage center for very-low activity wastes (CSTFA) sited in the Aube departement in the vicinity of the storage center for low- and intermediate activity wastes: storage concept, wastes confinement, center organization, environmental monitoring. (J.S.)

  3. National high-level waste systems analysis

    International Nuclear Information System (INIS)

    Kristofferson, K.; O'Holleran, T.P.

    1996-01-01

    Previously, no mechanism existed that provided a systematic, interrelated view or national perspective of all high-level waste treatment and storage systems that the US Department of Energy manages. The impacts of budgetary constraints and repository availability on storage and treatment must be assessed against existing and pending negotiated milestones for their impact on the overall HLW system. This assessment can give DOE a complex-wide view of the availability of waste treatment and help project the time required to prepare HLW for disposal. Facilities, throughputs, schedules, and milestones were modeled to ascertain the treatment and storage systems resource requirements at the Hanford Site, Savannah River Site, Idaho National Engineering Laboratory, and West Valley Demonstration Project. The impacts of various treatment system availabilities on schedule and throughput were compared to repository readiness to determine the prudent application of resources. To assess the various impacts, the model was exercised against a number of plausible scenarios as discussed in this paper

  4. Management of high level radioactive waste

    International Nuclear Information System (INIS)

    Redon, A.; Mamelle, J.; Chambon, M.

    1977-01-01

    The world wide needs in reprocessing will reach the value of 10.000 t/y of irradiated fuels, in the mid of the 80's. Several countries will have planned, in their nuclear programme, the construction of reprocessing plants with a 1500 t/y capacity, corresponding to 50.000 MWe installed. At such a level, the solidification of the radioactive waste will become imperative. For this reason, all efforts, in France, have been directed towards the realization of industrial plants able of solidifying the fission products as a glassy material. The advantages of this decision, and the reasons for it are presented. The continuing development work, and the conditions and methods of storing the high-level wastes prior to solidification, and of the interim storage (for thermal decay) and the ultimate disposal after solidification are described [fr

  5. Testing waste forms containing high radionuclide loadings

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Neilson, R.M. Jr.; Rogers, R.D.

    1986-01-01

    The Low-Level Waste Data Base Development - EPICOR-II Resin/Liner Investigation Program funded by the US Nuclear Regulatory Commission (NRC) is obtaining information on radioactive waste during NRC-prescribed tests and in a disposal environment. This paper describes the resin solidification task of that program, including the present status and results to date. An unusual aspect of this investigation is the use of commercial grade, ion exchange resins that have been loaded with over five times the radioactivity normally seen in a commercial application. That dramatically increases the total radiation dose to the resins. The objective of the resin solidification task is to determine the adequacy of test procedures specified by NRC for ion exchange resins having high radionuclide loadings

  6. The Radioactive Waste Management course: 14 High-yield editions

    International Nuclear Information System (INIS)

    Alonso, A.; Gallego, E.; Marco, M. L.; Falcon, S.

    2003-01-01

    The doctorate course on Radioactive Waste Management was initiated in February 1988, by initiative of the Chair of Nuclear Technology, under the sponsoring of the national radioactive waste management company (ENRESA), in a fruitful collaboration between the Institute Artigas of the Technical School of Industrial Engineering and the Institute of Formation on Energy of the research centre CIEMAT. The course is also offered as a post-graduate through both institutes. After completion of fourteen consecutive editions in 2002, the course constituted a landmark in the field of nuclear education in Spain. The last edition offered, along 35 lessons published in two books, the general aspects of generation, treatment and conditioning of radioactive wastes, the basic Safety and Radiological Protection criteria, the detailed technical questions of the management of both low-intermediate.activity and the high-activity level, together with the wastes generated during decommissioning and dismantling of installations, as well as the general and institutional aspects. Experts in each field, belonging either to ENRESA, CIEMAT, the Nuclear Safety Council, the UPM and the industry, present such wide programme. A technical visit to the low and intermediate radioactive waste repository of El Cabril was also offered to the participants as part of the course, as in previous years the visit to the dismantling workers of Vandellos I NPP. More than 500 engineers and graduates in different science branches have participated in the course along 14 years, with both students and professionals belonging to ENRESA, the Nuclear Safety Council, CIEMAT and other research centers, hospitals, civil protection at different levels, service and engineering companies related with the radioactive waste management. Altogether, it is possible to say, as the title is expressed, that the course has given in these 14 years a high-production yield. (Author)

  7. Repackaging of High Fissile TRU Waste at the Transuranic Waste Processing Center - 13240

    Energy Technology Data Exchange (ETDEWEB)

    Oakley, Brian; Heacker, Fred [WAI, TRU Waste Processing Center, 100 WIPP Road Lenoir City, TN 37771 (United States); McMillan, Bill [DOE, Oak Ridge Operations, Bldg. 2714, Oak Ridge, TN 37830 (United States)

    2013-07-01

    Twenty-six drums of high fissile transuranic (TRU) waste from Oak Ridge National Laboratory (ORNL) operations were declared waste in the mid-1980's and placed in storage with the legacy TRU waste inventory for future treatment and disposal at the Waste Isolation Pilot Plant (WIPP). Repackaging and treatment of the waste at the TRU Waste Packaging Center (TWPC) will require the installation of additional equipment and capabilities to address the hazards for handling and repackaging the waste compared to typical Contact Handled (CH) TRU waste that is processed at the TWPC, including potential hydrogen accumulation in legacy 6M/2R packaging configurations, potential presence of reactive plutonium hydrides, and significant low energy gamma radiation dose rates. All of the waste is anticipated to be repackaged at the TWPC and certified for disposal at WIPP. The waste is currently packaged in multiple layers of containers which presents additional challenges for repackaging activities due to the potential for the accumulation of hydrogen gas in the container headspace in quantities than could exceed the Lower Flammability Limit (LFL). The outer container for each waste package is a stainless steel 0.21 m{sup 3} (55-gal) drum which contains either a 0.04 m{sup 3} or 0.06 m{sup 3} (10-gal or 15-gal) 6M drum. The inner 2R container in each 6M drum is ∼12 cm (5 in) outside diameter x 30-36 cm (12-14 in) long and is considered to be a > 4 liter sealed container relative to TRU waste packaging criteria. Inside the 2R containers are multiple configurations of food pack cans, pipe nipples, and welded capsules. The waste contains significant quantities of high burn-up plutonium oxides and metals with a heavy weight percentage of higher atomic mass isotopes and the subsequent in-growth of significant quantities of americium. Significant low energy gamma radiation is expected to be present due to the americium in-growth. Radiation dose rates on inner containers are estimated

  8. Regulation of higher-activity NARM wastes by EPA

    International Nuclear Information System (INIS)

    Bandrowski, M.S.

    1988-01-01

    The US Environmental Protection Agency (EPA) is currently developing standards for the disposal of low-level radioactive waste (LLW). As part of this Standard, EPA is including regulations for the disposal of naturally occurring and accelerator-produced radioactive material (NARM) wastes not covered under the Atomic Energy Act (AEA). The regulations will cover only higher-activity NARM wastes, defined as NARM waste with specific activity exceeding two nanocuries per gram. The proposed regulations will specify that NARM wastes exceeding the above limits, except for specific exempted items, must be disposed of in regulated radioactive waste disposal facilities. The proposed EPA regulations for NARM wastes will be discussed, as well as the costs and benefits of the regulation, how it will be implemented by EPA, and the rationale for covering only higher-activity NARM wastes exceeding two nanocuries per gram

  9. The Japan Power Demonstration Reactor (JPDR) dismantling activities. Management of JPDR dismantling waste

    International Nuclear Information System (INIS)

    Abe, Masayoshi; Nakata, Susumu; Ito, Shinichi

    1996-01-01

    The management of wastes, both radioactive and non-radioactive, is one of the most important issues for a safe and reasonable dismantling operation of nuclear power plants. A large amount of radioactive wastes is arising from a reactor dismantling operation in a relatively short period time, ranging in a wide variety from very low level to relatively high level. Moreover non-radioactive waste is also in a huge amount. The dismantling operation of Japan Power Demonstration Reactor (JPDR) resulted in 24,440 tons of dismantling wastes, of which about 15% was radioactive and 85% non-radioactive. These wastes were managed successfully implementing a well developed management plan for JPDR dismantling waste. Research and development works for handling of JPDR dismantling wastes were performed, including fixation of loose contamination on surface, volume reduction and waste containers for on-site transportation and interim storage. The JPDR dismantling wastes generated were classified and categorized depending on their materials, characteristics and activity level. Approximately 2,100 tons of radioactive wastes were stored in the interim storage facilities on site using developed containers, and 1,670 tons of radioactive concrete waste were used for a safe demonstration test of a simple near-surface disposal for very low level waste. Other dismantling wastes such as steel and concrete which were categorized as non-radioactive were recycled and reused as useful resources. This paper describes the management of the JPDR dismantling wastes. (author)

  10. Characterization of low and medium active wastes

    International Nuclear Information System (INIS)

    Saas, A.

    1993-01-01

    For several years now, research on raw or packaged waste characterization has been carried out in France. Qualitative or quantitative analysis are given of radionuclides present in already packaged waste (including badly packaged waste) or in unpackaged waste; as far as possible, evaluation of the main physico-mechanical and confinement characteristics

  11. Vitrification of high-level liquid wastes

    International Nuclear Information System (INIS)

    Varani, J.L.; Petraitis, E.J.; Vazquez, Antonio.

    1987-01-01

    High-level radioactive liquid wastes produced in the fuel elements reprocessing require, for their disposal, a preliminary treatment by which, through a series of engineering barriers, the dispersion into the biosphere is delayed by 10 000 years. Four groups of compounds are distinguished among a great variety of final products and methods of elaboration. From these, the borosilicate glasses were chosen. Vitrification experiences were made at a laboratory scale with simulated radioactive wastes, employing different compositions of borosilicate glass. The installations are described. A series of tests were carried out on four basic formulae using always the same methodology, consisting of a dry mixture of the vitreous matrix's products and a dry simulated mixture. Several quality tests of the glasses were made 1: Behaviour in leaching following the DIN 12 111 standard; 2: Mechanical resistance; parameters related with the facility of the different glasses for increasing their surface were studied; 3: Degree of devitrification: it is shown that devitrification turns the glasses containing radioactive wastes easily leachable. From all the glasses tested, the composition SiO 2 , Al 2 O 3 , B 2 O 3 , Na 2 O, CaO shows the best retention characteristics. (M.E.L.) [es

  12. Ocean disposal of high level radioactive waste

    International Nuclear Information System (INIS)

    1983-01-01

    This study confirms, subject to limitations of current knowledge, the engineering feasibility of free fall penetrators for High Level Radioactive Waste disposal in deep ocean seabed sediments. Restricted sediment property information is presently the principal bar to an unqualified statement of feasibility. A 10m minimum embedment and a 500 year engineered barrier waste containment life are identified as appropriate basic penetrator design criteria at this stage. A range of designs are considered in which the length, weight and cross section of the penetrator are varied. Penetrators from 3m to 20m long and 2t to 100t in weight constructed of material types and thicknesses to give a 500 year containment life are evaluated. The report concludes that the greatest degree of confidence is associated with performance predictions for 75 to 200 mm thick soft iron and welded joints. A range of lengths and capacities from a 3m long single waste canister penetrator to a 20m long 12 canister design are identified as meriting further study. Estimated embedment depths for this range of penetrator designs lie between 12m and 90m. Alternative manufacture, transport and launch operations are assessed and recommendations are made. (author)

  13. Radioactive Waste Management Program Activities in Croatia

    International Nuclear Information System (INIS)

    Matanic, R.

    2000-01-01

    The concept of radioactive waste management in Croatia comprises three major areas: management of low and intermediate level radioactive waste (LILRW), spent fuel management and decommissioning. All the work regarding radioactive waste management program is coordinated by Hazardous Waste Management Agency (APO) and Croatian Power Utility (HEP) in cooperation with other relevant institutions. Since the majority of work has been done in developing low and intermediate level radioactive waste management program, the paper will focus on this part of radioactive waste management, mainly on issues of site selection and characterization, repository design, safety assessment and public acceptance. A short description of national radioactive waste management infrastructure will also be presented. (author)

  14. HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASSES FOR HANFORD'S WTP (WASTE TREATMENT PROJECT)

    International Nuclear Information System (INIS)

    Kruger, A.A.; Bowan, B.W.; Joseph, I.; Gan, H.; Kot, W.K.; Matlack, K.S.; Pegg, I.L.

    2010-01-01

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

  15. Waste minimization/pollution prevention study of high-priority waste streams

    International Nuclear Information System (INIS)

    Ogle, R.B.

    1994-03-01

    Although waste minimization has been practiced by the Metals and Ceramics (M ampersand C) Division in the past, the effort has not been uniform or formalized. To establish the groundwork for continuous improvement, the Division Director initiated a more formalized waste minimization and pollution prevention program. Formalization of the division's pollution prevention efforts in fiscal year (FY) 1993 was initiated by a more concerted effort to determine the status of waste generation from division activities. The goal for this effort was to reduce or minimize the wastes identified as having the greatest impact on human health, the environment, and costs. Two broad categories of division wastes were identified as solid/liquid wastes and those relating to energy use (primarily electricity and steam). This report presents information on the nonradioactive solid and liquid wastes generated by division activities. More specifically, the information presented was generated by teams of M ampersand C staff members empowered by the Division Director to study specific waste streams

  16. Status of the French nuclear high level waste disposal

    International Nuclear Information System (INIS)

    Sombret, C.

    1985-09-01

    French research on high level waste processing has led to the development of industrial vitrification facilities. Borosilicate glass is still being investigated for its long-term storage properties, since it is itself a component of the containment system. The other constituents of this system, the engineered barriers, are also being actively investigated. The geological barrier is now being assessed using a methodology applicable to various types of geological formations, and final site qualification should be possible before the end of 1992

  17. High-level waste processing and conditioning: vitrification

    International Nuclear Information System (INIS)

    Bonniaud, R.

    1981-02-01

    The vitrification process used to treat fission product solutions at the Marcoule Vitrification Plant is described. The type of waste processed is characterized by its very high activity and the long lifetimes of some of the emitters that it contains. The performance obtained with this process is given together with the future developments envisaged. The storage of glasses is described as well as their behavior with time [fr

  18. Glasses used for the high level radioactive wastes storage

    International Nuclear Information System (INIS)

    Sombret, C.

    1983-06-01

    High level radioactive wastes generated by the reprocessing of spent fuels is an important concern in the conditioning of radioactive wastes. This paper deals with the status of the knowledge about glasses used for the treatment of these liquids [fr

  19. Preparation of plutonium waste forms with ICPP calcined high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Staples, B.A.; Knecht, D.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); O`Holleran, T.P. [Argonne National Lab.-West, Idaho Falls, ID (United States)] [and others

    1997-05-01

    Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce{sup +4}) as a surrogate for plutonium (Pu{sup +4}) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study.

  20. Preparation of plutonium waste forms with ICPP calcined high-level waste

    International Nuclear Information System (INIS)

    Staples, B.A.; Knecht, D.A.; O'Holleran, T.P.

    1997-05-01

    Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce +4 ) as a surrogate for plutonium (Pu +4 ) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study

  1. Description of a ceramic waste form and canister for Savannah River Plant high-level waste

    International Nuclear Information System (INIS)

    Butler, J.L.; Allender, J.S.; Gould, T.H. Jr.

    1982-04-01

    A canistered ceramic waste form for possible immobilization of Savannah River Plant (SRP) high-level radioactive wastes is described. Characteristics reported for the form include waste loading, chemical composition, heat content, isotope inventory, mechanical and thermal properties, and leach rates. A conceptual design of a potential production process for making this canistered form are also described. The ceramic form was selected in November 1981 as the primary alternative to the reference waste form, borosilicate glass, for making a final waste form decision for SRP waste by FY-1983. 11 tables

  2. Technical basis for classification of low-activity waste fraction from Hanford site tanks

    Energy Technology Data Exchange (ETDEWEB)

    Petersen, C.A., Westinghouse Hanford

    1996-07-17

    The overall objective of this report is to provide a technical basis to support a U.S. Nuclear Regulatory Commission determination to classify the low-activity waste from the Hanford Site single-shell and double-shell tanks as `incidental` wastes after removal of additional radionuclides and immobilization.The proposed processing method, in addition to the previous radionuclide removal efforts, will remove the largest practical amount of total site radioactivity, attributable to high-level wastes, for disposal in a deep geologic repository. The remainder of the waste would be considered `incidental` waste and could be disposed onsite.

  3. Technical basis for classification of low-activity waste fraction from Hanford site tanks

    Energy Technology Data Exchange (ETDEWEB)

    Petersen, C.A.

    1996-09-20

    The overall objective of this report is to provide a technical basis to support a U.S. Nuclear Regulatory Commission determination to classify the low-activity waste from the Hanford Site single-shell and double-shell tanks as `incidental` wastes after removal of additional radionuclides and immobilization.The proposed processing method, in addition to the previous radionuclide removal efforts, will remove the largest practical amount of total site radioactivity, attributable to high-level waste, for disposal is a deep geologic repository. The remainder of the waste would be considered `incidental` waste and could be disposed onsite.

  4. The development of a strategy for the management of high level radioactive wastes

    International Nuclear Information System (INIS)

    Beale, H.

    1981-07-01

    An assessment is made of the options available for the management of high level radioactive wastes. This preliminary study leads to the conclusion that the high active liquor should be vitrified at the earliest possible date and points to the advantages of storing the vitrified waste for an extended period in reinforced concrete casks. (author)

  5. Quarterly Briefing Book on Environmental and Waste Management Activities

    International Nuclear Information System (INIS)

    Brown, M.C.

    1991-06-01

    The purpose of the Quarterly Briefing Book on Environmental and Waste Management Activities is to provide managers and senior staff at the US Department of Energy-Richland Operations Office and its contractors with timely and concise information on Hanford Site environmental and waste management activities. Each edition updates the information on the topics in the previous edition, deletes those determined not to be of current interest, and adds new topics to keep up to date with changing environmental and waste management requirements and issues. Section A covers current waste management and environmental restoration issues. In Section B are writeups on national or site-wide environmental and waste management topics. Section C has writeups on program- and waste-specific environmental and waste management topics. Section D provides information on waste sites and inventories on the site. 15 figs., 4 tabs

  6. Quarterly Briefing Book on Environmental and Waste Management Activities

    Energy Technology Data Exchange (ETDEWEB)

    Brown, M.C.

    1991-06-01

    The purpose of the Quarterly Briefing Book on Environmental and Waste Management Activities is to provide managers and senior staff at the US Department of Energy-Richland Operations Office and its contractors with timely and concise information on Hanford Site environmental and waste management activities. Each edition updates the information on the topics in the previous edition, deletes those determined not to be of current interest, and adds new topics to keep up to date with changing environmental and waste management requirements and issues. Section A covers current waste management and environmental restoration issues. In Section B are writeups on national or site-wide environmental and waste management topics. Section C has writeups on program- and waste-specific environmental and waste management topics. Section D provides information on waste sites and inventories on the site. 15 figs., 4 tabs.

  7. Stabilization of high and low solids Consolidated Incinerator Facility (CIF) waste with super cement

    International Nuclear Information System (INIS)

    Walker, B.W.

    2000-01-01

    This report details solidification activities using selected Mixed Waste Focus Area technologies with the High and Low Solid waste streams. Ceramicrete and Super Cement technologies were chosen as the best possible replacement solidification candidates for the waste streams generated by the SRS incinerator from a list of several suggested Mixed Waste Focus Area technologies. These technologies were tested, evaluated, and compared to the current Portland cement technology being employed. Recommendation of a technology for replacement depends on waste form performance, process flexibility, process complexity, and cost of equipment and/or raw materials

  8. DEFENSE HIGH LEVEL WASTE GLASS DEGRADATION

    International Nuclear Information System (INIS)

    Ebert, W.

    2001-01-01

    The purpose of this Analysis/Model Report (AMR) is to document the analyses that were done to develop models for radionuclide release from high-level waste (HLW) glass dissolution that can be integrated into performance assessment (PA) calculations conducted to support site recommendation and license application for the Yucca Mountain site. This report was developed in accordance with the ''Technical Work Plan for Waste Form Degradation Process Model Report for SR'' (CRWMS M andO 2000a). It specifically addresses the item, ''Defense High Level Waste Glass Degradation'', of the product technical work plan. The AP-3.15Q Attachment 1 screening criteria determines the importance for its intended use of the HLW glass model derived herein to be in the category ''Other Factors for the Postclosure Safety Case-Waste Form Performance'', and thus indicates that this factor does not contribute significantly to the postclosure safety strategy. Because the release of radionuclides from the glass will depend on the prior dissolution of the glass, the dissolution rate of the glass imposes an upper bound on the radionuclide release rate. The approach taken to provide a bound for the radionuclide release is to develop models that can be used to calculate the dissolution rate of waste glass when contacted by water in the disposal site. The release rate of a particular radionuclide can then be calculated by multiplying the glass dissolution rate by the mass fraction of that radionuclide in the glass and by the surface area of glass contacted by water. The scope includes consideration of the three modes by which water may contact waste glass in the disposal system: contact by humid air, dripping water, and immersion. The models for glass dissolution under these contact modes are all based on the rate expression for aqueous dissolution of borosilicate glasses. The mechanism and rate expression for aqueous dissolution are adequately understood; the analyses in this AMR were conducted to

  9. Sustainable Development and High Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Mikael [Swedish Radiation Protection Inst., Stockholm (Sweden)

    2001-07-01

    Sustainable development, defined by the BrundtIand Commission as 'development that meets the needs of the present without compromising the ability of future generations to meet their own needs', relates to a number of issues such as population, health, food, species and ecosystems, energy, industrial development, urbanization, societal issues and economy, and how these global challenges could be met within a long term strategy. It is not obvious how the principle may be applied to final disposal of radioactive waste, but the global scope of the principle suggests that no sector in society should be exempted from scrutinizing its practices in the light of the challenge presented by sustainable development. Waste management, as pointed out by the International Commission on Radiological Protection, ICRP, cannot be seen as a free standing practice in need of its own justification. The produced waste cannot be seen separately from the other components of nuclear production. However, the existence of very long-lived radioactive nuclei in the spent fuel warrants a careful examination of this subpractice. Health based post-closure criteria or standards for long-lived waste, usually make use of the concept of partitioning dose limit. ICRP recommends that individuals in the public do not receive a yearly dose in excess of 1 mSv as a result of releases in connection with activities involving the use of ionising radiation, and that any single facility does not generate a dose burden to individuals in excess of a fraction of this value. For an operating facility, this fraction is normally at least a factor of three. By definition, operational changes are not possible for a closed repository. It follows from this that the partitioning has another function. One interpretation is that it can allow for the simultaneous use and burdens of future generation's activities. Both the Swedish and the proposed US criteria and from EPA and NRC, as well as standards from

  10. Sustainable Development and High Level Waste

    International Nuclear Information System (INIS)

    Jensen, Mikael

    2001-01-01

    Sustainable development, defined by the BrundtIand Commission as 'development that meets the needs of the present without compromising the ability of future generations to meet their own needs', relates to a number of issues such as population, health, food, species and ecosystems, energy, industrial development, urbanization, societal issues and economy, and how these global challenges could be met within a long term strategy. It is not obvious how the principle may be applied to final disposal of radioactive waste, but the global scope of the principle suggests that no sector in society should be exempted from scrutinizing its practices in the light of the challenge presented by sustainable development. Waste management, as pointed out by the International Commission on Radiological Protection, ICRP, cannot be seen as a free standing practice in need of its own justification. The produced waste cannot be seen separately from the other components of nuclear production. However, the existence of very long-lived radioactive nuclei in the spent fuel warrants a careful examination of this subpractice. Health based post-closure criteria or standards for long-lived waste, usually make use of the concept of partitioning dose limit. ICRP recommends that individuals in the public do not receive a yearly dose in excess of 1 mSv as a result of releases in connection with activities involving the use of ionising radiation, and that any single facility does not generate a dose burden to individuals in excess of a fraction of this value. For an operating facility, this fraction is normally at least a factor of three. By definition, operational changes are not possible for a closed repository. It follows from this that the partitioning has another function. One interpretation is that it can allow for the simultaneous use and burdens of future generation's activities. Both the Swedish and the proposed US criteria and from EPA and NRC, as well as standards from Canada, UK and

  11. High-level radioactive waste disposal problem in Russia

    International Nuclear Information System (INIS)

    Velichkin, Vasily I.

    1999-01-01

    This presentation on radioactive waste management in Russia discusses criteria for the selection of disposal sites, how the various types of waste should be contained and stored, and gives a list showing the liable owner, type, volume, activity and storage place of the present amount of radioactive waste. The bulk of this waste, in volume and radioactivity, is at the enterprises of Minatom of the Russian Federation

  12. Glassy slags as novel waste forms for remediating mixed wastes with high metal contents

    International Nuclear Information System (INIS)

    Feng, X.; Wronkiewicz, D.J.; Bates, J.K.; Brown, N.R.; Buck, E.C.; Gong, M.; Ebert, W.L.

    1994-01-01

    Argonne National Laboratory (ANL) is developing a glassy slag final waste form for the remediation of low-level radioactive and mixed wastes with high metal contents. This waste form is composed of various crystalline and metal oxide phases embedded in a silicate glass phase. This work indicates that glassy slag shows promise as final waste form because (1) it has similar or better chemical durability than high-level nuclear waste (HLW) glasses, (2) it can incorporate large amounts of metal wastes, (3) it can incorporate waste streams having low contents of flux components (boron and alkalis), (4) it has less stringent processing requirements (e.g., viscosity and electric conductivity) than glass waste forms, (5) its production can require little or no purchased additives, which can result in greater reduction in waste volume and overall treatment costs. By using glassy slag waste forms, minimum additive waste stabilization approach can be applied to a much wider range of waste streams than those amenable only to glass waste forms

  13. Treatment of High-Level Waste Arising from Pyrochemical Processes

    International Nuclear Information System (INIS)

    Lizin, A.A.; Kormilitsyn, M.V.; Osipenko, A.G.; Tomilin, S.V.; Lavrinovich, Yu.G.

    2013-01-01

    JSC “SSC RIAR” has been performing research and development activities in support of closed fuel cycle of fast reactor since the middle of 1960s. Fuel cycle involves fabrication and reprocessing of spent nuclear fuel (SNF) using pyrochemical methods of reprocessing in molten alkali metal chlorides. At present pyrochemical methods of SNF reprocessing in molten chlorides has reached such a level in their development that makes it possible to compare their competitiveness with classic aqueous methods. Their comparative advantage lies in high safety, compactness, high protectability as to nonproliferation of nuclear materials, and reduction of high level waste volume

  14. Handbook of high-level radioactive waste transportation

    International Nuclear Information System (INIS)

    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

  15. The immobilization of High Level Waste Into Glass

    International Nuclear Information System (INIS)

    Aisyah; Martono, H.

    1998-01-01

    High level liquid waste is generated from the first step extraction in the nuclear fuel reprocessing. The waste is immobilized with boro-silicate glass. A certain composition of glass is needed for a certain type of waste, so that the properties of waste glass would meet the requirement either for further process or for disposal. The effect of waste loading on either density, thermal expansion, softening point and leaching rate has been studied. The composition of the high level liquid waste has been determined by ORIGEN 2 and the result has been used to prepare simulated high level waste. The waste loading in the waste glass has been set to be 19.48; 22.32; 25.27; and 26.59 weight percent. The result shows that increasing the waste loading has resulted in the higher density with no thermal expansion and softening point significant change. The increase in the waste loading increase that leaching rate. The properties of the waste glass in this research have not shown any deviation from the standard waste glass properties

  16. Processing vessel for high level radioactive wastes

    International Nuclear Information System (INIS)

    Maekawa, Hiromichi

    1998-01-01

    Upon transferring an overpack having canisters containing high level radioactive wastes sealed therein and burying it into an underground processing hole, an outer shell vessel comprising a steel plate to be fit and contained in the processing hole is formed. A bury-back layer made of dug earth and sand which had been discharged upon forming the processing hole is formed on the inner circumferential wall of the outer shell vessel. A buffer layer having a predetermined thickness is formed on the inner side of the bury-back layer, and the overpack is contained in the hollow portion surrounded by the layer. The opened upper portion of the hollow portion is covered with the buffer layer and the bury-back layer. Since the processing vessel having a shielding performance previously formed on the ground, the state of packing can be observed. In addition, since an operator can directly operates upon transportation and burying of the high level radioactive wastes, remote control is no more necessary. (T.M.)

  17. Aube very low activity waste storage Centre. Annual report 2009

    International Nuclear Information System (INIS)

    2010-01-01

    After a presentation of the ANDRA (the French national agency for radioactive waste management), its role and missions, its sites, its strategy with respect to a sustainable development, this report contains a description of waste storage installations and key figures of the activity in 2009 (origin and nature of very low activity wastes, brief description of the Aube centre installations, stored volumes, performed works). It describes arrangements related to security, safety and radioprotection, presents results of the radiological survey activity performed in the environment and on wastes, and activities related to public information

  18. High-level waste immobilization program: an overview

    International Nuclear Information System (INIS)

    Bonner, W.R.

    1979-09-01

    The High-Level Waste Immobilization Program is providing technology to allow safe, affordable immobilization and disposal of nuclear waste. Waste forms and processes are being developed on a schedule consistent with national needs for immobilization of high-level wastes stored at Savannah River, Hanford, Idaho National Engineering Laboratory, and West Valley, New York. This technology is directly applicable to high-level wastes from potential reprocessing of spent nuclear fuel. The program is removing one more obstacle previously seen as a potential restriction on the use and further development of nuclear power, and is thus meeting a critical technological need within the national objective of energy independence

  19. Urine: Waste product or biologically active tissue?

    Science.gov (United States)

    2018-03-01

    Historically, urine has been viewed primarily as a waste product with little biological role in the overall health of an individual. Increasingly, data suggest that urine plays a role in human health beyond waste excretion. For example, urine might act as an irritant and contribute to symptoms through interaction with-and potential compromise of-the urothelium. To explore the concept that urine may be a vehicle for agents with potential or occult bioactivity and to discuss existing evidence and novel research questions that may yield insight into such a role, the National Institute of Diabetes and Digestive and Kidney Disease invited experts in the fields of comparative evolutionary physiology, basic science, nephrology, urology, pediatrics, metabolomics, and proteomics (among others) to a Urinology Think Tank meeting on February 9, 2015. This report reflects ideas that evolved from this meeting and current literature, including the concept of urine quality, the biological, chemical, and physical characteristics of urine, including the microbiota, cells, exosomes, pH, metabolites, proteins, and specific gravity (among others). Additionally, the manuscript presents speculative, and hopefully testable, ideas about the functional roles of urine constituents in health and disease. Moving forward, there are several questions that need further understanding and pursuit. There were suggestions to consider actively using various animal models and their biological specimens to elaborate on basic mechanistic information regarding human bladder dysfunction. Published 2018. This article is a U.S. Government work and is in the public domain in the USA.

  20. Crystalline matter for solidification of highly radioactive wastes

    International Nuclear Information System (INIS)

    Grauer, R.

    1984-02-01

    Highly active wastes from reprocessed nuclear fuels must be incorporated into a solid chemically resistant inorganic matrix prior to final storage. One possible alternative to glassification is to embed the complex oxide mixture in a crystalline ceramic. A discussion from the structural and chemical viewpoint is presented giving guidelines for the selection and development of such a product. The chemical and phase composition concerning the most important developments are described. SYNROC is the most highly developed solid ceramic that has been evaluated to date for power reactor wastes. However, its testing and development so far has been restricted to simulated inactive materials. One of the most important aspects of solid high activity wastes is their behaviour in water. SYNROC reacts more slowly than glasses with water at temperatures over 100 0 C. Its low release of actinides under these conditions is remarkable. At temperatures under 100 0 C the important nuclide Cs 137 is released from SYNROC and from glasses at comparable rates. These assertions concerning chemical stability are however based on short term experiments, which have not considered the possibly complex interactions occurring during final storage. The information is therefore insufficient to describe the basic model required to predict long term behaviour under final storage conditions. Finally the report makes recommendations for a further programme of work. (Auth.)

  1. Highly Efficient Fecal Waste Incinerator, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Volume reduction is a critical element of Solid Waste Management for manned spacecraft and planetary habitations. To this end, the proposed fecal waste incinerator...

  2. Talc-silicon glass-ceramic waste forms for immobilization of high- level calcined waste

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1993-06-01

    Talc-silicon glass-ceramic waste forms are being evaluated as candidates for immobilization of the high level calcined waste stored onsite at the Idaho Chemical Processing Plant. These glass-ceramic waste forms were prepared by hot isostatically pressing a mixture of simulated nonradioactive high level calcined waste, talc, silicon and aluminum metal additives. The waste forms were characterized for density, chemical durability, and glass and crystalline phase compositions. The results indicate improved density and chemical durability as the silicon content is increased

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

    International Nuclear Information System (INIS)

    1982-03-01

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

  4. High-level radioactive waste glass and storage canister design

    International Nuclear Information System (INIS)

    Slate, S.C.; Ross, W.A.

    1979-01-01

    Management of high-level radioactive wastes is a primary concern in nuclear operations today. The main objective in managing these wastes is to convert them into a solid, durable form which is then isolated from man. A description is given of the design and evaluation of this waste form. The waste form has two main components: the solidified waste and the storage canister. The solid waste form discussed in this study is glass. Waste glasses have been designed to be inert to water attack, physically rugged, low in volatility, and stable over time. Two glass-making processes are under development at PNL. The storage canister is being designed to provide high-integrity containment for solidified wastes from processing to terminal storage. An outline is given of the steps in canister design: material selection, stress and thermal analyses, quality verification, and postfill processing. Examples are given of results obtained from actual nonradioactive demonstration tests. 14 refs

  5. High-Level waste process and product data annotated bibliography

    International Nuclear Information System (INIS)

    Stegen, G.E.

    1996-01-01

    The objective of this document is to provide information on available issued documents that will assist interested parties in finding available data on high-level waste and transuranic waste feed compositions, properties, behavior in candidate processing operations, and behavior on candidate product glasses made from those wastes. This initial compilation is only a partial list of available references

  6. High level waste canister emplacement and retrieval concepts study

    International Nuclear Information System (INIS)

    1975-09-01

    Several concepts are described for the interim (20 to 30 years) storage of canisters containing high level waste, cladding waste, and intermediate level-TRU wastes. It includes requirements, ground rules and assumptions for the entire storage pilot plant. Concepts are generally evaluated and the most promising are selected for additional work. Follow-on recommendations are made

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

    Science.gov (United States)

    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.

  8. Waste management activities and carbon emissions in Africa

    International Nuclear Information System (INIS)

    Couth, R.; Trois, C.

    2011-01-01

    This paper summarizes research into waste management activities and carbon emissions from territories in sub-Saharan Africa with the main objective of quantifying emission reductions (ERs) that can be gained through viable improvements to waste management in Africa. It demonstrates that data on waste and carbon emissions is poor and generally inadequate for prediction models. The paper shows that the amount of waste produced and its composition are linked to national Gross Domestic Product (GDP). Waste production per person is around half that in developed countries with a mean around 230 kg/hd/yr. Sub-Saharan territories produce waste with a biogenic carbon content of around 56% (+/-25%), which is approximately 40% greater than developed countries. This waste is disposed in uncontrolled dumps that produce large amounts of methane gas. Greenhouse gas (GHG) emissions from waste will rise with increasing urbanization and can only be controlled through funding mechanisms from developed countries.

  9. Removal of actinides from high activity wastes by solvent extraction: outline of the research work at Ispra J.R.C. laboratories

    International Nuclear Information System (INIS)

    Mannone, F.

    1976-07-01

    The development of an advanced waste management alternative such as the actinide nuclear incineration requires an almost quantitative removal of actinides from waste streams. Within the framework of the Ispra JRC Waste Disposal R and D programme, actinide separation studies were directed towards solvent extraction and precipitation methods. To develop a tentative waste partitioning flow-sheet based on solvent extraction, two conceptual process flow-sheet for actinide removal were evaluated on the basis of the currently used actinide recovery processes, i.e. removal after waste adjustment to low-acidity conditions and direct actinide removal from acidic wastes, as they are generated in actual reprocessing plants. No improvements have been devised for actinide recoveries within the conventional Purex reprocessing operations and a currently agreed value has been assumed for neptunium recovery (90%). According to these basic orientations some organic extractants have been selected for testing as promising candidates for waste partitioning and laboratory studies, designed to develop a satisfactory partitioning flow-sheet, have been proposed and described

  10. An overview of the AECB's strategy for regulating radioactive waste management activities

    International Nuclear Information System (INIS)

    Hamel, P.E.; Smythe, W.D.; Duncan, R.M.; Coady, J.R.

    1982-07-01

    The goal of the Canadian Atomic Energy Control Board in regulating the management of radioactive wastes is to ensure the protection of people and the environment. A program of cooperation with other agencies, identification and adoption of baselines for describing radioactive wastes, development of explicit criteria and requirements, publication of related regulatory documents, establishment of independent consultative processes with technical experts and the public, and maintenance of awareness and compatibility with international activities is underway. Activities related to high-level radioactive waste, uranium mine and mill tailings, low- and medium-level wastes, radioactive effluents from nuclear facilities, and decommissioning and decontamination are described

  11. Managing commercial high-level radioactive waste

    International Nuclear Information System (INIS)

    1983-01-01

    The article is a summary of issues raised during US Congress deliberations on nuclear waste policy legislation. It is suggested that, if history is not to repeat itself, and the current stalemate on nuclear waste is not to continue, a comprehensive policy is needed that addresses the near-term problems of interim storage as part of an explicit and credible program for dealing with the longer term problem of developing a final isolation system. Such a policy must: 1) adequately address the concerns and win the support of all the major interested parties, and 2) adopt a conservative technical and institutional approach - one that places high priority on avoiding the problems that have repeatedly beset the program in the past. It is concluded that a broadly supported comprehensive policy would contain three major elements, each designed to address one of the key questions concerning Federal credibility: commitment in law to the goals of a comprehensive policy; credible institutional mechanisms for meeting goals; and credible measures for addressing the specific concerns of the states and the various publics. Such a policy is described in detail. (Auth.)

  12. Identification of items and activities important to waste form acceptance by Westinghouse GoCo sites

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Marra, S.L.; Dempster, J.; Randklev, E.H.

    1993-01-01

    The Department of Energy has established specifications (Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms, or WAPS) for canistered waste forms produced at Hanford, Savannah River, and West Valley. Compliance with these specifications requires that each waste form producer identify the items and activities which must be controlled to ensure compliance. As part of quality assurance oversight activities, reviewers have tried to compare the methodologies used by the waste form producers to identify items and activities important to waste form acceptance. Due to the lack of a documented comparison of the methods used by each producer, confusion has resulted over whether the methods being used are consistent. This confusion has been exacerbated by different systems of nomenclature used by each producer, and the different stages of development of each project. The waste form producers have met three times in the last two years, most recently on June 28, 1993, to exchange information on each producer's program. These meetings have been sponsored by the Westinghouse GoCo HLW Vitrification Committee. This document is the result of this most recent exchange. It fills the need for a documented comparison of the methodologies used to identify items and activities important to waste form acceptance. In this document, the methodology being used by each waste form producer is summarized, and the degree of consistency among the waste form producers is determined

  13. Wastes Characterisation from Foundry Activities on European Level

    International Nuclear Information System (INIS)

    Andres, I.; Ruiz, C.; Ibanez, R.; Viguri, J.; Irabien, A.

    1999-01-01

    This work presents The results of the eco toxicological characterisation of 22 defined wastes from steel foundry activities. The wastes have been selected from three processes, steel mill (smelting). sand casting and cleaning and finishing of steel products,with the common characteristics of represent an important industrial activity in the area and generated the wastes considered in this study. The eco toxicological characterisation obtained applying the Spanish regulations on hazardous waste is compared to the hazardous attributions considered by the European Union in order to characterise a waste as hazardous (non hazardous). The results allow to conclude that a acceptable concordance between both methodologies is reached and remark the need to split the broad generic types of wastes given by the Spanish regulation (Eco toxic / non eco toxic) into clearly identifiable specific types of waste

  14. Engineering materials for high level radioactive waste repository

    International Nuclear Information System (INIS)

    Wen Zhijian

    2009-01-01

    Radioactive wastes can arise from a wide range of human activities and have different physical and chemical forms with various radioactivity. The high level radioactive wastes (HLW)are characterized by nuclides of very high initial radioactivity, large thermal emissivity and the long life-term. The HLW disposal is highly concerned by the scientists and the public in the world. At present, the deep geological disposal is regarded as the most reasonable and effective way to safely dispose high-level radioactive wastes in the world. The conceptual model of HLW geological disposal in China is based on a multi-barrier system that combines an isolating geological environment with an engineering barrier system(EBS). The engineering materials in EBS include the vitrified HLW, canister, overpack, buffer materials and backfill materials. Referring to progress in the world, this paper presents the function, the requirement for material selection and design, and main scientific projects of R and D of engineering materials in HLW repository. (authors)

  15. Special Analysis for Disposal of High-Concentration I-129 Waste in the Intermediate-Level Vaults at the E-Area Low-Level Waste Facility

    International Nuclear Information System (INIS)

    Collard, L.B.

    2000-01-01

    This revision was prepared to address comments from DOE-SR that arose following publication of revision 0. This Special Analysis (SA) addresses disposal of wastes with high concentrations of I-129 in the Intermediate-Level (IL) Vaults at the operating, low-level radioactive waste disposal facility (the E-Area Low-Level Waste Facility or LLWF) on the Savannah River Site (SRS). This SA provides limits for disposal in the IL Vaults of high-concentration I-129 wastes, including activated carbon beds from the Effluent Treatment Facility (ETF), based on their measured, waste-specific Kds

  16. Special Analysis for Disposal of High-Concentration I-129 Waste in the Intermediate-Level Vaults at the E-Area Low-Level Waste Facility

    Energy Technology Data Exchange (ETDEWEB)

    Collard, L.B.

    2000-09-26

    This revision was prepared to address comments from DOE-SR that arose following publication of revision 0. This Special Analysis (SA) addresses disposal of wastes with high concentrations of I-129 in the Intermediate-Level (IL) Vaults at the operating, low-level radioactive waste disposal facility (the E-Area Low-Level Waste Facility or LLWF) on the Savannah River Site (SRS). This SA provides limits for disposal in the IL Vaults of high-concentration I-129 wastes, including activated carbon beds from the Effluent Treatment Facility (ETF), based on their measured, waste-specific Kds.

  17. Review of high-level waste form properties. [146 bibliographies

    Energy Technology Data Exchange (ETDEWEB)

    Rusin, J.M.

    1980-12-01

    This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison.

  18. Review of high-level waste form properties

    International Nuclear Information System (INIS)

    Rusin, J.M.

    1980-12-01

    This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison

  19. The geochemistry of high-level waste disposal in granitic rocks

    International Nuclear Information System (INIS)

    Chapman, N.A.; Sargent, F.P.

    1984-01-01

    Under the auspices of the cooperative agreement between Euratom and Atomic Energy of Canada Ltd about radioactive waste management and disposal, a joint workshop was held on the topic of the geochemistry of high-level waste disposal in granitic rocks. The report covers (1) waste form leaching, (2) thermodynamics, (3) geochemical models, (4) the role of colloids, (5) sorption phenomena, (6) the linking of flow and geochemical models, (7) microbial activity

  20. Integrating the commercial and defense high level waste programs - A utility perspective

    International Nuclear Information System (INIS)

    Tomonto, J.R.

    1986-01-01

    The Nuclear Waste Policy Act of 1982 provided that disposal of high-level wastes resulting from defense activities be included in the authorized repository unless the President determined that separate facilities are required. President Reagan approved commingling of defense and civilian wastes on April 30, 1985. The impacts of this decision on the repository schedule, civilian spent fuel acceptance rates, and cost sharing are reviewed and recommendations for resolving these issues are presented

  1. Advanced waste form and Melter development for treatment of troublesome high-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (also with high Al2O3 concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both

  2. The conceptual design of waste repository for radioactive waste from medical, industrial and research facilities containing comparatively high radioactivity

    International Nuclear Information System (INIS)

    Yamamoto, Masayuki; Hashimoto, Naro

    2002-02-01

    Advisory Committee on Nuclear Fuel Cycle Backend Policy reported the basic approach to the RI and Institute etc. wastes on March 2002. According to it, radioactive waste form medical, industrial and research facilities should be classified by their radioactivity properties and physical and chemical properties, and should be disposed in the appropriate types of repository with that classification. For the radioactive waste containing comparatively high radioactivity generated from reactors, NSC has established the Concentration limit for disposal. NSC is now discussing about the limit for the radioactive waste from medical, industrial and research facilities containing comparatively high radioactivity. Japan Nuclear Cycle Development Institute (JNC) preliminary studied about the repository for radioactive waste from medical, industrial and research facilities and discussed about the problems for design on H12. This study was started to consider those problems, and to develop the conceptual design of the repository for radioactive waste from medical, industrial and research facilities. Safety assessment for that repository is also performed. The result of this study showed that radioactive waste from medical, industrial and research facilities of high activity should be disposed in the repository that has higher performance of barrier system comparing with the vault type near surface facility. If the conditions of the natural barrier and the engineering barrier are clearer, optimization of the design will be possible. (author)

  3. Composite quarterly technical report: long-term high-level waste technology, October-December 1980

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-04-01

    The technical information in this report summarizes work performed at participating sites to immobilize high-level radioactive wastes. The areas reported are in: program management and support; waste preparation; waste fixation; and final handling. Majority of the studies were in the area of waste fixation, some of which are: leaching tests of ceramic forms, high silica glass, graphite powder and other carbon preparations; viscosity measurements for a range of waste-glass compositions from references borosilicate glass to high-alumina glasses; neutron activation analysis for measuring leach rates; preparation of SYNROC D spheres; formulations for preparing ceramics from defense waste composition; development of a pilot-scale glass melter, and kinetic studies of slag formation in glass melters

  4. Intergenerational ethics of high level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Kunihiko [Nagoya Univ., Graduate School of Engineering, Nagoya, Aichi (Japan); Nasu, Akiko; Maruyama, Yoshihiro [Shibaura Inst. of Tech., Tokyo (Japan)

    2003-03-01

    The validity of intergenerational ethics on the geological disposal of high level radioactive waste originating from nuclear power plants was studied. The result of the study on geological disposal technology showed that the current method of disposal can be judged to be scientifically reliable for several hundred years and the radioactivity level will be less than one tenth of the tolerable amount after 1,000 years or more. This implies that the consideration of intergenerational ethics of geological disposal is meaningless. Ethics developed in western society states that the consent of people in the future is necessary if the disposal has influence on them. Moreover, the ethics depends on generally accepted ideas in western society and preconceptions based on racism and sexism. The irrationality becomes clearer by comparing the dangers of the exhaustion of natural resources and pollution from harmful substances in a recycling society. (author)

  5. Intergenerational ethics of high level radioactive waste

    International Nuclear Information System (INIS)

    Takeda, Kunihiko; Nasu, Akiko; Maruyama, Yoshihiro

    2003-01-01

    The validity of intergenerational ethics on the geological disposal of high level radioactive waste originating from nuclear power plants was studied. The result of the study on geological disposal technology showed that the current method of disposal can be judged to be scientifically reliable for several hundred years and the radioactivity level will be less than one tenth of the tolerable amount after 1,000 years or more. This implies that the consideration of intergenerational ethics of geological disposal is meaningless. Ethics developed in western society states that the consent of people in the future is necessary if the disposal has influence on them. Moreover, the ethics depends on generally accepted ideas in western society and preconceptions based on racism and sexism. The irrationality becomes clearer by comparing the dangers of the exhaustion of natural resources and pollution from harmful substances in a recycling society. (author)

  6. Activation of waste brewer's yeast Saccharomyces cerevisiae for bread production

    OpenAIRE

    Popov Stevan D.; Dodić Siniša N.; Mastilović Jasna S.; Dodić Jelena M.; Popov-Raljić Jovanka V.

    2005-01-01

    The waste brewer's yeast S. cerevisiae (activated and non-activated) was compared with the commercial baker's yeast regarding the volume of developed gas in dough, volume and freshness stability of produced bread. The activation of waste brewer's yeast resulted in the increased volume of developed gas in dough by 100% compared to non-activated brewer's yeast, and the obtained bread is of more stable freshness compared to bread produced with baker's yeast. The activation of BY affects positive...

  7. Radionuclide compositions of spent fuel and high level waste from commercial nuclear reactors

    International Nuclear Information System (INIS)

    Goodill, D.R.; Tymons, B.J.

    1984-10-01

    This report provides information on radionuclide compositions of spent fuel and high level waste produced during reprocessing. The reactor types considered are Magnox, AGR, PWR and CFR. The activities of the radionuclides are calculated using the FISPIN code. The results are presented in a form suitable for radioactive waste management calculations. (author)

  8. Tank waste remediation system high-level waste vitrification system development and testing requirements

    International Nuclear Information System (INIS)

    Calmus, R.B.

    1995-01-01

    This document provides the fiscal year (FY) 1995 recommended high-level waste melter system development and testing (D and T) requirements. The first phase of melter system testing (FY 1995) will focus on the feasibility of high-temperature operation of recommended high-level waste melter systems. These test requirements will be used to establish the basis for defining detailed testing work scope, cost, and schedules. This document includes a brief summary of the recommended technologies and technical issues associated with each technology. In addition, this document presents the key D and T activities and engineering evaluations to be performed for a particular technology or general melter system support feature. The strategy for testing in Phase 1 (FY 1995) is to pursue testing of the recommended high-temperature technologies, namely the high-temperature, ceramic-lined, joule-heated melter, referred to as the HTCM, and the high-frequency, cold-wall, induction-heated melter, referred to as the cold-crucible melter (CCM). This document provides a detailed description of the FY 1995 D and T needs and requirements relative to each of the high-temperature technologies

  9. High slope waste dumps – a proven possibility

    Directory of Open Access Journals (Sweden)

    Igor Svrkota

    2013-11-01

    Full Text Available This paper is an overview of dumping operations on High Slope Waste Dump at Veliki Krivelj open pit copper mine, RTB Bor, Serbia. The High Slope Waste Dump in Bor is the highest single slope waste dump in the world with the slope height of 405 m. The paper gives the basics and limitations of the designed dumping technology, the redesigned technology, gives an overview of the 13 year long operation and gathered experiences and addresses the main issues of dumping operations in high slope conditions as well as the present condition of the High Slope Waste Dump.

  10. Storage facility for highly radioactive solid waste

    International Nuclear Information System (INIS)

    Kitano, Shozo

    1996-01-01

    A heat insulation plate is disposed at an intermediate portion between a ceiling wall of a storage chamber and an upper plate of a storage pit in parallel with them. A large number of highly radioactive solid wastes contained in canisters are contained in the storage pit. Cooling air is introduced from an air suction port, passes a channel on the upper side of the heat insulation plate formed by the ceiling of the storage chamber and the heat insulation plate, and flows from a flow channel on the side of the wall of the storage chamber to the lower portion of the storage pit. Afterheat is removed by the air flown from the lower portion to ventilation tubes at the outer side of container tubes. The air heated to a high temperature through the flow channel on the lower side of the heat insulation plate between the heat insulation plate and the upper plate of the storage pit, and is exhausted to an exhaustion port. Further, a portion of a heat insulation plate as a boundary between the cooling air and a high temperature air formed on the upper portion of the storage pit is formed as a heat transfer plate, so that the heat of the high temperature air is removed by the cooling air flowing the upper flow channel. This can prevent heating of the ceiling wall of the storage chamber. (I.N.)

  11. International Symposium on Disposal of Low Activity Radioactive Waste, Cordoba, Spain, 13-17 December 2004

    CERN Document Server

    2004-01-01

    The topical issues addressed by the symposium were: policies and strategies for low activity radioactive waste; very low activity radioactive waste; low activity radioactive waste from decommissioning; long lived low activity radioactive waste and other materials; and unique low activity radioactive waste.

  12. Predicting the degradability of waste activated sludge.

    Science.gov (United States)

    Jones, Richard; Parker, Wayne; Zhu, Henry; Houweling, Dwight; Murthy, Sudhir

    2009-08-01

    The objective of this study was to identify methods for estimating anaerobic digestibility of waste activated sludge (WAS). The WAS streams were generated in three sequencing batch reactors (SBRs) treating municipal wastewater. The wastewater and WAS properties were initially determined through simulation of SBR operation with BioWin (EnviroSim Associates Ltd., Flamborough, Ontario, Canada). Samples of WAS from the SBRs were subsequently characterized through respirometry and batch anaerobic digestion. Respirometry was an effective tool for characterizing the active fraction of WAS and could be a suitable technique for determining sludge composition for input to anaerobic models. Anaerobic digestion of the WAS revealed decreasing methane production and lower chemical oxygen demand removals as the SRT of the sludge increased. BioWin was capable of accurately describing the digestion of the WAS samples for typical digester SRTs. For extended digestion times (i.e., greater than 30 days), some degradation of the endogenous decay products was assumed to achieve accurate simulations for all sludge SRTs.

  13. Stability of High-Level Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Besmann, Theodore M.; Vienna, John D.

    2005-09-30

    The objective of the proposed effort is to use a new approach to develop solution models of complex waste glass systems and spent fuel that are predictive with regard to composition, phase separation, and volatility. The effort will also yield thermodynamic values for waste components that are fundamentally required for corrosion models used to predict the leaching/corrosion behavior for waste glass and spent fuel material. This basic information and understanding of chemical behavior can subsequently be used directly in computational models of leaching and transport in geologic media, in designing and engineering waste forms and barrier systems, and in prediction of chemical interactions.

  14. Cermet high level waste forms: a pregress report

    International Nuclear Information System (INIS)

    Aaron, W.S.; Quinby, T.C.; Kobisk, E.H.

    1978-06-01

    The fixation of high level radioactive waste from both commercial and DOE defense sources as cermets is currently under study. This waste form consists of a continuous iron-nickel base metal matrix containing small particles of fission product oxides. Preliminary evaluations of cermets fabricated from a variety of simulated wastes indicate they possess properties providing advantages over other waste forms presently being considered, namely thermal conductivity, waste loading levels, and leach resistance. This report describes the progress of this effort, to date, since its initiation in 1977

  15. The IAEA's activities in the field of radioactive waste management

    International Nuclear Information System (INIS)

    Semenov, B.A.

    1984-01-01

    The IAEA has been concerned with radioactive waste management since its inception. Its programme in this area was expanded in the mid 1970s as questions related to the management and disposal of radioactive wastes came into focus in conjunction with the further industrial development of nuclear power. The objectives of the Agency's waste management programme are to assist its Member States in the safe and effective management of wastes by organizing the exchange and dissemination of information, providing guidance and technical assistance and supporting research. The current programme addresses all aspects of the industrial use of nuclear power under the aspects (a) technology of handling and treatment of wastes, (b) underground disposal of wastes, (c) environmental aspects of nuclear energy, including sea disposal of radioactive wastes. Systematic reviews have been made and publications issued concerning the technology of handling, treating, conditioning, and storing various categories of wastes, including liquid and gaseous wastes, wastes from nuclear power plants, spent fuel reprocessing and mining and milling of uranium ores, as well as wastes from decommissioning of nuclear facilities. As waste disposal is the current issue of highest interest, an Agency programme was set up in 1977 to develop a set of guidelines on the safe underground disposal of low-, intermediate- and high-level wastes in shallow ground, rock cavities or deep geological repositories. This programme will continue until 1990. Eleven Safety Series and Technical Documents and Reports have been published under this programme so far, which also addresses safety and other criteria for waste disposal. The environmental part of the waste management programme is concerned with the assessment of radiological and non-radiological consequences of discharges from nuclear facilities, including de minimis concepts in waste disposal and environmental models and data for radionuclide releases

  16. Handling and storage of conditioned high-level wastes

    International Nuclear Information System (INIS)

    1983-01-01

    This report deals with certain aspects of the management of one of the most important wastes, i.e. the handling and storage of conditioned (immobilized and packaged) high-level waste from the reprocessing of spent nuclear fuel and, although much of the material presented here is based on information concerning high-level waste from reprocessing LWR fuel, the principles, as well as many of the details involved, are applicable to all fuel types. The report provides illustrative background material on the arising and characteristics of high-level wastes and, qualitatively, their requirements for conditioning. The report introduces the principles important in conditioned high-level waste storage and describes the types of equipment and facilities, used or studied, for handling and storage of such waste. Finally, it discusses the safety and economic aspects that are considered in the design and operation of handling and storage facilities

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

    Energy Technology Data Exchange (ETDEWEB)

    Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Piepel, Gregory F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lindberg, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Heasler, Patrick G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mercier, Theresa M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Russell, Renee L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cozzi, Alex [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Daniel, William E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Eibling, Russell E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hansen, E. K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Reigel, Marissa M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Swanberg, David J. [Washington River Protection Solutions (WRPS), Aiken, SC (United States)

    2013-09-30

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

  18. Novel Activated Carbons from Agricultural Wastes and their Characterization

    Directory of Open Access Journals (Sweden)

    S. Karthikeyan

    2008-01-01

    Full Text Available Solid waste disposal has become a major problem in India, Either it has to be disposed safely or used for the recovery of valuable materials as agricultural wastes like turmeric waste, ferronia shell waste, jatropha curcus seed shell waste, delonix shell waste and ipomea carnia stem. Therefore these wastes have been explored for the preparation of activated carbon employing various techniques. Activated carbons prepared from agricultural solid wastes by chemical activation processes shows excellent improvement in the surface characteristics. Their characterization studies such as bulk density, moisture content, ash content, fixed carbon content, matter soluble in water, matter soluble in acid, pH, decolourising power, phenol number, ion exchange capacity, ion content and surface area have been carried out to assess the suitability of these carbons as absorbents in the water and wastewater. For anionic dyes (reactive, direct, acid a close relationship between the surface area and surface chemical groups of the modified activated carbon and percentage of dye removal by adsorption can be observed. Cationic dyes large amount of surface chemical groups present in the sample (mainly carboxylic, anhydrides, lactones and phenols etc. are good anchoring sites for adsorption. The present study reveals the recovery of valuable adsorbents from readily and cheaply available agriculture wastes.

  19. Pelleted waste form for high-level ICPP wastes

    International Nuclear Information System (INIS)

    Lamb, K.M.; Priebe, S.J.; Cole, H.S.; Taki, B.D.

    1979-01-01

    Simulated zirconia type calcined waste is pelletized on a 41-cm dia disc pelletizer using 5% bentonite, 2% metakaolin, and 2% boric acid as a solid binder and 7M phosphoric plus 4M nitric acid as a liquid binder. After heat treatment at 800 0 C for 2 hours, the pellets are impact resistant and have a leach resistance of 10 -4 g/cm 2 /day, based on Soxhlet leaching for 100 hours at 95 0 C with distilled water. An integrated pilot plant is being fabricated to verify the process. 1 figure, 4 tables

  20. Pelleted waste form for high-level ICPP wastes

    International Nuclear Information System (INIS)

    Lamb, K.M.; Priebe, S.J.; Cole, H.S.; Taki, B.d.

    1979-01-01

    Simulated zirconia-type calcined waste is pelletized on a 41-cm diameter disc pelletizer using 5% bentonite, 2% metakaolin, and 2% boric acid as a solid binder and 7M phosphoric plus 4M nitric acid as a liquid binder. After heat treatment at 800 0 C for 2 hours the pellets are impact resistant and have a leach resistance of 10 -4 g/cm 2 . day, based on Soxhlet leaching for 100 hours at 95 0 C with distilled water. An integrated pilot plant is being fabricated to verify the process. 1 figure, 4 tables

  1. Electrical processes for the treatment of medium active liquid wastes

    International Nuclear Information System (INIS)

    Turner, A.D.; Bowen, W.R.; Bridger, N.J.; Junkinson, A.R.; Cox, D.R.

    1985-07-01

    Cross-flow electrokinetic dewatering has been developed on a lab-scale into an effective process for the treatment of such wastes as gravity-settled flocs, or sludges arising from fuel storage. The product may be concentrated to 25-42% solids while still remaining fluid, prior to immobilization - e.g. by addition of cement powder. Complete retention of activity in the concentrate was observed during the treatment of Harwell low-level waste sludges due to the high solids separation factor ( > 10 4 ). It is a low pressure, low temperature process - consuming only 0.03-0.13 kWh/L at permeation rates of 0.3-1.5 m/h (depending on the stream), corresponding to 1 /67 - 1 /15 that needed for evaporation. An advanced electrochemical ion-exchange system has been developed in which ionic material can be electrically adsorbed and eluted by polarity reversal > 1000 times, without any change in performance. Decontamination factors of about 2000 were achieved for Cs removal, up to 75% loading of the exchanger at flow rates of 8 bed volumes/h. Elution into water can give concentrates of >= 0.25 M - with consequent high volume reduction factors. Inorganic ion-exchangers have also demonstrated system selectivity for the removal of specific cations. Overall energy consumption is 3 ( 1 /400 evaporation). Significant cost savings over conventional ion-exchange may accrue from the improved performance under electrical control, and the reduced volumes of waste requiring disposal. (author)

  2. Separating and stabilizing phosphate from high-level radioactive waste: process development and spectroscopic monitoring.

    Science.gov (United States)

    Lumetta, Gregg J; Braley, Jenifer C; Peterson, James M; Bryan, Samuel A; Levitskaia, Tatiana G

    2012-06-05

    Removing phosphate from alkaline high-level waste sludges at the Department of Energy's Hanford Site in Washington State is necessary to increase the waste loading in the borosilicate glass waste form that will be used to immobilize the highly radioactive fraction of these wastes. We are developing a process which first leaches phosphate from the high-level waste solids with aqueous sodium hydroxide, and then isolates the phosphate by precipitation with calcium oxide. Tests with actual tank waste confirmed that this process is an effective method of phosphate removal from the sludge and offers an additional option for managing the phosphorus in the Hanford tank waste solids. The presence of vibrationally active species, such as nitrate and phosphate ions, in the tank waste processing streams makes the phosphate removal process an ideal candidate for monitoring by Raman or infrared spectroscopic means. As a proof-of-principle demonstration, Raman and Fourier transform infrared (FTIR) spectra were acquired for all phases during a test of the process with actual tank waste. Quantitative determination of phosphate, nitrate, and sulfate in the liquid phases was achieved by Raman spectroscopy, demonstrating the applicability of Raman spectroscopy for the monitoring of these species in the tank waste process streams.

  3. LOW ACTIVITY WASTE FEED SOLIDS CARACTERIZATION AND FILTERABILITY TESTS

    Energy Technology Data Exchange (ETDEWEB)

    McCabe, D.; Crawford, C.; Duignan, M.; Williams, M.; Burket, P.

    2014-04-03

    The primary treatment of the tank waste at the DOE Hanford site will be done in the Waste Treatment and Immobilization Plant (WTP) that is currently under construction. The baseline plan for the WTP Pretreatment facility is to treat the waste, splitting it into High Level Waste (HLW) feed and Low Activity Waste (LAW) feed. Both waste streams are then separately vitrified as glass and sealed in canisters. The LAW glass will be disposed onsite in the Integrated Disposal Facility (IDF). There are currently no plans to treat the waste to remove technetium in the WTP Pretreatment facility, so its disposition path is the LAW glass. Options are being explored to immobilize the LAW portion of the tank waste, i.e., the LAW feed from the WTP Pretreatment facility. Removal of {sup 99}Tc from the LAW Feed, followed by off-site disposal of the {sup 99}Tc, would eliminate a key risk contributor for the IDF Performance Assessment (PA) for supplemental waste forms, and has potential to reduce treatment and disposal costs. Washington River Protection Solutions (WRPS) is developing some conceptual flow sheets for LAW treatment and disposal that could benefit from technetium removal. One of these flowsheets will specifically examine removing {sup 99}Tc from the LAW feed stream to supplemental immobilization. The conceptual flow sheet of the {sup 99}Tc removal process includes a filter to remove insoluble solids prior to processing the stream in an ion exchange column, but the characteristics and behavior of the liquid and solid phases has not previously been investigated. This report contains results of testing of a simulant that represents the projected composition of the feed to the Supplemental LAW process. This feed composition is not identical to the aqueous tank waste fed to the Waste Treatment Plant because it has been processed through WTP Pretreatment facility and therefore contains internal changes and recycle streams that will be generated within the WTP process. Although

  4. Properties and characteristics of high-level waste glass

    International Nuclear Information System (INIS)

    Ross, W.A.

    1977-01-01

    This paper has briefly reviewed many of the characteristics and properties of high-level waste glasses. From this review, it can be noted that glass has many desirable properties for solidification of high-level wastes. The most important of these include: (1) its low leach rate; (2) the ability to tolerate large changes in waste composition; (3) the tolerance of anticipated storage temperatures; (4) its low surface area even after thermal shock or impact

  5. Synroc - a multiphase ceramic for high level nuclear waste immobilisation

    International Nuclear Information System (INIS)

    Reeve, K.D.; Vance, E.R.; Hart, K.P.; Smith, K.L.; Lumpkin, G.R.; Mercer, D.J.

    1992-01-01

    Many natural minerals - particularly titanates - are very durable geochemically, having survived for millions of years with very little alteration. Moreover, some of these minerals have quantitatively retained radioactive elements and their daughter products over this time. The Synroc concept mimics nature by providing an all-titanate synthetic mineral phase assemblage to immobilise high level waste (HLW) from nuclear fuel reprocessing operations for safe geological disposal. In principle, many chemically hazardous inorganic wastes arising from industry could also be immobilised in highly durable ceramics and disposed of geologically, but in practice the cost structure of most industries is such that lower cost waste management solutions - for example, the development of reusable by-products or the use of cements rather than ceramics - have to be devised. In many thousands of aqueous leach tests at ANSTO, mostly at 70-90 deg C, Synroc has been shown to be exceptionally durable. The emphases of the recent ANSTO program have been on tailoring of the Synroc composition to varying HLW compositions, leach testing of Synroc containing radioactive transuranic actinides, study of leaching mechanisms by SEM and TEM, and the development and costing of a conceptual fully active Synroc fabrication plant design. A summary of recent results on these topics will be presented. 29 refs., 4 figs

  6. Waste minimization activity report for 1991

    International Nuclear Information System (INIS)

    Shoemaker, J.D.

    1992-01-01

    This is a waste reduction report for the Lawrence Livermore National Laboratory (LLNL) for 1991. The report covers the Main Site at Livermore and Site 300. Each research program at LLNL is described by its operation, administrative procedures, and waste minimization. Examples of the programs at LLNL are biomedical and environmental research, chemistry and materials science, and energy program and earth sciences. (MB)

  7. High-Level Waste Vitrification Facility Feasibility Study

    International Nuclear Information System (INIS)

    D. A. Lopez

    1999-01-01

    A ''Settlement Agreement'' between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste now stored at the Idaho Nuclear Technology and Engineering Center will be treated so that it is ready to be moved out of Idaho for disposal by a compliance date of 2035. This report investigates vitrification treatment of the high-level waste in a High-Level Waste Vitrification Facility based on the assumption that no more New Waste Calcining Facility campaigns will be conducted after June 2000. Under this option, the sodium-bearing waste remaining in the Idaho Nuclear Technology and Engineering Center Tank Farm, and newly generated liquid waste produced between now and the start of 2013, will be processed using a different option, such as a Cesium Ion Exchange Facility. The cesium-saturated waste from this other option will be sent to the Calcine Solids Storage Facilities to be mixed with existing calcine. The calcine and cesium-saturated waste will be processed in the High-Level Waste Vitrification Facility by the end of calendar year 2035. In addition, the High-Level Waste Vitrification Facility will process all newly-generated liquid waste produced between 2013 and the end of 2035. Vitrification of this waste is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the waste and pouring it into stainless-steel canisters that will be ready for shipment out of Idaho to a disposal facility by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory until they are sent to a national geologic repository. The operating period for vitrification treatment will be from the end of 2015 through 2035

  8. High-Level Waste Vitrification Facility Feasibility Study

    Energy Technology Data Exchange (ETDEWEB)

    D. A. Lopez

    1999-08-01

    A ''Settlement Agreement'' between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste now stored at the Idaho Nuclear Technology and Engineering Center will be treated so that it is ready to be moved out of Idaho for disposal by a compliance date of 2035. This report investigates vitrification treatment of the high-level waste in a High-Level Waste Vitrification Facility based on the assumption that no more New Waste Calcining Facility campaigns will be conducted after June 2000. Under this option, the sodium-bearing waste remaining in the Idaho Nuclear Technology and Engineering Center Tank Farm, and newly generated liquid waste produced between now and the start of 2013, will be processed using a different option, such as a Cesium Ion Exchange Facility. The cesium-saturated waste from this other option will be sent to the Calcine Solids Storage Facilities to be mixed with existing calcine. The calcine and cesium-saturated waste will be processed in the High-Level Waste Vitrification Facility by the end of calendar year 2035. In addition, the High-Level Waste Vitrification Facility will process all newly-generated liquid waste produced between 2013 and the end of 2035. Vitrification of this waste is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the waste and pouring it into stainless-steel canisters that will be ready for shipment out of Idaho to a disposal facility by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory until they are sent to a national geologic repository. The operating period for vitrification treatment will be from the end of 2015 through 2035.

  9. Testing waste forms containing high radionuclide loadings

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Neilson, R.M. Jr.; Rogers, R.D.

    1986-01-01

    The Low-Level Waste Data Base Development - EPICOR-II Resin/Liner Investigation Program of the US Nuclear Regulatory Commission (NRC) is obtaining information on radioactive waste during NRC-prescribed tests and in a disposal environment. This paper describes the resin solidification task of that program, including the present status and results to date

  10. Wastes from selected activities in two light-water reactor fuel cycles

    International Nuclear Information System (INIS)

    Palmer, C.R.; Hill, O.F.

    1980-07-01

    This report presents projected volumes and radioactivities of wastes from the production of electrical energy using light-water reactors (LWR). The projections are based upon data developed for a recent environmental impact statement in which the transuranic wastes (i.e., those wastes containing certain long-lived alpha emitters at concentrations of at least 370 becquerels, or 10 nCi, per gram of waste) from fuel cycle activities were characterized. In addition, since the WG.7 assumed that all fuel cycle wastes except mill tailings are placed in a mined geologic repository, the nontransuranic wastes from several activities are included in the projections reported. The LWR fuel cycles considered are the LWR, once-through fuel cycle (Strategy 1), in which spent fuel is packaged in metal canisters and then isolated in geologic formations; and the LWR U/Pu recycle fuel cycle (Strategy 2), wherein spent fuel is reprocessed for recovery and recycle of uranium and plutonium in LWRs. The wastes projected for the two LWR fuel cycles are summarized. The reactor operations and decommissioning were found to dominate the rate of waste generation in each cycle. These activities account for at least 85% of the fuel cycle waste volume (not including head-end wastes) when normalized to per unit electrical energy generated. At 10 years out of reactor, however, spent fuel elements in Strategy 1 represent 98% of the fuel cycle activity but only 4% of the volume. Similarly, the packaged high-level waste, fuel hulls and hardware in Strategy 2 concentrate greater than 95% of the activity in 2% of the waste volume

  11. Hierarchically porous nitrogen-doped carbon derived from the activation of agriculture waste by potassium hydroxide and urea for high-performance supercapacitors

    Science.gov (United States)

    Zou, Kaixiang; Deng, Yuanfu; Chen, Juping; Qian, Yunqian; Yang, Yuewang; Li, Yingwei; Chen, Guohua

    2018-02-01

    Nitrogen-doped carbon with an ultra-high specific surface area and a hierarchically interconnected porous structure is synthesized in large scale from a green route, that is, the activation of bagasse via a one-step method using KOH and urea. KOH and urea play a synergistic effect for the enhancement of the specific surface area and the modification of pore size of the as-prepared material. Benefiting from the multiple synergistic roles originated from an ultra-high specific area (2905.4 m2 g-1), a high porous volume (2.05 mL g-1 with 75.6 vol% micropores, which is an ideal proportion of micropores for obtaining high specific capacitance), a suitable nitrogen content (2.63 wt%), and partial graphitization, the hierarchically interconnected porous N-doped carbon exhibits an excellent electrochemical performance with a high specific capacitance (350.8, 301.9, and 259.5 F g-1 at 1.0 A g-1 in acidic, alkaline, and neutral electrolytes, respectively), superior rate capability and excellent cycling stability (almost no capacitance loss up to 5000 cycles). Furthermore, the symmetric device assembled by this material achieves high energy densities of 39.1 and 23.5 Wh kg-1 at power densities of 1.0 and 20 kW kg-1, respectively, and exhibits an excellent long-term cycling stability (with capacitance retention above 95.0% after 10 000 cycles).

  12. Managing the nation's commercial high-level radioactive waste

    International Nuclear Information System (INIS)

    1985-03-01

    This report presents the findings and conclusions of OTA's analysis of Federal policy for the management of commercial high-level radioactive waste. It represents a major update and expansion of the Analysis presented to Congress in our summary report, Managing Commercial High-Level Radioactive Waste, published in April of 1982 (NWPA). This new report is intended to contribute to the implementation of NWPA, and in particular to Congressional review of three major documents that DOE will submit to the 99th Congress: a Mission Plan for the waste management program; a monitored retrievable storage (MRS) proposal; and a report on mechanisms for financing and managing the waste program. The assessment was originally focused on the ocean disposal of nuclear waste. OTA later broadened the study to include all aspects of high-level waste disposal. The major findings of the original analysis were published in OTA's 1982 summary report

  13. Techniques for the solidification of high-level wastes

    International Nuclear Information System (INIS)

    1977-01-01

    The problem of the long-term management of the high-level wastes from the reprocessing of irradiated nuclear fuel is receiving world-wide attention. While the majority of the waste solutions from the reprocessing of commercial fuels are currently being stored in stainless-steel tanks, increasing effort is being devoted to developing technology for the conversion of these wastes into solids. A number of full-scale solidification facilities are expected to come into operation in the next decade. The object of this report is to survey and compare all the work currently in progress on the techniques available for the solidification of high-level wastes. It will examine the high-level liquid wastes arising from the various processes currently under development or in operation, the advantages and disadvantages of each process for different types and quantities of waste solutions, the stages of development, the scale-up potential and flexibility of the processes

  14. Radioactive Demonstrations Of Fluidized Bed Steam Reforming As A Supplementary Treatment For Hanford's Low Activity Waste And Secondary Wastes

    International Nuclear Information System (INIS)

    Jantzen, C.; Crawford, C.; Cozzi, A.; Bannochie, C.; Burket, P.; Daniel, G.

    2011-01-01

    The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP's LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as Cs-137, I-129, Tc-99, Cl, F, and SO4 that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap. The current waste disposal path for the WTP-SW is to recycle it to the supplemental LAW treatment to avoid a large steady state accumulation in the pretreatment-vitrification loop. Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750 C) continuous method by which LAW and/or WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides

  15. High levels of antimony in dust from e-waste recycling in southeastern China

    International Nuclear Information System (INIS)

    Bi, Xiangyang; Li, Zhonggen; Zhuang, Xiaochun; Han, Zhixuan; Yang, Wenlin

    2011-01-01

    Environmental contamination due to uncontrolled e-waste recycling is an emerging global issue. Antimony (Sb) is a toxic element used in semiconductor components and flame retardants for circuit board within electronic equipment. When e-waste is recycled, Sb is released and contaminates the surrounding environment; however, few studies have characterized the extent of this problem. In this study, we investigated Sb and arsenic (As) distributions in indoor dust from 13 e-waste recycling villages in Guiyu, Guangdong Province, southeastern China. Results revealed significantly elevated concentrations of Sb (6.1-232 mg/kg) in dust within all villages, which were 3.9-147 times higher than those from the non e-waste sites, indicating e-waste recycling was an important source of Sb pollution. On the contrary, As concentrations (5.4-17.7 mg/kg) in e-waste dusts were similar to reference values from the control sites. Therefore, dusts emitted from e-waste recycling may be characterized by high Sb/As ratios, which may help identify the contamination due to the e-waste recycling activities. - Highlights: → Antimony and arsenic concentrations in dust from e-waste recycling were investigated. → E-waste recycling is an important emerging source of Sb pollution. → Sb/As ratios may help identify the e-waste contamination.

  16. High levels of antimony in dust from e-waste recycling in southeastern China

    Energy Technology Data Exchange (ETDEWEB)

    Bi, Xiangyang, E-mail: bixy@cug.edu.cn [State Key Laboratory of Biogeology and Environmental Geology, China University of Geosciences, Wuhan 430074 (China); Faculty of Earth Science, China University of Geosciences, Wuhan 430074 (China); Li, Zhonggen [State Key Laboratory of Environmental Geochemistry, Institute of Geochemistry, Chinese Academy of Sciences, Guiyang 550002 (China); Zhuang, Xiaochun [Faculty of Materials Science and Chemical Engineering, China University of Geosciences, Wuhan 430074 (China); Han, Zhixuan [State Key Laboratory of Biogeology and Environmental Geology, China University of Geosciences, Wuhan 430074 (China); Faculty of Earth Science, China University of Geosciences, Wuhan 430074 (China); Yang, Wenlin [Faculty of Earth Science, China University of Geosciences, Wuhan 430074 (China)

    2011-11-01

    Environmental contamination due to uncontrolled e-waste recycling is an emerging global issue. Antimony (Sb) is a toxic element used in semiconductor components and flame retardants for circuit board within electronic equipment. When e-waste is recycled, Sb is released and contaminates the surrounding environment; however, few studies have characterized the extent of this problem. In this study, we investigated Sb and arsenic (As) distributions in indoor dust from 13 e-waste recycling villages in Guiyu, Guangdong Province, southeastern China. Results revealed significantly elevated concentrations of Sb (6.1-232 mg/kg) in dust within all villages, which were 3.9-147 times higher than those from the non e-waste sites, indicating e-waste recycling was an important source of Sb pollution. On the contrary, As concentrations (5.4-17.7 mg/kg) in e-waste dusts were similar to reference values from the control sites. Therefore, dusts emitted from e-waste recycling may be characterized by high Sb/As ratios, which may help identify the contamination due to the e-waste recycling activities. - Highlights: {yields} Antimony and arsenic concentrations in dust from e-waste recycling were investigated. {yields} E-waste recycling is an important emerging source of Sb pollution. {yields} Sb/As ratios may help identify the e-waste contamination.

  17. Calculations on the development in space and time of the temperature field around a repository of medium and high active wastes in a salt formation

    International Nuclear Information System (INIS)

    Delisle, G.

    1980-01-01

    The concept of nuclear waste disposal of th of the Federal Republic of Germany calls for the burial of the wastes within a salt formation. A small portion of the wastes will generate heat after the disposal procedure. A temperature rise within the salt formation, in space and time limited, will be the consequence. The temperature change at any point in the near or far field of the disporal area can be calculated with the aid of numerical models. The thermal parameters representative for the bulk material of the Zechstein formation in NW-Germany, on which the calculations are based, will be discussed in detail. The interrelation between the concentration of heat producing wastes in the disposal field and the maximum average temperature in the salt formation will be treated. By defining numerical models, which are based on assumed shapes of a salt dome and a disposal area, the temperature development in the near and far field of a nuclear repository are shown. (orig.) [de

  18. Capability and limitation study of the DDT passive-active neutron waste assay instrument

    International Nuclear Information System (INIS)

    Nicholas, N.J.; Coop, K.L.; Estep, R.J.

    1992-05-01

    The differential-dieaway-technique passive-active neutron assay system is widely used by transuranic waste generators to certify their drummed waste for eventual shipment to the Waste Isolation Pilot Plant (WIPP). Stricter criteria being established for waste emplacement at the WIPP site has led to a renewed interest in improvements to and a better understanding of current nondestructive assay (NDA) techniques. Our study includes the effects of source position, extreme matrices, high neutron backgrounds, and source self-shielding to explore the system's capabilities and limitations and to establish a basis for comparison with other NDA systems. 11 refs

  19. Quarterly Progress Report Research And Development Activities Waste Fixation Program October Through December 1976

    International Nuclear Information System (INIS)

    McElroy, J. L.

    1977-01-01

    Research and development activities of the Waste Fixation Program for October through December 1976 are described in this report. The objective of this program is to develop processes to convert high-level radioactive liquid waste (HLLW) to solid forms that are demonstrated to be physically, chemically, and radiolytically stable and inert. The scope of this program encompasses plans to make available a flexible advancing technology for the solidification of radioactive waste. Early technology will produce borosilicate glass by in-can melting and continuous electric melters. Multibarrier waste forms will be developed for future application

  20. Mining techniques and some aspects of high-level waste disposal

    International Nuclear Information System (INIS)

    Hoefnagels, J.A.R.

    1980-01-01

    The solutions to many problems of underground waste disposal involve mine engineering. This article attempts to highlight chosen issues and thereby create an overall impression, avoiding emphasis on single-aspect calculation. High level waste (H.L.W.) dominates current radioactive waste studies because of its specific characteristics and is therefore dealt with in this paper. However, depending on the method of disposal the other categories of radio active waste might become problems by themselves because of the relatively large quantities involved. (Auth.)

  1. Active waste disposal monitoring at the Radioactive Waste Management Complex, Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Hubbell, J.M.

    1990-10-01

    This report describes an active waste disposal monitoring system proposed to be installed beneath the low-level radioactive disposal site at the Radioactive Waste Management Complex (RWMC), Idaho National Engineering Laboratory, Idaho. The monitoring instruments will be installed while the waste is being disposed. Instruments will be located adjacent to and immediately beneath the disposal area within the unsaturated zone to provide early warning of contaminant movement before contaminants reach the Snake River Plain Aquifer. This study determined the optimum sampling techniques using existing monitoring equipment. Monitoring devices were chosen that provide long-term data for moisture content, movement of gamma-emitting nuclides, and gas concentrations in the waste. The devices will allow leachate collection, pore-water collection, collection of gasses, and access for drilling through and beneath the waste at a later time. The optimum monitoring design includes gas sampling devices above, within, and below the waste. Samples will be collected for methane, tritium, carbon dioxide, oxygen, and volatile organic compounds. Access tubes will be utilized to define the redistribution of radionuclides within, above, and below the waste over time and to define moisture content changes within the waste using spectral and neutron logging, respectively. Tracers will be placed within the cover material and within waste containers to estimate transport times by conservative chemical tracers. Monitoring the vadose zone below, within, and adjacent to waste while it is being buried is a viable monitoring option. 12 refs., 16 figs., 1 tab

  2. Suggestions on R and D work of high-level radioactive waste disposal in China

    International Nuclear Information System (INIS)

    Xu Guoqing

    2012-01-01

    The difference between repository and generic underground facilities is described. Some differences and similarities of site selection between the low and medium radioactive waste disposal, nuclear power station and high-level radioactive waste repository are also discussed here. We trend to extremely emphasize the safety of high-level radioactive waste disposal because of high toxicity, long half-life and long safety disposal period of this kind of radioactive wastes; because radioactive waste in the repository is of high specific activities and buried in depth, it would be difficult to meddle with its safety. In case of repository system being destroyed, the author considers that in the stages of regional and area site selection, the first task is to investigate regional tectonic stability. Some problems about disposal options and others are also discussed in this paper. (author)

  3. Characteristics of solidified high-level waste products

    International Nuclear Information System (INIS)

    1979-01-01

    The object of the report is to contribute to the establishment of a data bank for future preparation of codes of practice and standards for the management of high-level wastes. The work currently in progress on measuring the properties of solidified high-level wastes is being studied

  4. Process for solidifying high-level nuclear waste

    Science.gov (United States)

    Ross, Wayne A.

    1978-01-01

    The addition of a small amount of reducing agent to a mixture of a high-level radioactive waste calcine and glass frit before the mixture is melted will produce a more homogeneous glass which is leach-resistant and suitable for long-term storage of high-level radioactive waste products.

  5. Safe management of waste from health-care activities

    International Nuclear Information System (INIS)

    Pruess, A.; Giroult, E.; Rushbrook, P.

    1999-01-01

    The waste produced in the course of health-care activities, from contaminated needles to radioactive isotopes, carries a greater potential for causing infection and injury than any other type of waste, and inadequate or inappropriate management is likely to have serious public health consequences and deleterious effects on the environment. This handbook - the result of extensive international consultation and collaboration - provides comprehensive guidance on safe, efficient, and environmentally sound methods for the handling and disposal of health-care wastes. The various categories of waste are clearly defined and the particular hazards that each poses are described. Considerable prominence is given to the careful planning that is essential for the success of waste management; workable means of minimizing waste production are outlined and the role of reuse and recycling of waste is discussed. Most of the text, however, is devoted to the collection, segregation, storage, transport, and disposal of wastes. Details of containers for each category of waste, labelling of waste packages, and storage conditions are provided, and the various technologies for treatment of waste and disposal of final residues are discussed at length. Advice is given on occupational safety for all personnel involved with waste handling, and a separate chapter is devoted to the closely related topic of hospital hygiene and infection control. The handbook pays particular attention to basic processes and technologies that are not only safe but also affordable, sustainable, and culturally appropriate. For health-care settings in which resources are severely limited there is a separate chapter on minimal programmes; this summarizes all the simplest and least costly techniques that can be employed for the safe management of health-care wastes. The guide is aimed at public health managers and policy-makers, hospital managers, environmental health professionals, and all administrators with an

  6. Method to determine the activity concentration and total activity of radioactive waste

    International Nuclear Information System (INIS)

    Angeles C, A.

    2001-02-01

    A characteristic system of radioactive waste is described to determine the concentration of radionuclides activity and the total activity of bundles of radioactive waste. The system this integrated by three subsystems: - Elevator of drums. - Electromechanics. - Gamma spectroscopy. In the system it is analyzed waste of issuing gamma specifically, and this designed for materials of relative low density and it analyzes materials of cylindrical recipients

  7. Development of MHI's induction melting system for low level radio active solid waste treatment

    International Nuclear Information System (INIS)

    Murakami, Tadashi; Hashiba, Kenji; Fukui, Hiroshi; Sato, Akio; Minemoto, Masaki

    1999-01-01

    Mitsubishi Heavy Industries, Ltd., (MHI) has developed melting facilities that reduce radioactive waste volume. The system uses a high-frequency induction to separately melt nonmetallic waste in SUS containers and metallic waste. Use of system extends refractory life. To validate system feasibility, major components were tested with the following results: (1) Two 200-liter drum cans of molten solid waste are produced per work day, (2) Radioactivity in molten solid was homogeneous with a coefficient of variation ≤10%, clarifying residue properties, (3) The radioactive decontamination factor of off-gas facilities --DF=Activity to system/Activity at the system exit --exceeded 10 7 . We confirmed system to fill the requirements for molten solid waste and have the merit of high volume-reduction and long-life refractory. (author)

  8. Final report on cermet high-level waste forms

    International Nuclear Information System (INIS)

    Kobisk, E.H.; Quinby, T.C.; Aaron, W.S.

    1981-08-01

    Cermets are being developed as an alternate method for the fixation of defense and commercial high level radioactive waste in a terminal disposal form. Following initial feasibility assessments of this waste form, consisting of ceramic particles dispersed in an iron-nickel base alloy, significantly improved processing methods were developed. The characterization of cermets has continued through property determinations on samples prepared by various methods from a variety of simulated and actual high-level wastes. This report describes the status of development of the cermet waste form as it has evolved since 1977. 6 tables, 18 figures

  9. Evaluation of solidified high-level waste forms

    International Nuclear Information System (INIS)

    1981-01-01

    One of the objectives of the IAEA waste management programme is to coordinate and promote development of improved technology for the safe management of radioactive wastes. The Agency accomplished this objective specifically through sponsoring Coordinated Research Programmes on the ''Evaluation of Solidified High Level Waste Products'' in 1977. The primary objectives of this programme are to review and disseminate information on the properties of solidified high-level waste forms, to provide a mechanism for analysis and comparison of results from different institutes, and to help coordinate future plans and actions. This report is a summary compilation of the key information disseminated at the second meeting of this programme

  10. Waste package designs for disposal of high-level waste in salt formations

    International Nuclear Information System (INIS)

    Basham, S.J. Jr.; Carr, J.A.

    1984-01-01

    In the United States of America the selected method for disposal of radioactive waste is mined repositories located in suitable geohydrological settings. Currently four types of host rocks are under consideration: tuff, basalt, crystalline rock and salt. Development of waste package designs for incorporation in mined salt repositories is discussed. The three pertinent high-level waste forms are: spent fuel, as disassembled and close-packed fuel pins in a mild steel canister; commercial high-level waste (CHLW), as borosilicate glass in stainless-steel canisters; defence high-level waste (DHLW), as borosilicate glass in stainless-steel canisters. The canisters are production and handling items only. They have no planned long-term isolation function. Each waste form requires a different approach in package design. However, the general geometry and the materials of the three designs are identical. The selected waste package design is an overpack of low carbon steel with a welded closure. This container surrounds the waste forms. Studies to better define brine quantity and composition, radiation effects on the salt and brines, long-term corrosion behaviour of the low carbon steel, and the leaching behaviour of the spent fuel and borosilicate glass waste forms are continuing. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-12-30

    This Test Plan describes work to support the development and testing of high waste loading glass formulations that achieve high glass melting rates for Hanford high aluminum high level waste (HLW). In particular, the present testing is designed to evaluate the effect of using low activity waste (LAW) waste streams as a source of sodium in place ofchemical additives, sugar or cellulose as a reductant, boehmite as an aluminum source, and further enhancements to waste processing rate while meeting all processing and product quality requirements. The work will include preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM 100) tests designed to examine the effects of enhanced glass formulations, glass processing temperature, incorporation of the LAW waste stream as a sodium source, type of organic reductant, and feed solids content on waste processing rate and product quality. Also included is a confirmatory test on the HLW Pilot Melter (DM1200) with a composition selected from those tested on the DM100. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy's (DOE's) Office of River Protection (ORP) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same waste composition. This Test Plan is prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is about 12,500. This estimate is based upon the inventory ofthe tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat

  12. Radiation transport in high-level waste form

    International Nuclear Information System (INIS)

    Arakali, V.S.; Barnes, S.M.

    1992-01-01

    The waste form selected for vitrifying high-level nuclear waste stored in underground tanks at West Valley, NY is borosilicate glass. The maximum radiation level at the surface of a canister filled with the high-level waste form is prescribed by repository design criteria for handling and disposition of the vitrified waste. This paper presents an evaluation of the radiation transport characteristics for the vitreous waste form expected to be produced at West Valley and the resulting neutron and gamma dose rates. The maximum gamma and neutron dose rates are estimated to be less than 7500 R/h and 10 mRem/h respectively at the surface of a West Valley canister filled with borosilicate waste glass

  13. Quality checking task force destructive testing of active waste forms

    International Nuclear Information System (INIS)

    James, J.M.; Smith, D.L.

    1987-03-01

    The implications of sampling and testing of full size active packages of intermediate level wastes are summarised in this report. Sampling operations are technically feasible but a major difficulty will be the disposal of secondary waste. A literature survey indicated that destructive testing of wasteforms is not carried out as a routine operation in Europe or the USA. (author)

  14. Waste monitoring of the uranium ore processing activities in Romania

    International Nuclear Information System (INIS)

    Nica, L.

    2002-01-01

    The uranium ore processing activities at the Feldioara site produce a range of liquid and solid waste that are monitored. Liquids are treated through decantation, pH correction and uranium precipitation before their release into the environment. The solid waste is gathered into ore specific area and are covered regularly with clay materials. (author)

  15. Volume reduction through incineration of low-activity radioactive wastes

    International Nuclear Information System (INIS)

    Eymeri, J.; Gauthey, J.C.; Chaise, D.; Lafite, G.

    1993-01-01

    The aim of the waste treatment plant, designed by Technicatome (CEA) for an Indonesian Nuclear Research Center, is to reduce through incineration the volume of low-activity radioactive wastes such as technological solids (cotton, PVC, paper board), biological solids (animal bones) and liquids (cutting fluids...). The complete combustion is realized with a total air multi-fuel burner (liquid wastes) and flash pyrolysis-complete combustion (solid wastes). A two stage flue gas filtration system, a flue gas washing system, and an ash recovery system are used. A test platform has been built. 3 figs

  16. Decontamination of high-level waste canisters

    International Nuclear Information System (INIS)

    Nesbitt, J.F.; Slate, S.C.; Fetrow, L.K.

    1980-12-01

    This report presents evaluations of several methods for the in-process decontamination of metallic canisters containing any one of a number of solidified high-level waste (HLW) forms. The use of steam-water, steam, abrasive blasting, electropolishing, liquid honing, vibratory finishing and soaking have been tested or evaluated as potential techniques to decontaminate the outer surfaces of HLW canisters. Either these techniques have been tested or available literature has been examined to assess their applicability to the decontamination of HLW canisters. Electropolishing has been found to be the most thorough method to remove radionuclides and other foreign material that may be deposited on or in the outer surface of a canister during any of the HLW processes. Steam or steam-water spraying techniques may be adequate for some applications but fail to remove all contaminated forms that could be present in some of the HLW processes. Liquid honing and abrasive blasting remove contamination and foreign material very quickly and effectively from small areas and components although these blasting techniques tend to disperse the material removed from the cleaned surfaces. Vibratory finishing is very capable of removing the bulk of contamination and foreign matter from a variety of materials. However, special vibratory finishing equipment would have to be designed and adapted for a remote process. Soaking techniques take long periods of time and may not remove all of the smearable contamination. If soaking involves pickling baths that use corrosive agents, these agents may cause erosion of grain boundaries that results in rough surfaces

  17. High-level radioactive waste fixation in sintered vitreous matrix

    International Nuclear Information System (INIS)

    Russo, D.O.; Messi de Bernasconi, N.; Audero, M.A.

    1987-01-01

    The safe storage of high-level wastes from fuel elements reprocessing includes, as a first step, the fixation of the same in materials having a good resistance to the leaching in aqueous medium, such as borosilicate glass. As an alternative to the usual method of the molten glasses, a procedure for the sintering of a powdered glass and waste mixture at lower temperatures (600-700 deg C) has been developed, which minimizes the volatilization of active compounds during the process. Two glasses matrices of different composition and characteristics were used, to which the simulated wastes were added in the ratio of a 10% in weight of oxides. Two sintering techniques were employed 1: cold pressing and further sintering; 2: hot pressing and sintering under pressure. The densities were measured, the microstructure of the samples was analyzed and leaching essays were made in distilled water. The pellet's microstructure was observed by means of optical microscopy, by reflection in polished samples and by transparency in thin slices. The presence of crystalline compounds was analyzed by means of x rays and electron microprobe. The results have shown the convenience to continue with hot pressing essays, because a denser product with a higher resistance to the leaching is thus obtained. (M.E.L.) [es

  18. High density Polyethylene plastic waste treatment with microwave heating pyrolysis method using coconut-shell activated carbon to produce alternative fuels

    Science.gov (United States)

    Juliastuti, S. R.; Hisbullah, M. I.; Abdillah, M.

    2018-03-01

    Pyrolysis is a technology that could crack polimer such as plastic waste into alternative fuels. This research uses microwave heating methode, which more efficient than conventional heating methode. The plastic waste used is 200 grams of HDPE, with feed to catalyst weight ratio are 1:1, 0.6:1, 0.4:1. Pyrolysis was run at temperatures of 250, 300, 350, & 400 °C for 15, 30 and 45 min. From the experimental result, the best variable of pyrolysis process with microwave method is at 45 minutes, at 400°C, and 1:1 feed to catalyst weight ratio. Result shows that yield of liquid and gas product is 99.22%; yield of residue is 0.78%; value of liquid product’s composition (cycloparaffin and n-paraffin) is 54.09% and concentration of methane gas is 10.2%.

  19. Activities in department of energy hazardous and mixed waste defense waste management

    International Nuclear Information System (INIS)

    Eyman, L.D.

    1988-01-01

    In January 1986, the U.S. Department of Energy (DOE) Office of Assistant Secretary for Defense Programs (DP) created the Hazardous Waste and Remedial Actions Division within the Office of Defense Waste and Transportation Management. The Oak Ridge Operations Office (ORO) was assigned the responsibility for supporting DOE Headquarters (HQ) in planning nationally integrated activities for Resource Conservation and Recovery Act/Comprehensive Environmental Response, Compensation, and Liability Act/Superfund Amendments and Reauthorization Act (RCRA/CERCLA/SARA) compliance. In turn, ORO created the Hazardous Waste Remedial Actions Program Support Contractor Office (HAZWRAPSCO) to assist with the expanded lead assignment. The HAZWRAPSCO activities are currently supported by three distinct DOE-HQ funding elements: the Environmental Restoration Program, the Hazardous Waste Compliance Technology Program, and the Hazardous Waste Research and Development R and D Program. The Environmental Restoration Program is discussed in the paper, entitled The DOE Defense Program for Environmental Restoration

  20. High level waste management in France

    International Nuclear Information System (INIS)

    Sombret, C.; Bonniaud, R.; Jouan, A.

    1979-01-01

    This paper deals with solidification of radioactive wastes. The vitrification process has been selected, the Marcoule vitrification plant and the operation are described. Prospects for vitrification of fission product solutions from La Hague reprocessing plant are given

  1. Handling and storage of conditioned high-level wastes

    International Nuclear Information System (INIS)

    Heafield, W.

    1984-01-01

    This paper deals with certain aspects of the management of one of the most important radioactive wastes arising from the nuclear fuel cycle, i.e. the handling and storage of conditioned high-level wastes. The paper is based on an IAEA report of the same title published during 1983 in the Technical Reports Series. The paper provides illustrative background material on the characteristics of high-level wastes and, qualitatively, their requirements for conditioning. The principles important in the storage of high-level wastes are reviewed in conjunction with the radiological and socio-political considerations involved. Four fundamentally different storage concepts are described with reference to published information and the safety aspects of particular storage concepts are discussed. Finally, overall conclusions are presented which confirm the availability of technology for constructing and operating conditioned high-level waste storage facilities for periods of at least several decades. (author)

  2. High Resolution Sensor for Nuclear Waste Characterization

    International Nuclear Information System (INIS)

    Kanai Shah; William Higgins; Edgar V. Van Loef

    2006-01-01

    Gamma ray spectrometers are an important tool in the characterization of radioactive waste. Important requirements for gamma ray spectrometers used in this application include good energy resolution, high detection efficiency, compact size, light weight, portability, and low power requirements. None of the available spectrometers satisfy all of these requirements. The goal of the Phase I research was to investigate lanthanum halide and related scintillators for nuclear waste clean-up. LaBr 3 :Ce remains a very promising scintillator with high light yield and fast response. CeBr 3 is attractive because it is very similar to LaBr 3 :Ce in terms of scintillation properties and also has the advantage of much lower self-radioactivity, which may be important in some applications. CeBr 3 also shows slightly higher light yield at higher temperatures than LaBr 3 and may be easier to produce with high uniformity in large volume since it does not require any dopants. Among the mixed lanthanum halides, the light yield of LaBr x I 3-x :Ce is lower and the difference in crystal structure of the binaries (LaBr 3 and LaI 3 ) makes it difficult to grow high quality crystals of the ternary as the iodine concentration is increased. On the other hand, LaBr x I 3-x :Ce provides excellent performance. Its light output is high and it provides fast response. The crystal structures of the two binaries (LaBr 3 and LaCl 3 ) are very similar. Overall, its scintillation properties are very similar to those for LaBr 3 :Ce. While the gamma-ray stopping efficiency of LaBr x I 3-x :Ce is lower than that for LaBr 3 :Ce (primarily because the density of LaCl 3 is lower than that of LaBr 3 ), it may be easier to grow large crystals of LaBr x I 3-x :Ce than LaBr 3 :Ce since in some instances (for example, Cd x Zn 1-x Te), the ternary compounds provide increased flexibility in the crystal lattice. Among the new dopants, Eu 2+ and Pr 3+ , tried in LaBr 3 host crystals, the Eu 2+ doped samples exhibited

  3. Future directions of defense programs high-level waste technology programs

    International Nuclear Information System (INIS)

    Chee, T.C.; Shupe, M.W.; Turner, D.A.; Campbell, M.H.

    1987-01-01

    The Department of Energy has been managing high-level waste from the production of nuclear materials for defense activities over the last forty years. An objective for the Defense Waste and Transportation Management program is to develop technology which ensures the safe, permanent disposal of all defense radioactive wastes. Technology programs are underway to address the long-term strategy for permanent disposal of high-level waste generated at each Department of Energy site. Technology is being developed for assessing the hazards, environmental impacts, and costs of each long-term disposal alternative for selection and implementation. This paper addresses key technology development areas, and consideration of recent regulatory requirements associated with the long-term management of defense radioactive high-level waste

  4. High-level waste processing at the Savannah River Site: An update

    International Nuclear Information System (INIS)

    Marra, J.E.; Bennett, W.M.; Elder, H.H.; Lee, E.D.; Marra, S.L.; Rutland, P.L.

    1997-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) in Aiken, SC mg began immobilizing high-level radioactive waste in borosilicate glass in 1996. Currently, the radioactive glass is being produced as a ''sludge-only'' composition by combining washed high-level waste sludge with glass frit. The glass is poured in stainless steel canisters which will eventually be disposed of in a permanent, geological repository. To date, DWPF has produced about 100 canisters of vitrified waste. Future processing operations will, be based on a ''coupled'' feed of washed high-level waste sludge, precipitated cesium, and glass frit. This paper provides an update of the processing activities completed to date, operational/flowsheet problems encountered, and programs underway to increase production rates

  5. Winery waste recycling through anaerobic co-digestion with waste activated sludge.

    Science.gov (United States)

    Da Ros, C; Cavinato, C; Pavan, P; Bolzonella, D

    2014-11-01

    In this study biogas and high quality digestate were recovered from winery waste (wine lees) through anaerobic co-digestion with waste activated sludge both in mesophilic and thermophilic conditions. The two conditions studied showed similar yields (0.40 m(3)/kgCODfed) but different biological process stability: in fact the mesophilic process was clearly more stable than the thermophilic one in terms of bioprocess parameters. The resulting digestates showed good characteristics for both the tested conditions: heavy metals, dioxins (PCDD/F), and dioxin like bi-phenyls (PCBs) were concentred in the effluent if compared with the influent because of the important reduction of the solid dry matter, but remained at levels acceptable for agricultural reuse. Pathogens in digestate decreased. Best reductions were observed in thermophilic condition, while at 37°C the concentration of Escherichia coli was at concentrations level as high as 1000 UFC/g. Dewatering properties of digestates were evaluated by means of the capillary suction time (CST) and specific resistance to filtration (SRF) tests and it was found that a good dewatering level was achievable only when high doses of polymer (more than 25 g per kg dry solids) were added to sludge. Copyright © 2014 Elsevier Ltd. All rights reserved.

  6. Electrochemical ion-exchange for active liquid waste treatment

    International Nuclear Information System (INIS)

    Turner, A.D.; Bridger, N.J.; Jones, C.P.

    1992-10-01

    Electrochemical ion exchange (EIX) has been firmly established as an effective process for the treatment of a wide range of liquid radioactive wastes. Both organic (for low specific activity streams) and inorganic systems (for higher activity wastes) have been demonstrated. A low cost current feeder electrode has also been developed, with a projected lifetime of > 6 years. While cation EIX can be used for the treatment of low salt content streams, combination with anion EIX to control the pH can extend its range of application. At the same time, it is also able to remove activity complexed in an anionic form. AEIX has also demonstrated its ability to remove radionuclides with insoluble hydroxides (eg Co, U and Pu) from both high and low salt content streams. EIX has been successfully scaled-up form the bench-top scale by increasing electrode size by a factor of 11, and then by operating five units in parallel. An improvement in performance of by a factor 3 was observed over a simple increase in area, due to the minimization of edge effects in the larger units. The most significant advantage of EIX is its compactness -with plant sizes of 1000). (Author)

  7. Canadian high-level radioactive waste management system issues

    International Nuclear Information System (INIS)

    Allan, C.J.; Gray, B.R.

    1992-01-01

    In Canada responsibility for the management of radioactive wastes rests with the producer of those wastes. This fundamental principle applies to such diverse wastes as uranium mine and mill tailings, low-level wastes from universities and hospitals, wastes produced at nuclear research establishments, and wastes produced at nuclear generating stations. The federal government has accepted responsibility for historical wastes for which the original producer can no longer be held accountable. Management of radioactive wastes is subject to the regulatory control of the Atomic Energy Control Board, the federal agency responsible for regulating the nuclear industry. In this paper the authors summarize the current situation concerning the management of high level (used nuclear fuel) wastes. In 1981 the two governments also announced that selection of a disposal site would not proceed, and responsibility for site selection and operation would not be assigned until the Concept for used fuel disposal had been reviewed and assessed. Thus the concept assessment is generic rather than site specific. The Concept that has been developed has been designed to conform with safety and performance criteria established by the Atomic Energy Control Board. It is based on burial deep in plutonic rock of the Canadian Shield, using a multi-barrier approach with a series of engineered and natural barriers: these include the waste form, container, buffer and backfill, and the host rock

  8. Risk communication system for high level radioactive waste disposal

    International Nuclear Information System (INIS)

    Kugo, Akihide; Uda, Akinobu; Shimoda, Hirosi; Yoshikawa, Hidekazu; Ito, Kyoko; Wakabayashi, Yasunaga

    2005-01-01

    In order to gain a better understanding and acceptance of the task of implementing high level radioactive waste disposal, a study on new communication system about social risk information has been initiated by noticing the rapid expansion of Internet in the society. First, text mining method was introduced to identify the core public interest, examining public comments on the technical report of high level radioactive waste disposal. Then we designed the dialog-mode contents based on the theory of norm activation by Schwartz. Finally, the discussion board was mounted on the web site. By constructing such web communication system which includes knowledge base contents, introspective contents, and interactive discussion board, we conducted the experiment for verifying the principles such as that the basic technical knowledge and trust, and social ethics are indispensable in this process to close the perception gap between nuclear specialists and the general public. The participants of the experiment increased their interest in the topics with which they were not familiar and actively posted their opinions on the BBS. The dialog-mode contents were significantly more effective than the knowledge-based contents in promoting introspection that brought people into a greater awareness of problems such as social dilemma. (author)

  9. Chemical analysis of simulated high level waste glasses to support stage III sulfate solubility modeling

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-17

    The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms within the DOE complex. These wastes can contain relatively high concentrations of sulfate, which has low solubility in borosilicate glass. This is a significant issue for low-activity waste (LAW) glass and is projected to have a major impact on the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Sulfate solubility has also been a limiting factor for recent high level waste (HLW) sludge processed at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). The low solubility of sulfate in glass, along with melter and off-gas corrosion constraints, dictate that the waste be blended with lower sulfate concentration waste sources or washed to remove sulfate prior to vitrification. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerate mission completion.The objective of the current scope being pursued by SHU is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DWPF and WTP, allowing for enhanced waste loadings and waste throughput at these facilities. A series of targeted glass compositions was selected to resolve data gaps in the model and is identified as Stage III. SHU fabricated these glasses and sent samples to SRNL for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for the Stage III, simulated HLW glasses fabricated by SHU in support of the sulfate solubility model development.

  10. Learning the ABCs: Activity based costing in waste operations

    International Nuclear Information System (INIS)

    Zocher, Marc A.

    1992-01-01

    The United States Department of Energy (DOE) is facing a challenging new national role based on current world events, changing public perception and awareness, and a legacy of wastes generated in the past. Clearly, the DOE must put mechanisms in place to comply with environmental rules, regulations, and good management practices so that public health risk is minimized while programmatic costs are controlled. DOE has begun this process and has developed a Five-Year Plan to describe the activities necessary to comply with both cleanup, or environmental restoration, and waste management of existing waste streams. The focus of this paper is how to best manage the treatment, storage, disposal, and transportation of waste throughout the DOE weapons complex by using Activity Based Costing (ABC) to both plan and control expenditures in DOE Waste Management (WM). The basics of ABC, along with an example, will be detailed. (author)

  11. Development for low-activation concrete design reducing radioactive waste

    International Nuclear Information System (INIS)

    Kimura, Ken-ichi; Kinno, Masaharu; Hasegawa, Akira

    2008-01-01

    Full text: Concrete is very valuable and inexpensive material, however it can be changed to be expensive and hard to deal with in use of a nuclear plant after long operation. One of the counter plans for the above is to use low-activation concrete instead of the ordinary concrete, that will reduce radioactive waste and could be even below clearance level in decommissioning and that is very useful in term of life cycle cost. Radioactive analysis showed that Co and Eu were the major target elements which decide the radioactivity level of reinforced concrete in decommissioning stage, and a several material were selected as a low-activation raw material from wide survey of raw materials for concrete (typically aggregates and cements). With the canditate of raw materials, several low-activation concrete were proposed for various portion of light water reactor plant, which reduction ratio were 1/10 to 1/30 which were mainly consist of limestone and low heat cement or white cement, and 1/100 to 1/300 which were mainly consist of alumina aggregate or quartz and high almina cement, comparing to the ordinary concrete in ΣDi/Ci unit, where 'Di' indicates concentration of each residual radioisotope, Ci defined by IAEA as a clearance level, and suffition of 'i' indicates each radioisotope. National funded project for development of low-activation design method for reduction of radioactive waste below clearance level were started from 2005 with aiming (1) development of a database on the content of target elements, which transform radioactive nuclides, in raw materials of reinforced concrete, (2) development of calculation tools for estimation of residual radioactivity of plant components, and (3) development of low-activation materials for concrete such as cements and reinforcing steel bars for structural components. For the optimized design for applying low-activation concrete to the reactor portion, effective evaluation of neutron spectrum in the certain portion including

  12. Activated Carbon from the Chinese Herbal Medicine Waste by H3PO4 Activation

    Directory of Open Access Journals (Sweden)

    Tie Mi

    2015-01-01

    Full Text Available Large amounts of Chinese herbal medicine wastes produced by the medicinal factories have been mainly landfilled as waste. In this study, via phosphoric acid activation, a Chinese herbal medicine waste from Magnolia officinalis (CHMW-MO was prepared for activated carbon (CHMW-MO-AC. The effect of preparation conditions (phosphoric acid/CHMW-MO impregnation ratio, activation temperature, and time of activated carbon on yield of CHMW-MO-AC was investigated. The surface area and porous texture of the CHMW-MO-ACs were characterized by nitrogen adsorption at 77 K. The SBET and pore volume were achieved in their highest value of 920 m2/g and 0.703 cm3/g, respectively. Thermal gravity analysis and scanning electron microscope images showed that CHMW-MO-ACs have a high thermal resistance and pore development. The results indicated that CHMW-MO is a good precursor material for preparing activated carbon, and CHMW-MO-AC with well-developed mesopore volume can be prepared by H3PO4 activation.

  13. Managing the nation's commercial high-level radioactive waste

    International Nuclear Information System (INIS)

    1985-03-01

    This report presents the findings and conclusions of OTA's analysis of Federal policy for the management of commercial high-level radioactive waste. It is intended to contribute to the implementation of Nuclear Waste Policy Act of 1982 (NWPA). The major conclusion of that review is that NWPA provides sufficient authority for developing and operating a waste management system based on disposal in geologic repositories. Substantial new authority for other facilities will not be required unless major unexpected problems with geologic disposal are encountered. OTA also concludes that DOE's Draft Mission Plan published in 1984 falls short of its potential for enhancing the credibility and acceptability of the waste management program

  14. Solidification of Savannah River Plant high-level waste

    International Nuclear Information System (INIS)

    Maher, R.; Shafranek, L.F.; Stevens, W.R. III.

    1983-01-01

    The Department of Energy, in accord with recommendations from the Du Pont Company, has started construction of a Defense Waste Processing Facility (DWPF) at the Savannah River Plant. The facility should be completed by the end of 1988, and full-scale operation should begin in 1990. This facility will immobilize in borosilicate glass the large quantity of high-level radioactive waste now stored at the plant plus the waste to be generated from continued chemical reprocessing operations. The existing wastes at the Savannah River Plant will be completely converted by about 2010. 21 figures

  15. Engineering-scale vitrification of commercial high-level waste

    International Nuclear Information System (INIS)

    Bonner, W.F.; Bjorklund, W.J.; Hanson, M.S.; Knowlton, D.E.

    1980-04-01

    To date, technology for immobilizing commercial high-level waste (HLW) has been extensively developed, and two major demonstration projects have been completed, the Waste Solidification Engineering Prototypes (WSEP) Program and the Nuclear Waste Vitrification Project (NWVP). The feasibility of radioactive waste solidification was demonstrated in the WSEP program between 1966 and 1970 (McElroy et al. 1972) using simulated power-reactor waste composed of nonradioactive chemicals and HLW from spent, Hanford reactor fuel. Thirty-three engineering-scale canisters of solidified HLW were produced during the operations. In early 79, the NWVP demonstrated the vitrification of HLW from the processing of actual commercial nuclear fuel. This program consisted of two parts, (1) waste preparation and (2) vitrification by spray calcination and in-can melting. This report presents results from the NWVP

  16. Cementitious Stabilization of Mixed Wastes with High Salt Loadings

    International Nuclear Information System (INIS)

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

    1999-01-01

    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

  17. [Co-composting high moisture vegetable waste and flower waste in a sequential fed operation].

    Science.gov (United States)

    Zhang, Xiangfeng; Wang, Hongtao; Nie, Yongfeng

    2003-11-01

    Co-composting of high moisture vegetable wastes (celery and cabbage) and flower wastes (carnation) were studied in a sequential fed bed. The preliminary materials of composting were celery and carnation wastes. The sequential fed materials of composting were cabbage wastes and were fed every 4 days. Moisture content of mixture materials was between 60% and 70%. Composting was done in an aerobic static bed of composting based temperature feedback and control via aeration rate regulation. Aeration was ended when temperature of the pile was about 40 degrees C. Changes of composting of temperature, aeration rate, water content, organic matter, ash, pH, volume, NH4(+)-N, and NO3(-)-N were studied. Results show that co-composting of high moisture vegetable wastes and flower wastes, in a sequential fed aerobic static bed based temperature feedback and control via aeration rate regulation, can stabilize organic matter and removal water rapidly. The sequential fed operation are effective to overcome the difficult which traditional composting cannot applied successfully where high moisture vegetable wastes in more excess of flower wastes, such as Dianchi coastal.

  18. Nuclear waste management

    International Nuclear Information System (INIS)

    1982-12-01

    The subject is discussed, with special reference to the UK, under the headings: radiation; origins of the waste (mainly from nuclear power programme; gas, liquid, solid; various levels of activity); dealing with waste (methods of processing, storage, disposal); high-active waste (storage, vitrification, study of means of eventual disposal); waste management (UK organisation to manage low and intermediate level waste). (U.K.)

  19. R ampersand D activities at DOE applicable to mixed waste

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  20. Anaerobic bioleaching of metals from waste activated sludge

    KAUST Repository

    Meulepas, Roel J W; Gonzalez-Gil, Graciela; Teshager, Fitfety Melese; Witharana, Ayoma; Saikaly, Pascal; Lens, Piet Nl L

    2015-01-01

    Heavy metal contamination of anaerobically digested waste activated sludge hampers its reuse as fertilizer or soil conditioner. Conventional methods to leach metals require aeration or the addition of leaching agents. This paper investigates whether

  1. Viscosity-based high temperature waste form compositions

    International Nuclear Information System (INIS)

    Reimann, G.A.

    1994-01-01

    High-temperature waste forms such as iron-enriched basalt are proposed to immobilize and stabilize a variety of low-level wastes stored at the Idaho National Engineering Laboratory. The combination of waste and soil anticipated for the waste form results in high SiO 2 + Al 2 O 3 producing a viscous melt in an arc furnace. Adding a flux such as CaO to adjust the basicity ratio (the molar ratio of basic to acid oxides) enables tapping the furnace without resorting to extreme temperatures, but adds to the waste volume. Improved characterization of wastes will permit adjusting the basicity ratio to between 0.7 and 1.0 by blending of wastes and/or changing the waste-soil ratio. This minimizes waste form volume. Also, lower pouring temperatures will decrease electrode and refractory attrition, reduce vaporization from the melt, and, with suitable flux, facilitate crystallization. Results of laboratory tests were favorable and pilot-scale melts are planned; however, samples have not yet been subjected to leach testing

  2. Office of River Protection Advanced Low-Activity Waste Glass Research and Development Plan

    International Nuclear Information System (INIS)

    Kruger, A. A.; Peeler, D. K.; Kim, D. S.; Vienna, J. D.; Piepel, G. F.; Schweiger, M. J.

    2015-01-01

    The U.S. Department of Energy Office of River Protection (ORP) has initiated and leads an integrated Advanced Waste Glass (AWG) program to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product performance requirements. The integrated ORP program is focused on providing a technical, science-based foundation for making key decisions regarding the successful operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) facilities in the context of an optimized River Protection Project (RPP) flowsheet. The fundamental data stemming from this program will support development of advanced glass formulations, key product performance and process control models, and tactical processing strategies to ensure safe and successful operations for both the low-activity waste (LAW) and high-level waste vitrification facilities. These activities will be conducted with the objective of improving the overall RPP mission by enhancing flexibility and reducing cost and schedule.

  3. Office of River Protection Advanced Low-Activity Waste Glass Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, A. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Peeler, D. K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kim, D. S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vienna, J. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Piepel, G. F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schweiger, M. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-11-23

    The U.S. Department of Energy Office of River Protection (ORP) has initiated and leads an integrated Advanced Waste Glass (AWG) program to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product performance requirements. The integrated ORP program is focused on providing a technical, science-based foundation for making key decisions regarding the successful operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) facilities in the context of an optimized River Protection Project (RPP) flowsheet. The fundamental data stemming from this program will support development of advanced glass formulations, key product performance and process control models, and tactical processing strategies to ensure safe and successful operations for both the low-activity waste (LAW) and high-level waste vitrification facilities. These activities will be conducted with the objective of improving the overall RPP mission by enhancing flexibility and reducing cost and schedule.

  4. Answers to your questions on high-level nuclear waste

    International Nuclear Information System (INIS)

    1987-11-01

    This booklet contains answers to frequently asked questions about high-level nuclear wastes. Written for the layperson, the document contains basic information on the hazards of radiation, the Nuclear Waste Management Program, the proposed geologic repository, the proposed monitored retrievable storage facility, risk assessment, and public participation in the program

  5. Long-term management of wastes resulting from dismantling operations. Storing the very low-level activity wastes at Morvilliers

    International Nuclear Information System (INIS)

    Duret, F.; Dutzer, M.; Beranger, V.; Lecoq, P.

    2003-01-01

    Extension of dismantling operations in France in the years to come poses the question of availability of long-term waste facility. Large amount of such wastes will be produced after progressive shutdown of the 58 pressurized water reactors now in operation, not before 2010. However, France is already confronted with dismantling of 9 power reactors (6 of which of gas cooled graphite type), the first reprocessing plant at Marcoule, as well as, dismantling of other installations, for instance the CEA reactors or laboratories. The systems of processing the dismantling waste are not different from those used for wastes resulting from nuclear operations. For the high-level or long-term intermediate level activity disposal the debates must start by 2006, as based on the results of the research conducted according to different provisions of the December 30, 1991 law. These wastes represent however small amounts from the dismantling (around 2000 t for the 9 reactors at shutdown) and they will be stored until a decision will be made. A specific storing system should be implemented by 2008-2010 for the graphite wastes (around 23,000 t) which contain significant amount of long-lived radioelements, although their gross activity is low. But the most significant amount will come from low-level or intermediate-level of short lifetime or from wastes of very low activity. The first category is stored at Storage Center at Aube (CSA), its capacity being of 1,000,000 m 3 of drums. The total volume stored by the end of 2002 amounted 136,500 m 3 with an annual delivering of 12-15,000 m 3 at design rate of 30,000 m 3 /y. This center will be able to absorb the flux increase resulting from dismantling of the decommissioned nuclear installations (around 50,000 t from the dismantling of the 9 power reactor). The Center at Aube can be also adapted for storing wastes of large sizes as for instance the lid of the reactor vessel. According to the French regulation, the wastes produced within a

  6. Material chemistry challenges in vitrification of high level radioactive waste

    International Nuclear Information System (INIS)

    Kaushik, C.P.

    2008-01-01

    Full text: Nuclear technology with an affective environmental management plan and focused attention on safety measures is a much cleaner source of electricity generation as compared to other sources. With this perspective, India has undertaken nuclear energy program to share substantial part of future need of power. Safe containment and isolation of nuclear waste from human environment is an indispensable part of this programme. Majority of radioactivity in the entire nuclear fuel cycle is high level radioactive liquid waste (HLW), which is getting generated during reprocessing of spent nuclear fuels. A three stage strategy for management of HLW has been adopted in India. This involves (i) immobilization of waste oxides in stable and inert solid matrices, (ii) interim retrievable storage of the conditioned waste product under continuous cooling and (iii) disposal in deep geological formation. Borosilicate glass matrix has been adopted in India for immobilization of HLW. Material issue are very important during the entire process of waste immobilization. Performance of the materials used in nuclear waste management determines its safety/hazards. Material chemistry therefore has a significant bearing on immobilization science and its technological development for management of HLW. The choice of suitable waste form to deploy for nuclear waste immobilization is difficult decision and the durability of the conditioned product is not the sole criterion. In any immobilization process, where radioactive materials are involved, the process and operational conditions play an important role in final selection of a suitable glass formulation. In remotely operated vitrification process, study of chemistry of materials like glass, melter, materials of construction of other equipment under high temperature and hostile corrosive condition assume significance for safe and un-interrupted vitrification of radioactive to ensure its isolation waste from human environment. The present

  7. Neotectonic movement feature in preselection area for high level radioactive waste repository

    International Nuclear Information System (INIS)

    Huang Xianfang; Gao Yang; He Jianguo; Li Jianzhong; Gao Honglei; Xu Guoqing

    2010-01-01

    Neotectonic activity intensity is an important criteria for evaluating high level radioactive waste repository. The guiding ideology, methods and application of neotectonic study are elaborated in the paper. According to comparison research between the south and north part of east Tianshan area, the south part of east Tianshan is regarded as relative stable or relative weak in neotectonic movement in Neogene period and was selected as preselection area for high level radioactive waste repository. (authors)

  8. Application of new technologies for characterization of Hanford Site high-level waste

    International Nuclear Information System (INIS)

    Winters, W.I.

    1998-01-01

    To support remediation of Hanford Site high-level radioactive waste tanks, new chemical and physical measurement technologies must be developed and deployed. This is a major task of the Chemistry Analysis Technology Support (CATS) group of the Hanford Corporation. New measurement methods are required for efficient and economical resolution of tank waste safety, waste retrieval, and disposal issues. These development and deployment activities are performed in cooperation with Waste Management Federal Services of Hanford, Inc. This paper provides an overview of current analytical technologies in progress. The high-level waste at the Hanford Site is chemically complex because of the numerous processes used in past nuclear fuel reprocessing there, and a variety of technologies is required for effective characterization. Programmatic and laboratory operational needs drive the selection of new technologies for characterizing Hanford Site high-level waste, and these technologies are developed for deployment in laboratories, hot cells or in the field. New physical methods, such as the propagating reactive systems screening tool (PRSST) to measure the potential for self-propagating reactions in stored wastes, are being implemented. Technology for sampling and measuring gases trapped within the waste matrix is being used to evaluate flammability hazards associated with gas releases from stored wastes. Application of new inductively coupled plasma and laser ablation mass spectrometry systems at the Hanford Site's 222-S Laboratory will be described. A Raman spectroscopy probe mounted in a cone penetrometer to measure oxyanions in wastes or soils will be described. The Hanford Site has used large volumes of organic complexants and acids in processing waste, and capillary zone electrophoresis (CZE) methods have been developed for determining several of the major organic components in complex waste tank matrices. The principles involved, system installation, and results from

  9. Hanford Waste Vitrification Plant Quality Assurance Program description for defense high-level waste form development and qualification

    International Nuclear Information System (INIS)

    Hand, R.L.

    1992-01-01

    This document describes the quality assurance (QA) program of the Hanford Waste Vitrification Plant (HWVP) Project. The purpose of the QA program is to control project activities in such a manner as to achieve the mission of the HWVP Project in a safe and reliable manner. A major aspect of the HWVP Project QA program is the control of activities that relate to high-level waste (HLW) form development and qualification. This document describes the program and planned actions the Westinghouse Hanford Company (Westinghouse Hanford) will implement to demonstrate and ensure that the HWVP Project meets the US Department of Energy (DOE) and ASME regulations. The actions for meeting the requirements of the Waste Acceptance Preliminary Specifications (WAPS) will be implemented under the HWVP product qualification program with the objective of ensuring that the HWVP and its processes comply with the WAPS established by the federal repository

  10. The 1 000-year prediction. A state-of-the-art review on the research activity for the structural integrity of geological disposal packages of high-level nuclear waste

    International Nuclear Information System (INIS)

    Akashi, Masatsune

    1996-01-01

    The geological disposal package for high-level nuclear waste to be buried deep underground must be assured of structural integrity for ultra-long services of 1 000 years or even longer. The greatest and essentially the sole adversary to those packages in such a service is corrosion by ground water. Therefore, quantitative assessment of the corrosion form, the corrosion rate, and the corrosion lifetime is indispensable. This paper reviews the research activities to clarify what has been known, and discusses the future items to be studied. The largest detriment to the integrity of the package is not the uniform corrosion but the localized corrosion. The critical potential concept can quantify the safety usage domain for the material concerned. (author)

  11. 2002 summary of the report on studies and activities. Research on the Disposal of High-level Long-lived Radioactive Wastes

    International Nuclear Information System (INIS)

    2004-01-01

    This synthesis presents the studies underway and the major results obtained in the framework of the second research area prescribed by the Law of 1991 (the feasibility study of a deep geological repository). It also clarifies the French and international contexts of that research, namely with regard to the development of scientific partnerships. Content: 1 - scientific knowledge gained in 2002: packages, materials evolution, works about the clayey environment, works about the granitic environment, modeling approach; 2 - the 2002 granite dossier: context and goals, content; 3 - the 2002 concept choices for the clayey environment: overall architecture, packages and storage cells for B-type wastes, packages and storage cells for C-type vitrified wastes, packages and storage cells for spent fuels; 4 - international context: IAEA, OECD/NEA, integrated group for the safety case (IGSC), forum for stakeholders confidence (FSC), working party on decommissioning and dismantling (WPDD), general direction of energy transport, general direction of research, status of partnerships and bilateral agreements, advance of research cooperations, European Union; 5 - ANDRA's participations to international meetings; 6 - scientific partnerships: Ecole des Mines, BRGM, GdR FORPRO, MoMas, CEA, IFP, PRACTIS, PARIS, 7 - information and evaluation: national commission of evaluation (CNE), scientific committee, permanent group, orientation and follow-up committee; 8 - ANDRA's theses; 9 - ANDRA's publications. (J.S.)

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

    International Nuclear Information System (INIS)

    McDowell, Kip; Forrester, Tim; Saunders, Mark

    2013-01-01

    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)

  13. Leaching of gold from a mechanically and mechanochemically activated waste

    Directory of Open Access Journals (Sweden)

    Jana Ficeriová

    2010-03-01

    Full Text Available The intensification of leaching of gold from a waste using mechanical activation (milling in water and mechanochemical activation(milling in thiourea solution were studied as the pretreatment steps. The leaching of “as-received“ sample in an acid thiourea solutionresulted in 78 % Au dissolution, after mechanical activation 98 % and mechanochemical activation up to 99 % of the gold was leachedduring 120 min. The mechanochemical activation resulted in an increase of the specific surface area of the waste from 0.6 m2g-1to a maximum value of 20.5 m2g-1. The activation was performed in an attritor using variable milling times. The physico-chemical changesin the waste as a consequence of mechanochemical activation had a pronounced influence on the subsequent gold extraction.

  14. Radioactive gaseous waste management activities at CWMF

    International Nuclear Information System (INIS)

    Sumangala, R.K; Cheralathan, M.; Hariharan, P.T.; Chitra, S.; Paul, B.

    2015-01-01

    HEPA and iodine filter banks are used as an important engineering safeguard to prevent the release of airborne activity to the environment during normal and accident conditions in all nuclear installations. CWMF is responsible for the periodical testing and certification these filter banks as per the technical specification of the nuclear facilities at Kalpakkam site. An efficiency of >99.9% is ensured for both the HEPA as well as iodine filter banks. The larger radioactive particulates are trapped in the micro glass fibre filter paper medium by the mechanism of interception and inertial impaction whereas particulates of submicron size are caught by diffusion. The major activity removed in particulate form is 137 Cs and 90 Sr. The elemental iodine is removed by physico-chemical adsorption on high surface area activated charcoal and organic compounds of iodine are removed by isotopic exchange with KI/KOH impregnated activated charcoal or silver impregnated silica gel. Silver impregnated molecular sieves 13-X and AR-1 were developed for the removal of iodine from reprocessing atmosphere. Studies on pressure swing adsorption technique have been carried out for isolating Argon from air. Using Molecular sieve 5A (45psi-50psi) and Carbon molecular sieves (100psi to 120psi) based PSA systems in series an enrichment of 30% Ar is possible. Pressurized Heavy Water Reactors with calandria vault coolant as air produces 41 Ar due to neutron activation of 40 Ar present in air which main contributor to the air is borne activity in MAPS, RAPS and Dhruva reactors. The isolated argon can be stored for decay and activity release can be minimized as per ALARA principle. (author)

  15. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    Pettit, N. E.

    2001-01-01

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms [IPWF]) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. US Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as co-disposal. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister inserted in the center and/or one or more DOE SNF canisters displacing a HLW canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by

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

    Energy Technology Data Exchange (ETDEWEB)

    G. Radulesscu; J.S. Tang

    2000-06-07

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

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

    International Nuclear Information System (INIS)

    Radulesscu, G.; Tang, J.S.

    2000-01-01

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

  18. LLNL/YMP Waste Container Fabrication and Closure Project; GFY technical activity summary

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-10-01

    The Department of Energy`s Office of Civilian Radioactive Waste Management (OCRWM) Program is studying Yucca Mountain, Nevada as a suitable site for the first US high-level nuclear waste repository. Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing and developing the waste package for the permanent storage of high-level nuclear waste. This report is a summary of the technical activities for the LLNL/YMP Nuclear Waste Disposal Container Fabrication and Closure Development Project. Candidate welding closure processes were identified in the Phase 1 report. This report discusses Phase 2. Phase 2 of this effort involved laboratory studies to determine the optimum fabrication and closure processes. Because of budget limitations, LLNL narrowed the materials for evaluation in Phase 2 from the original six to four: Alloy 825, CDA 715, CDA 102 (or CDA 122) and CDA 952. Phase 2 studies focused on evaluation of candidate material in conjunction with fabrication and closure processes.

  19. Removing radio-active wastes from nuclear power stations by the STEAG system

    International Nuclear Information System (INIS)

    Baatz, H.

    1978-01-01

    The mobile STEAG System for conditioning radio-active wastes from nuclear power stations represents a particularly safe and economic method of removing them in present day conditions. Cementation by the FAFNIR System is used for the greater part of the waste, the liquid concentrate (evaporator concentrate and filter slurry). For the special case of the medium active resin balls from the primary circuits, embedding in plastic by the FAMA process has proved to be the only available successful process so far. The highly active solid waste from the reactor core is decomposed by the MOSAIK System, is packed in transportable and storable containers and is removed from the fuel element storage pond. The systems are so safe that faults or interruptions of power station operation due to faults in removing radio-active wastes can be excluded. (orig.) [de

  20. Activation and characterization of waste coffee grounds as bio-sorbent

    Science.gov (United States)

    Mariana; Marwan; Mulana, F.; Yunardi; Ismail, T. A.; Hafdiansyah, M. F.

    2018-03-01

    As the city well known for its culture of coffee drinkers, modern and traditional coffee shops are found everywhere in Banda Aceh, Indonesia. High number of coffee shops in the city generates large quantities of spent coffee grounds as waste without any effort to convert them as other valuable products. In an attempt to reduce environmental problems caused by used coffee grounds, this research was conducted to utilize waste coffee grounds as an activated carbon bio-sorbent. The specific purpose of this research is to improve the performance of coffee grounds bio-sorbent through chemical and physical activation, and to characterize the produced bio-sorbent. Following physical activation by carbonization, a chemical activation was achieved by soaking the carbonized waste coffee grounds in HCl solvent and carbonization process. The activated bio-sorbent was characterized for its morphological properties using Scanning Electron Microscopy (SEM), its functional groups by Fourier Transform Infra-Red Spectrophotometer (FTIR), and its material characteristics using X-Ray Diffraction (XRD). Characterization of the activated carbon prepared from waste coffee grounds shows that it meets standard quality requirement in accordance with Indonesian National Standard, SNI 06-3730-1995. Activation process has modified the functional groups of the waste coffee grounds. Comparing to natural waste coffee grounds, the resulted bio-sorbent demonstrated a more porous surface morphology following activation process. Consequently, such bio-sorbent is a potential source to be used as an adsorbent for various applications.

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

    Energy Technology Data Exchange (ETDEWEB)

    Troyer, G.L.

    1997-03-17

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

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

    International Nuclear Information System (INIS)

    Troyer, G.L.

    1997-01-01

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

  3. National high-level waste systems analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Kristofferson, K.; Oholleran, T.P.; Powell, R.H.

    1995-09-01

    This report documents the assessment of budgetary impacts, constraints, and repository availability on the storage and treatment of high-level waste and on both existing and pending negotiated milestones. The impacts of the availabilities of various treatment systems on schedule and throughput at four Department of Energy sites are compared to repository readiness in order to determine the prudent application of resources. The information modeled for each of these sites is integrated with a single national model. The report suggests a high-level-waste model that offers a national perspective on all high-level waste treatment and storage systems managed by the Department of Energy.

  4. National high-level waste systems analysis report

    International Nuclear Information System (INIS)

    Kristofferson, K.; Oholleran, T.P.; Powell, R.H.

    1995-09-01

    This report documents the assessment of budgetary impacts, constraints, and repository availability on the storage and treatment of high-level waste and on both existing and pending negotiated milestones. The impacts of the availabilities of various treatment systems on schedule and throughput at four Department of Energy sites are compared to repository readiness in order to determine the prudent application of resources. The information modeled for each of these sites is integrated with a single national model. The report suggests a high-level-waste model that offers a national perspective on all high-level waste treatment and storage systems managed by the Department of Energy

  5. Design concepts of definitive disposal for high level radioactive wastes

    International Nuclear Information System (INIS)

    Badillo A, V.E.; Alonso V, G.

    2007-01-01

    It is excessively known the importance about finding a solution for the handling and disposition of radioactive waste of all level. However, the polemic is centered in the administration of high level radioactive waste and the worn out fuel, forgetting that the more important volumes of waste its are generated in the categories of low level wastes or of very low level. Depending on the waste that will be confined and of the costs, several technological modalities of definitive disposition exist, in function of the depth of the confinement. The concept of deep geologic storage, technological option proposed more than 40 years ago, it is a concept of isolation of waste of long half life placed in a deep underground installation dug in geologic formations that are characterized by their high stability and their low flow of underground water. In the last decades, they have registered countless progresses in technical and scientific aspects of the geologic storage, making it a reliable technical solution supported with many years of scientific work carried out by numerous institutions in the entire world. In this work the design concepts that apply some countries for the high level waste disposal that its liberate heat are revised and the different geologic formations that have been considered for the storage of this type of wastes. (Author)

  6. Disposal of high level and intermediate level radioactive wastes

    International Nuclear Information System (INIS)

    Flowers, R.H.

    1991-01-01

    The waste products from the nuclear industry are relatively small in volume. Apart from a few minor gaseous and liquid waste streams, containing readily dispersible elements of low radiotoxicity, all these products are processed into stable solid packages for disposal in underground repositories. Because the volumes are small, and because radioactive wastes are latecomers on the industrial scene, a whole new industry with a world-wide technological infrastructure has grown up alongside the nuclear power industry to carry out the waste processing and disposal to very high standards. Some of the technical approaches used, and the Regulatory controls which have been developed, will undoubtedly find application in the future to the management of non-radioactive toxic wastes. The repository site outlined would contain even high-level radioactive wastes and spent fuels being contained without significant radiation dose rates to the public. Water pathway dose rates are likely to be lowest for vitrified high-level wastes with spent PWR fuel and intermediate level wastes being somewhat higher. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-01-04

    This report describes the development and testing of new glass formulations for high aluminum waste streams that achieve high waste loadings while maintaining high processing rates. The testing was based on the compositions of Hanford High Level Waste (HLW) with limiting concentrations of aluminum specified by the Office of River Protection (ORP). The testing identified glass formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts and small scale melt rate screening tests. The results were used to select compositions for subsequent testing in a DuraMelter 100 (DM100) system. These tests were used to determine processing rates for the selected formulations as well as to examine the effects of increased glass processing temperature, and the form of aluminum in the waste simulant. Finally, one of the formulations was selected for large-scale confirmatory testing on the HLW Pilot Melter (DM1200), which is a one third scale prototype of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW melter and off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy (DOE) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same high-aluminum waste composition used in the present work and other Hanford HLW compositions. The scope of this study was outlined in a Test Plan that was prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the WTP is about 13,500 (equivalent to 40,500 MT glass). This estimate is based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form

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

    International Nuclear Information System (INIS)

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

    2010-01-01

    This report describes the development and testing of new glass formulations for high aluminum waste streams that achieve high waste loadings while maintaining high processing rates. The testing was based on the compositions of Hanford High Level Waste (HLW) with limiting concentrations of aluminum specified by the Office of River Protection (ORP). The testing identified glass formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts and small scale melt rate screening tests. The results were used to select compositions for subsequent testing in a DuraMelter 100 (DM100) system. These tests were used to determine processing rates for the selected formulations as well as to examine the effects of increased glass processing temperature, and the form of aluminum in the waste simulant. Finally, one of the formulations was selected for large-scale confirmatory testing on the HLW Pilot Melter (DM1200), which is a one third scale prototype of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW melter and off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy (DOE) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same high-aluminum waste composition used in the present work and other Hanford HLW compositions. The scope of this study was outlined in a Test Plan that was prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the WTP is about 13,500 (equivalent to 40,500 MT glass). This estimate is based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form

  9. Remote ignitability analysis of high-level radioactive waste

    International Nuclear Information System (INIS)

    Lundholm, C.W.; Morgan, J.M.; Shurtliff, R.M.; Trejo, L.E.

    1992-09-01

    The Idaho Chemical Processing Plant (ICPP), was used to reprocess nuclear fuel from government owned reactors to recover the unused uranium-235. These processes generated highly radioactive liquid wastes which are stored in large underground tanks prior to being calcined into a granular solid. The Resource Conservation and Recovery Act (RCRA) and state/federal clean air statutes require waste characterization of these high level radioactive wastes for regulatory permitting and waste treatment purposes. The determination of the characteristic of ignitability is part of the required analyses prior to calcination and waste treatment. To perform this analysis in a radiologically safe manner, a remoted instrument was needed. The remote ignitability Method and Instrument will meet the 60 deg. C. requirement as prescribed for the ignitability in method 1020 of SW-846. The method for remote use will be equivalent to method 1020 of SW-846

  10. Disposal of Kitchen Waste from High Rise Apartment

    Science.gov (United States)

    Ori, Kirki; Bharti, Ajay; Kumar, Sunil

    2017-09-01

    The high rise building has numbers of floor and rooms having variety of users or tenants for residential purposes. The huge quantities of heterogenous mixtures of domestic food waste are generated from every floor of the high rise residential buildings. Disposal of wet and biodegradable domestic kitchen waste from high rise buildings are more expensive in regards of collection and vertical transportation. This work is intended to address the technique to dispose of the wet organic food waste from the high rise buildings or multistory building at generation point with the advantage of gravity and vermicomposting technique. This innovative effort for collection and disposal of wet organic solid waste from high rise apartment is more economical and hygienic in comparison with present system of disposal.

  11. Activity measurements at a waste volume reduction facility

    International Nuclear Information System (INIS)

    Richardson, J.; Lee, D.A.

    1979-01-01

    The monitoring program for Ontario Hydro's radioactive waste management site will be described, several aspects of which will be discussed in detail. The program at this facility includes categorization, volume reduction processing, and storage of solid radioactive wastes from nuclear generating stations of the CANDU type. At the present time, two types of volume reduction process are in operation - incineration and compaction. Following categorization and processing, wastes are stored in in-ground concrete trenches or tile-holes, or in above-ground quadricells. The monitoring program is divided into three areas: public safety, worker safety, and structural integrity. Development projects with respect to the monitoring program have been undertaken to achieve activity accounting for the total waste management program. In particular, a field measurement for the radioactivity content of radioactive ash containers and compacted waste drums

  12. Advanced waste form and melter development for treatment of troublesome high-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-02

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these "troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approached to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.

  13. Midwestern High-Level Radioactive Waste Transportation Project

    International Nuclear Information System (INIS)

    1993-01-01

    On February 17,1989, the Midwestern Office of The Council of State Governments and the US Department of Energy entered into a cooperative agreement authorizing the initiation of the Midwestern High-Level Radioactive Waste Transportation Project. The transportation project continued to receive funding from DOE through amendments to the original cooperative agreement, with December 31, 1993, marking the end of the initial 5-year period. This progress report reflects the work completed by the Midwestern Office from February 17,1989, through December 31,1993. In accordance with the scopes of work governing the period covered by this report, the Midwestern Office of The Council of State Governments has worked closely with the Midwestern High-Level Radioactive Waste Committee. Project staff have facilitated all eight of the committee's meetings and have represented the committee at meetings of DOE's Transportation Coordination Group (TCG) and Transportation External Coordination Working Group (TEC/WG). Staff have also prepared and submitted comments on DOE activities on behalf of the committee. In addition to working with the committee, project staff have prepared and distributed 20 reports, including some revised reports (see Attachment 1). Staff have also developed a library of reference materials for the benefit of committee members, state officials, and other interested parties. To publicize the library, and to make it more accessible to potential users, project staff have prepared and distributed regular notices of resource availability

  14. A Strategy for Maintenance of the Long-Term Performance Assessment of Immobilized Low-Activity Waste Glass

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, Joseph V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Freedman, Vicky L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-28

    Approximately 50 million gallons of high-level radioactive mixed waste has accumulated in 177 buried single- and double-shell tanks at the Hanford Site in southeastern Washington State as a result of the past production of nuclear materials, primarily for defense uses. The United States Department of Energy (DOE) is proceeding with plans to permanently dispose of this waste. Plans call for separating the tank waste into high-level waste (HLW) and low-activity waste (LAW) fractions, which will be vitrified at the Hanford Waste Treatment and Immobilization Plant (WTP). Principal radionuclides of concern in LAW are 99Tc, 129I, and U, while non-radioactive contaminants of concern are Cr and nitrate/nitrite. HLW glass will be sent off-site to an undetermined federal site for deep geological disposal while the much larger volume of immobilized low-activity waste will be placed in the on-site, near-surface Integrated Disposal Facility (IDF).

  15. High level waste management in Asia: R and D perspectives

    International Nuclear Information System (INIS)

    Deokattey, Sangeeta; Bhanumurthy, K.

    2010-01-01

    The present work is an attempt to provide an overview, about the status of R and D and current trends in high level radioactive waste management, particularly in Asian countries. The INIS database (for the period 1976 to 2010) was selected for this purpose, as this is the most authoritative global source of information, in the area of Nuclear Science and Technology. Appropriate query formulations on the database, resulted in the retrieval of 4322 unique bibliographic records. Using the content analysis method (which is both a qualitative as well as a quantitative research method), all the records were analyzed. Part One of the analysis details Scientometric R and D indicators, such as the countries and the institutions involved in R and D, the types of publications, and programmes and projects related to High Level Waste management. Part Two is a subject-based analysis, grouped under the following broad categories: I. Waste Processing 1. Partitioning and transmutation (including ADS) II. Waste Immobilization 1. Glass waste forms and 2. Crystalline ceramics and other waste forms III. Waste Disposal 1. Performance assessment and safety evaluation studies 2. Geohydrological studies a. Site selection and characterization, b. In situ underground experiments, c. Rock mechanical characterization 3. Deep geological repositories a. Sorption, migration and groundwater chemistry b. Engineered barrier systems and IV. Waste Packaging Materials. The results of this analysis are summarized in the study. (author)

  16. Overview of high-level waste management accomplishments

    International Nuclear Information System (INIS)

    Lawroski, H.; Berreth, J.R.; Freeby, W.A.

    1980-01-01

    Storage of power reactor spent fuel is necessary at present because of the lack of reprocessing operations particularly in the U.S. By considering the above solidification and storage scenario, there is more than reasonable assurance that acceptable, stable, low heat generation rate, solidified waste can be produced, and safely disposed. The public perception of no waste disposal solutions is being exploited by detractors of nuclear power application. The inability to even point to one overall system demonstration lends credibility to the negative assertions. By delaying the gathering of on-line information to qualify repository sites, and to implement a demonstration, the actions of the nuclear power detractors are self serving in that they can continue to point out there is no demonstration of satisfactory high-level waste disposal. By maintaining the liquid and solidified high-level waste in secure above ground storage until acceptable decay heat generation rates are achieved, by producing a compatible, high integrity, solid waste form, by providing a second or even third barrier as a compound container and by inserting the enclosed waste form in a qualified repository with spacing to assure moderately low temperature disposal conditions, there appears to be no technical reason for not progressing further with the disposal of high-level wastes and needed implementation of the complete nuclear power fuel cycle

  17. Technical career opportunities in high-level radioactive waste management

    International Nuclear Information System (INIS)

    1993-01-01

    Technical career opportunities in high-level radioactive waste management are briefly described in the areas of: Hydrology; geology; biological sciences; mathematics; engineering; heavy equipment operation; and skilled labor and crafts

  18. Heat exchangers and recuperators for high temperature waste gases

    Science.gov (United States)

    Meunier, H.

    General considerations on high temperature waste heat recovery are presented. Internal heat recovery through combustion air preheating and external heat recovery are addressed. Heat transfer and pressure drop in heat exchanger design are discussed.

  19. Development of technical information database for high level waste disposal

    International Nuclear Information System (INIS)

    Kudo, Koji; Takada, Susumu; Kawanishi, Motoi

    2005-01-01

    A concept design of the high level waste disposal information database and the disposal technologies information database are explained. The high level waste disposal information database contains information on technologies, waste, management and rules, R and D, each step of disposal site selection, characteristics of sites, demonstration of disposal technology, design of disposal site, application for disposal permit, construction of disposal site, operation and closing. Construction of the disposal technologies information system and the geological disposal technologies information system is described. The screen image of the geological disposal technologies information system is shown. User is able to search the full text retrieval and attribute retrieval in the image. (S.Y. )

  20. Biogas from organically high polluted industrial waste waters

    Energy Technology Data Exchange (ETDEWEB)

    Sixt, H

    1985-06-01

    Organically high polluted waste water sets special claims for an economical purification and the process treatment. Up to now these waste waters are being purified by anaerobic processes with simultaneous biogas generation. The fourstep anaerobic degradation is influenced by a lot of important parameters. Extensive researchers in the field of anaerobic microbiology has improved the knowledge of the fundamental principles. Parallel the reactor technology is developed worldwide. In general it seems that the fixed-film-reactor with immobilized bacteria has the best future to purify organically high polluted industrial waste water with short retention times under stable operation conditions.

  1. Treatment technologies for non-high-level wastes (USA)

    International Nuclear Information System (INIS)

    Cooley, C.R.; Clark, D.E.

    1976-06-01

    Non-high-level waste arising from operations at nuclear reactors, fuel fabrication facilities, and reprocessing facilities can be treated using one of several technical alternatives prior to storage. Each alternative and the associated experience and status of development are summarized. The technology for treating non-high-level wastes is generally available for industrial use. Improved techniques applicable to the commercial nuclear fuel cycle are being developed and demonstrated to reduce the volume of waste and to immobilize it for storage. 36 figures, 59 references

  2. High-level radioactive waste disposal type and theoretical analyses

    International Nuclear Information System (INIS)

    Lu Yingfa; Wu Yanchun; Luo Xianqi; Cui Yujun

    2006-01-01

    Study of high-level radioactive waste disposal is necessary for the nuclear electrical development; the determination of nuclear waste depository type is one of importance safety. Based on the high-level radioactive disposal type, the relative research subjects are proposed, then the fundamental research characteristics of nuclear waste disposition, for instance: mechanical and hydraulic properties of rock mass, saturated and unsaturated seepage, chemical behaviors, behavior of special soil, and gas behavior, etc. are introduced, the relative coupling equations are suggested, and a one dimensional result is proposed. (authors)

  3. Developments in the management of low and intermediate activity solid wastes at the Cadarache Centre

    International Nuclear Information System (INIS)

    Barbreau, A.; Marcaillou, J.; Mery, J.; Pinto, D.; Rancon, D.

    1975-01-01

    The Cadarache Nuclear Studies Centre is located in a thinly populated region. Covering a total area of 1600 hectares, it has been able to accommodate numerous and important research facilities. In 1970, 11 reactors or critical assemblies were in operation. More than 164000 m 2 are devoted to laboratories, testing areas, installations for the inspection of irradiated fuel elements and plutonium technology workshops. Up to 1968 the low- and intermediate-activity solid wastes (categories 1, 2 and 30) collected at the Centre were divided into two classes for disposal purposes: (a) burnable wastes which, after sorting, were destroyed in an incinerator; (b) compressible wastes which were compacted in concrete containers after recovery of the packing, by means of a 250-ton press. The situation at Cadarache and the results obtained in hydrogeological studies have prompted the Centre to improve the processing of these wastes and reduce the cost. The treatment of solid wastes should, in effect, be regarded as a step towards their final elimination. The measure envisaged at Cadarache were thus aimed at permitting final storage on site, in order to reduce the volume of waste, contain the activity and keep the cost to a minimum. The management of solid wastes is at present based on the following methods: (a) storage in trenches with PVC packing for non-burnable solid wastes of categories 1 and 4, after monitoring of specific activities; (b) compacting and storage in leak-proof pools for solid wastes of categories 2 and 3, the most highly active undergoing a period of decay storage beforehand; (c) incineration of burnable solid wastes of categories 1 and 2 and also of contaminated oils and solvents. (author)

  4. Treatment of low alpha activity liquid wastes

    International Nuclear Information System (INIS)

    Nannicini, R.; Fenoglio, F.; Pozzi, L.

    1984-01-01

    The nuclear industry considers so big safety problems that the purifying treatment of liquid wastes must always provide for a complete recycle of the liquid strems from the production processes as regard this problem. ''Enea-Comb-Ifec'' people from saluggia, already previously engages with verifying and setting-up ''Sol-Gel'' process for the recover of uranium-plutonium solutions coming from irradiated fuel reprocessing, started an experimental work, with the assistance of ''Cnr-Irsa'' from Rome, on the applicability of the biological treatment to the purification of liquid wastes coming from the production process itself. The present technical report gives, besides a short description of the ''Sol-Gel'' process, the first results, only relating to the biological stage of the whole proposed purifyng treatment, included the final results of the experimental work, object of a contract between ''Enea-Ifec'' and ''Snam progetti'' from Fano

  5. Segregation of low-level dry active waste

    International Nuclear Information System (INIS)

    Kornblith, L. Jr.; Naughton, M.D.; Welsh, L.

    1984-01-01

    A program has been carried out to characterize the Dry Active Waste (DAW) stream from a typical PWR power plant in order to determine the usefulness of large-volume DAW monitors for segregating such waste in order to dispose of it in appropriate facilities. A waste monitor using plastic scintillation counters was used for measuring the waste. The monitor had a volume of about 300 liters and an overall efficiency of about 12% for a typical fission product mixture. It provides automatic compensation for background radioactivity and can measure a bag of waste in less than a minute, including background measurements. Six hundred consecutively generated bags of DAW were measured. These had a total activity of about one millicurie and an average specific activity of about 540 nanocuries per kilogram. About half of the bags contained less than 1000 nanocuries and had specific activities of less than 100 nanocuries per kilogram. Based on simplified preliminary calculations, it appears that an evaluation of the risks of disposal of bags such as these in a landfill other than a low-level waste disposal facility could be carried out that would demonstrate that such disposal of half or more of these bags would not result in any substantial hazard, either short or long term

  6. Storage drums for radio-active waste

    International Nuclear Information System (INIS)

    Knights, H.C.

    1982-01-01

    The lid of a storage drum for radioactive waste is secured by a series of clamps each of which has a hook for engaging the rim of the drum. Each clamp has an indicating means whereby a remote operator can check that the lid is secured to the drum. In a second embodiment, the position of an arm acts as a visual indication as to whether or not the clamp is in engagement with the container rim. (author)

  7. Minor component study for simulated high-level nuclear waste glasses (Draft)

    International Nuclear Information System (INIS)

    Li, H.; Langowskim, M.H.; Hrma, P.R.; Schweiger, M.J.; Vienna, J.D.; Smith, D.E.

    1996-02-01

    Hanford Site single-shell tank (SSI) and double-shell tank (DSI) wastes are planned to be separated into low activity (or low-level waste, LLW) and high activity (or high-level waste, HLW) fractions, and to be vitrified for disposal. Formulation of HLW glass must comply with glass processibility and durability requirements, including constraints on melt viscosity, electrical conductivity, liquidus temperature, tendency for phase segregation on the molten glass surface, and chemical durability of the final waste form. A wide variety of HLW compositions are expected to be vitrified. In addition these wastes will likely vary in composition from current estimates. High concentrations of certain troublesome components, such as sulfate, phosphate, and chrome, raise concerns about their potential hinderance to the waste vitrification process. For example, phosphate segregation in the cold cap (the layer of feed on top of the glass melt) in a Joule-heated melter may inhibit the melting process (Bunnell, 1988). This has been reported during a pilot-scale ceramic melter run, PSCM-19, (Perez, 1985). Molten salt segregation of either sulfate or chromate is also hazardous to the waste vitrification process. Excessive (Cr, Fe, Mn, Ni) spinel crystal formation in molten glass can also be detrimental to melter operation

  8. Anaerobic co-digestion of winery waste and waste activated sludge: assessment of process feasibility.

    Science.gov (United States)

    Da Ros, C; Cavinato, C; Cecchi, F; Bolzonella, D

    2014-01-01

    In this study the anaerobic co-digestion of wine lees together with waste activated sludge in mesophilic and thermophilic conditions was tested at pilot scale. Three organic loading rates (OLRs 2.8, 3.3 and 4.5 kgCOD/m(3)d) and hydraulic retention times (HRTs 21, 19 and 16 days) were applied to the reactors, in order to evaluate the best operational conditions for the maximization of the biogas yields. The addition of lee to sludge determined a higher biogas production: the best yield obtained was 0.40 Nm(3)biogas/kgCODfed. Because of the high presence of soluble chemical oxygen demand (COD) and polyphenols in wine lees, the best results in terms of yields and process stability were obtained when applying the lowest of the three organic loading rates tested together with mesophilic conditions.

  9. Development of melt compositions for sulphate bearing high level waste

    International Nuclear Information System (INIS)

    Jahagirdar, P.B.; Wattal, P.K.

    1997-09-01

    The report deals with the development and characterization of vitreous matrices for sulphate bearing high level waste. Studies were conducted in sodium borosilicate and lead borosilicate systems with the introduction of CaO, BaO, MgO etc. Lead borosilicate system was found to be compatible with sulphate bearing high level wastes. Detailed product evaluation carried on selected formulations is also described. (author)

  10. High-Level Waste System Process Interface Description

    International Nuclear Information System (INIS)

    D'Entremont, P.D.

    1999-01-01

    The High-Level Waste System is a set of six different processes interconnected by pipelines. These processes function as one large treatment plant that receives, stores, and treats high-level wastes from various generators at SRS and converts them into forms suitable for final disposal. The three major forms are borosilicate glass, which will be eventually disposed of in a Federal Repository, Saltstone to be buried on site, and treated water effluent that is released to the environment

  11. Deep reversible storage. Safety options for the storage in deep geological formation - High-medium activity, long living wastes 2009 milestone

    International Nuclear Information System (INIS)

    2010-09-01

    This large document aims at presenting safety options which have been adopted for the current design status (notably for the installation architecture), elements of description of envisaged technical solutions and exploitation principles which are required for the control of risks (either internal or external) and uncertainties on a long term which could lead to radiological consequences for the project of storage of nuclear wastes in a deep geological formation. After a presentation of the context and of input data, this report discusses the principle of a modular construction and then discusses the safety approach. One part deals with risk analysis for surface installations and aims at showing how internal risks (handling, fire) and external risks (earthquake, plane crash) are taken into account in terms of design choices, processes and control measures. Another part deals with risk analysis for underground installations during the reversible exploitation phase (the considered risks are about the same as in the previous part). The next part addresses risk analysis after closing, and tries to describe how the location, storage construction elements and its architecture ensure a passive safety. Uncertainty management is presented in relationship with envisaged technical solutions and scientific knowledge advances. Additional elements (detailed study, researches and experimentations) for the establishment of the future creation authorization request are identified all along the report

  12. Log live high activity radioactive wastes / Researches and results law of the 30 December 1991. Separation and transmutation of long lived radionuclides; Les dechets radioactifs a haute activite et a vie longue / recherches et resultats Loi du 30 decembre 1991. Separation et transmutation des radionucleides a vie longue

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-12-15

    The law of the 30 December 1991 on the high activity long lived radioactive wastes reached the end. This synthesis final document presents the scientific and technological results, obtained still the end of 2005, on the separation and the transmutation of long lived radionuclides of high activity long lived radioactive wastes. It is organized in five chapters: a presentation of the context and the historical aspects, the researches, the objectives and the strategy of the axis 1, the researches results on the advanced separation, the researches results on the transmutation, the scenario of separation-transmutation and their environmental, technical and economical impacts. (A.L.B.)

  13. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    2000-01-01

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different materials

  14. Vitrification of high-level radioactive and hazardous wastes

    International Nuclear Information System (INIS)

    Lutze, W.

    1993-12-01

    The main objective is to summarize work conducted on glasses as waste forms for high-level radioactive fission product solutions up to the late 1980's (section I and II). Section III addresses the question, whether waste forms designed for the immobilization of radioactive residues can be used for the same purpose for hazardous wastes. Of particular interest are those types of hazardous wastes, e.g., fly ashes from municipal combustion plants, easy to convert into glasses or ceramic materials. A large number of base glass compositions has been studied to vitrify waste from reprocessing but only borosilicate glasses with melting temperatures between 1100 C and 1200 C and very good hydrolytic stability is used today. (orig./HP) [de

  15. Phase 1 immobilized low-activity waste operational source term

    International Nuclear Information System (INIS)

    Burbank, D.A.

    1998-01-01

    This report presents an engineering analysis of the Phase 1 privatization feeds to establish an operational source term for storage and disposal of immobilized low-activity waste packages at the Hanford Site. The source term information is needed to establish a preliminary estimate of the numbers of remote-handled and contact-handled waste packages. A discussion of the uncertainties and their impact on the source term and waste package distribution is also presented. It should be noted that this study is concerned with operational impacts only. Source terms used for accident scenarios would differ due to alpha and beta radiation which were not significant in this study

  16. Incineration plant for low active waste at Inshass, LAWI

    International Nuclear Information System (INIS)

    Krug, W.; Thoene, L.; Schmitz, H.J.; Abdelrazek, I.D.

    1993-10-01

    The LAWI (Low Active Waste Incinerator) prototype incinerating plant was devised and constructed according to the principle of the Juelich thermoprocess and installed at the Egyptian research centre Inshass. In parallel, AEA Cairo devised and constructed their own operations building for this plant with all the features, infrastructural installations and rooms required for operating the plant and handling and treating low-level radioactive wastes. The dimensions of this incinerator were selected so as to be sufficient for the disposal of solid, weakly radioactive combustible wastes from the Inshass Research Centre and the environment (e.g. Cairo hospitals). (orig./DG) [de

  17. Criteria for high-level waste disposal

    International Nuclear Information System (INIS)

    Sousselier, Y.

    1981-01-01

    Disposal of radioactive wastes is storage without the intention of retrieval. But in such storage, it may be useful and in some cases necessary to have the possibility of retrieval at least for a certain period of time. In order to propose some criteria for HLW disposal, one has to examine how this basic concept is to be applied. HLW is waste separated as a raffinate in the first cycle of solvent extraction in reprocessing. Such waste contains the bulk of fission products which have long half lives, therefore the safety of a disposal site, at least after a certain period of time, must be intrinsic, i.e. not based on human intervention. There is a consensus that such a disposal is feasible in a suitable geological formation in which the integrity of the container will be reinforced by several additional barriers. Criteria for disposal can be proposed for all aspects of the question. The author discusses the aims of the safety analysis, particularly the length of time for this analysis, and the acceptable dose commitments resulting from the release of radionuclides, the number and role of each barrier, and a holistic analysis of safety external factors. (Auth.)

  18. Systems costs for disposal of Savannah River high-level waste sludge and salt

    International Nuclear Information System (INIS)

    McDonell, W.R.; Goodlett, C.B.

    1984-01-01

    A systems cost model has been developed to support disposal of defense high-level waste sludge and salt generated at the Savannah River Plant. Waste processing activities covered by the model include decontamination of the salt by a precipitation process in the waste storage tanks, incorporation of the sludge and radionuclides removed from the salt into glass in the Defense Waste Processing Facility (DWPF), and, after interim storage, final disposal of the DWPF glass waste canisters in a federal geologic repository. Total costs for processing of waste generated to the year 2000 are estimated to be about $2.9 billion (1984 dollars); incremental unit costs for DWPF and repository disposal activities range from $120,000 to $170,000 per canister depending on DWPF processing schedules. In a representative evaluation of process alternatives, the model is used to demonstrate cost effectiveness of adjustments in the frit content of the waste glass to reduce impacts of wastes generated by the salt decontamination operations. 13 references, 8 tables

  19. Probabilistic safety assessment for Hanford high-level waste tanks

    International Nuclear Information System (INIS)

    MacFarlane, D.R.; Stack, D.S.; Kindinger, J.P.; Deremer, R.K.

    1995-01-01

    This paper gives results from the first comprehensive level-3 probabilistic safety assessment (PSA), including consideration of external events, for the Hanford tank farm (HTF). This work was sponsored by the U.S. Department of Energy/Environmental Restoration and Waste Management Division (DOE/EM). At the HTF, there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/saltcake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is ∼60 million gal, containing ∼200 million Ci of radioactivity

  20. Test plan: Effects of phase separation on waste loading for high level waste glasses

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    2000-01-01

    As part of the Tanks Focus Area's (TFA) effort to increase waste loading for high-level waste (HLW) vitrification at various facilities in the Department of Energy (DOE) complex, the occurrence of phase separation in waste glasses spanning the Savannah River Site (SRS) and Idaho National Engineering and Environmental Laboratory (INEEL) composition ranges were studied during FY99. The type, extent, and impact of phase separation on glass durability for a series of HLW glasses, e.g., SRS-type and INEEL-type, were examined

  1. Quantitative analysis of impact of awareness-raising activities on organic solid waste separation behaviour in Balikpapan City, Indonesia.

    Science.gov (United States)

    Murase, Noriaki; Murayama, Takehiko; Nishikizawa, Shigeo; Sato, Yuriko

    2017-10-01

    Many cities in Indonesia are under pressure to reduce solid waste and dispose of it properly. In response to this pressure, the Japan International Cooperation Agency and the Indonesian Government have implemented a solid waste separation and collection project to reduce solid waste in the target area (810 households) of Balikpapan City. We used a cluster randomised controlled trial method to measure the impact of awareness-raising activities that were introduced by the project on residents' organic solid waste separation behaviour. The level of properly separated organic solid waste increased by 6.0% in areas that conducted awareness-raising activities. Meanwhile, the level decreased by 3.6% in areas that did not conduct similar activities. Therefore, in relative comparison, awareness-raising increased the level by 9.6%. A comparison among small communities in the target area confirmed that awareness-raising activities had a significant impact on organic solid waste separation. High frequencies of monitoring at waste stations and door-to-door visits by community members had a positive impact on organic solid waste separation. A correlation between the proximity of environmental volunteers' houses to waste stations and a high level of separation was also confirmed. The awareness-raising activities introduced by the project led to a significant increase in the separation of organic solid waste.

  2. Incineration of alpha-active solid waste by microwaves

    Energy Technology Data Exchange (ETDEWEB)

    Mallik, G K; Bhargava, V K; Kamath, H S; Purushotham, D S.C. [Bhabha Atomic Research Centre, Tarapur (India). Advanced Fuel Fabrication Facility

    1996-12-31

    The conventional techniques for treatment of alpha-active compressible solid waste involve incineration using electrically heated incinerators and subsequent recovery of special nuclear materials (SNM) from the ash by acid leaching. A microwave incineration followed by microwave digestion and SNM recovery from ash has specific advantages from maintenance and productivity consideration. The paper describes a preliminary work carried out with simulated uranium containing compressible solid waste using microwave heating technique. (author). 3 refs., 1 tab.

  3. Characterization of silicoaluminates for low and medium activity wastes packaging

    International Nuclear Information System (INIS)

    Rivoallan, A.; Berson, X.

    1996-01-01

    Studies are done in order to demonstrate many advantages (as an important volume reduction and a greater chemical stability) of packaging low and medium activity wastes in crystal structures compared with concrete and bitumen. In order to understand the consequences of hazardous chemical composition (especially anions) in the waste on the characteristics of the mineral packaging, a simulation study is developed with inactive concentrates. It leads to well crystallized structures which have not the same major crystallized phase. (authors)

  4. The defense waste processing facility: the final processing step for defense high-level waste disposal

    International Nuclear Information System (INIS)

    Cowan, S.P.; Sprecher, W.M.; Walton, R.D.

    1983-01-01

    The policy of the U.S. Department of Energy is to pursue an aggressive and credible waste management program that advocates final disposal of government generated (defense) high-level nuclear wastes in a manner consistent with environmental, health, and safety responsibilities and requirements. The Defense Waste Processing Facility (DWPF) is an essential component of the Department's program. It is the first project undertaken in the United States to immobilize government generated high-level nuclear wastes for geologic disposal. The DWPF will be built at the Department's Savannah River Plant near Aiken, South Carolina. When construction is complete in 1989, the DWPF will begin processing the high-level waste at the Savannah River Plant into a borosilicate glass form, a highly insoluble and non-dispersable product, in easily handled canisters. The immobilized waste will be stored on site followed by transportation to and disposal in a Federal repository. The focus of this paper is on the DWPF. The paper discusses issues which justify the project, summarizes its technical attributes, analyzes relevant environmental and insitutional factors, describes the management approach followed in transforming technical and other concepts into concrete and steel, and concludes with observations about the future role of the facility

  5. Radioactive Waste Management Research Program Plan for high-level waste: 1987

    International Nuclear Information System (INIS)

    1987-05-01

    This plan will identify and resolve technical and scientific issues involved in the NRC's licensing and regulation of disposal systems intended to isolate high level hazardous radioactive wastes (HLW) from the human environment. The plan describes the program goals, discusses the research approach to be used, lays out peer review procedures, discusses the history and development of the high level radioactive waste problem and the research effort to date and describes study objectives and research programs in the areas of materials and engineering, hydrology and geochemistry, and compliance assessment and modeling. The plan also details the cooperative interactions with international waste management research programs. Proposed Earth Science Seismotectonic Research Program plan for radioactive waste facilities is appended

  6. Development of activated carbon pore structure via physical and chemical activation of biomass fibre waste

    International Nuclear Information System (INIS)

    Williams, Paul T.; Reed, Anton R.

    2006-01-01

    Biomass waste in the form of biomass flax fibre, produced as a by-product of the textile industry was processed via both physical and chemical activation to produce activated carbons. The surface area of the physically activated carbons were up to 840 m 2 g -1 and the carbons were of mesoporous structure. Chemical activation using zinc chloride produced high surface area activated carbons up to 2400 m 2 g -1 and the pore size distribution was mainly microporous. However, the process conditions of temperature and zinc chloride concentration could be used to manipulate the surface area and porosity of the carbons to produce microporous, mesoporous and mixed microporous/mesoporous activated carbons. The physically activated carbons were found to be a mixture of Type I and Type IV carbons and the chemically activated carbons were found to be mainly Type I carbons. The development of surface morphology of physically and chemically activated carbons observed via scanning electron microscopy showed that physical activation produced activated carbons with a nodular and pitted surface morphology whereas activated carbons produced through chemical activation had a smooth surface morphology. Transmission electron microscopy analysis could identify mesopore structures in the physically activated carbon and microporous structures in the chemically activated carbons

  7. Final report, Task 2: alternative waste management options, Nuclear Fuel Services, Inc., high level waste

    International Nuclear Information System (INIS)

    1978-01-01

    Of the alternatives considered for disposal of the high-level waste in tanks 8D2 and 8D4, the following process is recommended: homogenization of the contents of tank 8D2, centrifugation of the sludge and supernate, mixing of the 8D4 acid waste with the centrifuged sludge, and converting the mixture to a borosilicate glass using the Hanford spray calciner/in-can melter

  8. Problems related to final disposal of high-level radioactive waste in Russia

    International Nuclear Information System (INIS)

    Velichkin, Vasily I.

    1999-01-01

    According to this presentation, the radioactivity of the total amount of radioactive waste accumulated in Russia to date is 1.5*10 9 Ci and of spent fuel 4.5*10 9 Ci. A table is given that shows the source, type, volume activity and storage type under the responsibility of the different departments and enterprises. 99.9% of the wastes are accumulated at the enterprises of Minatom of the Russian Federation. Some companies inject their liquid wastes from ionisation sources and intermediate liquid waste from the nuclear power industry into deep-seated reliably isolated aquifers. The Mayak plant has released liquid low-level and intermediate wastes into artificial reservoirs and Lake Karachay. Liquid high-level wastes are always stored in special tanks at interim storage facilities. A large number of nuclear submarines are laid up in North-Western Russia and East Russia, with spent fuel still in place as the interim storages in these regions are filled up and there are no conditioning plants. Underground disposal is considered the best way of isolating radioactive waste for as long as it is hazardous to the environment. Two new technologies are discussed. One involves including long-lived isotopes in high-stable mineral matrices, the other uses selective separation from the bulk of wastes. The matrices should be disposed of deep in the Earth's crust, at least 2-3 km down. Liquid waste of caesium-strontium fraction must be transformed into glass-like form and stored underground at a depth of a few hundred metres. Short-lived low level and intermediate level wastes should be conditioned and then deposited in subsurface ferroconcrete repositories constructed in clays. Finally, the presentation discusses the selection of sites and conditions for radioactive waste disposal. Two sites are discussed, the Mayak plant and a possible site at Mining Chemical Combine in Krasnoyarsk-26

  9. Underground excavation methods for a high-level waste repository

    International Nuclear Information System (INIS)

    Peshel, J.; Gupta, D.; Nataraja, M.

    1990-01-01

    This paper reports on rock excavation methods for a High-Level Waste repository that should be selected to limit the potential for creating preferential pathways for groundwater to travel to the waste packages or for radionuclides to migrate to the accessible environment. The use of water and other foreign substances should be controlled so that the repository performance is not compromised. The excavated openings should remain stable so that operations can be carried out safely and the retrievability option maintained. As per the current conceptual designs presented by the Department of Energy, the exploratory shaft facility becomes a part of the repository if the Yucca Mountain site is found suitable for repository development. Therefore, the methods of constructing the underground openings should be compatible with the performance requirements for the repository. Also, the degree of damage to the rock surrounding the openings and the extent of the damage zone should not preclude adequate site characterization. The ESf construction and operation should be compatible with the site data gathering activities, such as geological, thermomechanical, hydrological and geochemical testing

  10. The disposal of high-level radioactive waste. Vol. 1

    International Nuclear Information System (INIS)

    Parker, F.L.; Broshears, R.E.; Pasztor, J.

    1984-01-01

    The Beijer Institute received request from the Swedish Board for Spent Nuclear Fuel (Naemnden for Anvaent Kaernbraensle - NAK) to undertake an international review of the major programmes which were currently making arrangements for the future disposal of high-level radioactive wastes and spent nuclear fuel. The request was accepted, a detailed proposal was worked out and agreed to by NAK, for a critical technical review which concentrated on the following three main tasks: 1. a 'state-of-the-art' review of selected ongoing disposal programmes, both national and international; 2. an assessment of the scientific and technical controversies involved, and 3. recommendations for further research in this field. This review work was to be built on a survey of the available technical literature which was to serve as a basis for a series of detailed interviews, consultations and discussions with scientific and technical experts in Japan, Canada, USA, Belgium, Federal Republic of Germany, France, Switzerland and the United Kingdom. This first volume contains: disposal options; review of the state-of-the-art (international activities, national programs); analysis of waste disposal systems. (orig./HP)

  11. Electrochemical ion-exchange for medium active liquid waste treatment

    International Nuclear Information System (INIS)

    Bridger, N.J.; Turner, A.D.

    1987-01-01

    Electrochemical ion-exchange has already been demonstrated to be a robust, effective process for the treatment of active liquid wastes -with high decontamination and volume reduction factors, and only a low energy requirement. The primary aim of this new programme is to scale up this process - initially to 0.1m 3 /h, and ultimately to 1 3 m/h. A new 0.4m 2 electrode module has been designed and constructed, together with 3m 3 feed tanks for the first phase of this work. Further development work is also being carried out on alternative electrode designs and fabrication methods, as well as new exchange media (including inorganic absorbers and organic chelating resins) in order to optimize selectivity performance. (author)

  12. High-level radioactive-waste-disposal investigations in Texas

    International Nuclear Information System (INIS)

    Smith, R.D.

    1983-01-01

    The Texas Energy and Natural Resources Advisory Council (TENRAC) was designated in 1980 to coordinate the interaction between the State of Texas and the federal government relating to the high-level radioactive waste disposal issue. This report was prepared to summarize the many aspects of that issue with particular emphasis on the activities in Texas. The report is intended to provide a comprehensive introduction for individuals with little or no previous exposure to the issue and to provide a broader perspective for those individuals who have addressed specific aspects of the issue but have not had the opportunity to study it in a broader context. Following the introduction, contents of this report are as follows: (1) general status of major repository siting investigations in the US; (2) detailed review of Texas studies; (3) possible facilities to be sited in Texas; (4) current Texas policy; (5) federal regulations; and (6) federal legislation. 9 figures, 2 tables

  13. Recovery of nitric acid from simulated acidic high level radioactive waste using pore-filled anion exchange membranes

    International Nuclear Information System (INIS)

    Chavan, Vivek; Agarwal, Chhavi; Pandey, A.K.; Goswami, A.

    2014-01-01

    Acidic waste is generated at different stages of nuclear fuel cycle. The waste contains minor amounts of actinides ( 241 Am, Pu, Np) along with large number of long-lived radionuclides such as 137 Cs, 90 Sr, 106 Ru etc. Before disposal or storage, the overall activity of the waste needs to be reduced. Along with this, the high amount of acid present in the waste needs to be removed. In this study, DD has been used to recover nitric acid from acidic solutions with compositions similar to radioactive waste using pore-filled anion exchange membranes

  14. High-level radioactive waste incorporation into (special) cements

    International Nuclear Information System (INIS)

    Roy, D.M.; Gouda, G.R.

    1978-01-01

    A feasibility study has demonstrated that very strong, durable, relatively impermeable cylinders may be prepared by hot pressing combinations of cements with simulated radioactive waste solids. While the properties have not been studied exhaustively, the results suggest an optional method for immobilization and isolation of radioactive waste. Samples prepared with calcium aluminate cements appeared to have properties superior to those with Portland cements. Four simulated radioactive waste compositions having high rare-earth oxide contents, and some containing a large excess of NaNO 3 , were studied. Modest temperatures [423 to 673 K (150 to 400 0 C)] were used for hot pressing at pressures from 178 to 345 MPa. Dense strong very low porosity specimens resulted when mixtures containing from 10 to 50% waste were hot pressed, incorporating also a small percentage of water. In addition, high-strength cement cylinders were prepared with the waste solid (approximately 20 wt% waste) in a separate core and were very resistant to leaching by water near its boiling point. With this configuration, even the NaNO 3 -containing wastes were resistant to leaching by water

  15. Alternative processes for managing existing commercial high-level radioactive wastes

    International Nuclear Information System (INIS)

    1976-04-01

    A number of alternatives are discussed for managing high-level radioactive waste presently stored at the West Valley, New York, plant owned by Nuclear Fuel Services, Inc. These alternatives (liquid storage, conversion to cement, shale fracturing, shale cement, calcination, aqueous silicate, conversion to glass, and salt cake) are limited to concepts presently under active investigation by ERDA. Each waste management option is described and examined regarding the status of the technology; its applications to managing NFS waste; its advantages and disadvantages; the research and development needed to implement the option; safety considerations; and estimated costs and time to implement the process

  16. Status of technology for isolating high-level radioactive wastes in geologic repositories

    International Nuclear Information System (INIS)

    Klingsberg, C.; Duguid, J.

    1980-10-01

    This report attempts to summarize the status of scientific and technological knowledge relevant to long-term isolation of high-level and transuranic wastes in a mined geologic repository. It also identifies and evaluates needed information and identifies topics in which work is under way or needed to reduce uncertainties. The major findings and conclusions on the following topics are presented: importance of the systems approach; prospects for successful isolation of wastes; need for site-specific investigations; human activities in the future; importance of modelling; disposal of transuranic wastes; status of technology of isolation barriers, performance assessment, site selection and characterization, and potential host rocks

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

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Steven E.

    2013-11-11

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

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

    International Nuclear Information System (INIS)

    Kelly, Steven E.

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Manaktala, H.K.; Interrante, C.G.

    1990-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-12-01

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

  1. High temperature vitrification of surrogate Savannah River Site (SRS) mixed waste materials

    International Nuclear Information System (INIS)

    Applewhite-Ramsey, A.; Schumacher, R.F.; Spatz, T.L.; Newsom, R.A.; Circeo, L.J.; Danjaji, M.B.

    1995-01-01

    The Savannah River Technology Center (SRTC) has been funded through the DOE Office of Technology Development (DOE-OTD) to investigate high-temperature vitrification technologies for the treatment of diverse low-level and mixed wastes. High temperature vitrification is a likely candidate for processing heterogeneous solid wastes containing low levels of activity. Many SRS wastes fit into this category. Plasma torch technology is one high temperature vitrification method. A trial demonstration of plasma torch processing is being performed at the Georgia Institute of Technology on surrogate SRS wastes. This effort is in cooperation with the Engineering Research and Development Association of Georgia Universities (ERDA) program. The results of phase 1 of these plasma torch trials will be presented

  2. Modeling by GASP-IV simulation of high-level nuclear waste disposal

    International Nuclear Information System (INIS)

    Kurstedt, H.A. Jr.; DePorter, E.L.; Turek, J.L.; Funk, S.K.; Rasbach, C.E.

    1981-01-01

    High-level nuclear waste generated by defense-oriented and commercial nuclear energy activities are to be stored ultimately in underground repositories. Research continues on the waste-form and waste-form processing. DOE managers must coordinate the results of this research, the capacities and availability times of the permanent geologic storage repositories, and the capacities and availability times of interim storage facilities (pending availability of permanent repositories). Comprehensive and active DOE program-management information systems contain predicted generation of nuclear wastes from defense and commercial activities; milestones on research on waste-forms; and milestones on research and development, design, acquisition, and construction of facilities and repositories. A GASP IV simulation model is presented which interfaces all of these data. The model accepts alternate management decisions; relates all critical milestones, all research and development data, and the generation of waste nuclear materials; simulates the passage of time; then, predicts the impact of those alternate decisions on the availability of storage capacity for waste nuclear materials. 3 references, 3 figures

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

    International Nuclear Information System (INIS)

    Mahidhara, R.K.; Elleman, T.S.; Murty, K.L.

    1992-11-01

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

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

    International Nuclear Information System (INIS)

    Garcia, E.

    1996-01-01

    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. Development and evaluation of candidate high-level waste forms

    International Nuclear Information System (INIS)

    Bernadzikowski, T.A.

    1981-01-01

    Some seventeen candidate waste forms have been investigated under US Department of Energy programs as potential media for the immobilization and geologic disposal of the high-level radioactive wastes (HLW) resulting from chemical processing of nuclear reactor fuels and targets. Two of these HLW forms were selected at the end of fiscal year (FY) 1981 for intensive development if FY 1982 to 1983. Borosilicate glass was continued as the reference form. A crystalline ceramic waste form, SYNROC, was selected for further product formulation and process development as the alternative to borosilicate glass. This paper describes the bases on which this decision was made

  6. High temperature slagging incinerator for alpha contaminated wastes

    International Nuclear Information System (INIS)

    Van de Voorde, N.

    1985-01-01

    This report describes the experiences collected by the treatment of plutonium-contaminated wastes, in the High Temperature Slagging Incinerator at the C.E.N./S.C.K. at Mol, with the support of the Commission of the European Communities. The major objective of the exercise is to demonstrate the operability of this facility for the treatment of mixed transuranic (TRU) and beta-gamma solid waste material. The process will substantially reduce the TRU waste volume by burning the combustibles and converting the non-combustibles into a chemically inert and physically stable basalt-like slag product, suitable for safe transport and final disposal. (Auth.)

  7. Spanish high level radioactive waste management system issues

    International Nuclear Information System (INIS)

    Espejo, J.M.; Beceiro, A.R.

    1992-01-01

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

  8. Spanish high level radioactive waste management system issues

    International Nuclear Information System (INIS)

    Ulibarri, A.; Veganzones, A.

    1993-01-01

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

  9. High-level nuclear waste disposal: Ethical considerations

    International Nuclear Information System (INIS)

    Maxey, M.N.

    1985-01-01

    Popular skepticism about, and moral objections to, recent legislation providing for the management and permanent disposal of high-level radioactive wastes have derived their credibility from two major sources: government procrastination in enacting waste disposal program, reinforcing public perceptions of their unprecedented danger and the inflated rhetoric and pretensions to professional omnicompetence of influential scientists with nuclear expertise. Ethical considerations not only can but must provide a mediating framework for the resolution of such a polarized political controversy. Implicit in moral objections to proposals for permanent nuclear waste disposal are concerns about three ethical principles: fairness to individuals, equitable protection among diverse social groups, and informed consent through due process and participation

  10. The precautionary principle and high-level nuclear waste policy

    International Nuclear Information System (INIS)

    Frishman, S.

    1999-01-01

    The 'Precautionary Principle' has grown from the broadening observation that there is compelling evidence that damage to humans and the world-wide environment is of such a magnitude and seriousness that new principles for conducting human activities are necessary. One of the various statements of the Precautionary Principle is: when an activity raises threats of harm to human health or the environment, precautionary measures should be taken even if some cause and effect relationships are not fully established scientifically. The use of a precautionary principle was a significant recommendation emerging from the 1992 United Nations Conference on Environment and Development, held in Rio de Janeiro, Brazil, and it is gaining acceptance in discussions ranging from global warming to activities that affect the marine environment, and far beyond. In the US high-level nuclear waste policy, there is a growing trend on the part of geologic repository proponents and regulators to shift the required safety evaluation from a deterministic analysis of natural and engineered barriers and their interactions to risk assessments and total system waste containment and isolation performance assessment. This is largely a result of the realisation that scientific 'proof' of safety cannot be demonstrated to the level repository proponents have led the American public to expect. Therefore, they are now developing other methods in an attempt to effectively lower the repository safety expectations of the public. Implicit in this shift in demonstration of 'proof' is that levels of uncertainty far larger than those generally taken as scientifically acceptable must be accepted in repository safety, simply because greater certainty is either too costly, in time and money, or impossible to achieve at the potential Yucca Mountain repository site. In the context of the Precautionary Principle, the repository proponent must bear the burden of providing 'Acceptable' proof, established by an open

  11. Effect of ultrasonic specific energy on waste activated sludge ...

    African Journals Online (AJOL)

    The effect of ultrasonic specific energy on waste activated sludge (WAS) solubilization and enzyme activity was investigated in this study. Experimental results showed that the increase of ultrasonic specific energy in the range of 0 - 90000 kJ/kg dried sludge (DS) benefited WAS particle size reduction and the solubilization ...

  12. Evaporation Of Hanford Waste Treatment Plant Direct Feed Low Activity Waste Effluent Management Facility Core Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Mcclane, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-01

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate, LMOGC) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation, and recycled to the LAW vitrification facility. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream is to evaporate it in a new evaporator, in the Effluent Management Facility (EMF), and then return it to the LAW melter. It is important to understand the composition of the effluents from the melter and new evaporator, so that the disposition of these streams can be accurately planned and accommodated. Furthermore, alternate disposition of the LMOGC stream would eliminate recycling of problematic components, and would reduce the need for closely integrated operation of the LAW melter and the Pretreatment Facilities. Long-term implementation of this option after WTP start-up would decrease the LAW vitrification mission duration and quantity of glass waste, amongst the other operational complexities such a recycle stream presents. In order to accurately plan for the disposition path, it is key to experimentally determine the fate of contaminants. To do this, testing is needed to accurately account for the buffering chemistry of the components, determine the achievable evaporation end point, identify insoluble solids that form, and determine the distribution of key regulatory-impacting constituents. The LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures, have limited solubility in the glass waste form, and represent a materials corrosion concern, such as halides and sulfate. Because this stream will recycle within WTP, these components will accumulate in the Melter Condensate

  13. Active Waste Materials Corrosion and Decontamination Tests

    International Nuclear Information System (INIS)

    Danielson, M.J.; Elmore, M.R.; Pitman, S.G.

    2000-01-01

    Stainless steel alloys, 304L and 316L, were corrosion tested in representative radioactive samples of three actual Hanford tank waste solutions (Tanks AW-101, C-104, AN-107). Both the 304L and 316L exhibited good corrosion performance when immersed in boiling waste solutions. The maximum general corrosion rate was 0.015 mm/y (0.60 mils per year). Generally, the 304L had a slightly higher rate than the 316L. No localized attack was observed after 122 days of testing in the liquid phase, liquid/vapor phase, or vapor phase. Radioactive plate-out decontamination tests indicated that a 24-hour exposure to 1 und M HNO 3 could remove about 99% of the radioactive components in the metal film when exposed to the C-104 and AN-107 solutions. The decontamination results are less certain for the AW-101 solution, since the initial contamination readings exceeded the capacity of the meter used for this test

  14. In-situ nitrite analysis in high level waste tanks

    International Nuclear Information System (INIS)

    O'Rourke, P.E.; Prather, W.S.; Livingston, R.R.

    1992-01-01

    The Savannah River Site produces special nuclear materials used in the defense of the United States. Most of the processes at SRS are primarily chemical separations and purifications. In-situ chemical analyses help improve the safety, efficiency and quality of these operations. One area where in situ fiberoptic spectroscopy can have a great impact is the management of high level radioactive waste. High level radioactive waste at SRS is stored in more than 50 large waste tanks. The waste exists as a slurry of nitrate salts and metal hydroxides at pH's higher than 10. Sodium Nitrite is added to the tanks as a corrosion inhibitor. In-situ fiberoptic probes are being developed to measure the nitrate, nitrite and hydroxide concentrations in both liquid and solid fractions. Nitrite levels can be measured between 0.01M and 1M in a 1mm pathlength optical cell

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

    Energy Technology Data Exchange (ETDEWEB)

    WILLIS, W.L.

    2000-06-15

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

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

    International Nuclear Information System (INIS)

    WILLIS, W.L.

    2000-01-01

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

  17. Interaction of cementitious materials with high-level waste

    International Nuclear Information System (INIS)

    Lemmens, Karel; Cachoir, Christelle; Ferrand, Karine; Mennecart, Thierry; Gielen, Ben; Vercauter, Regina

    2012-01-01

    vitrified HLW, the high pH is expected to cause a faster dissolution of the glass matrix, driven by secondary phase formation. Low long-term dissolution rates, such as expected in neutral pH conditions, will probably be reached only after the formation of a sufficiently thick alteration layer separating the pristine glass from the bulk of the concrete, thus causing a pH gradient. So far, this process cannot yet be quantified well enough. For spent fuel, the preliminary results are not entirely consistent. The tests with spent fuel suggest that the high pH is not detrimental for the fuel stability. The tests with (doped) UO 2 suggest, however, that it may take more time to reach the threshold fuel activity where radiolytic oxidation becomes unimportant. Moreover, there are indications that the non-oxidative fuel dissolution could be accelerated at high pH by colloid formation in certain conditions. The factors affecting this colloid formation are not understood so far. Therefore it is not yet possible to assess whether this process will be relevant under in situ conditions. The uncertainties about the present dissolution rates are still large due to the small number of data. A continuation of the programme is necessary to improve the rate estimations and to better understand the underlying mechanisms. For vitrified waste, the focus will be on the evolution at the glass/concrete interface (secondary phase formation, pH evolution...). For spent fuel, it is important to understand better the nature of the dissolution process (oxidative or non-oxidative) as a function of the fuel activity, the role of colloids, and the differences between the behaviour of alpha-doped UO 2 and real spent fuel. Tests with spent fuel pellets and hydrogen gas will be started at INE to see if hydrogen reinforces the reducing capacity enough to suppress the oxidative fuel dissolution when the alpha activity of the fuel is still high. Another important remaining uncertainty is the potential effect of

  18. Managing the nation's commercial high-level radioactive waste

    International Nuclear Information System (INIS)

    Cotton, T.

    1985-01-01

    With the passage of the Nuclear Waste Policy Act of 1982 (NWPA), Congress for the first time established in law a comprehensive Federal policy for commercial high-level radioactive waste management, including interim storage and permanent disposal. NWPA provides sufficient authority for developing and operating a high-level radioactive waste management system based on disposal in mined geologic repositories. Authorization for other types of waste facilities will not be required unless major problems with geologic disposal are discovered, and studies to date have identified no insurmountable technical obstacles to developing geologic repositories. The NWPA requires the Department of Energy (DOE) to submit to Congress three key documents: (1) a Mission Plan, containing both a waste management plan with a schedule for transferring waste to Federal facilities and an implementation program for choosing sites and developing technologies to carry out that plan; (2) a monitored retrievable storage (MRS) proposal, to include a site-specific design for a long-term federal storage facility, an evaluation of whether such an MRS facility is needed and feasible, and an analysis of how an MRS facility would be integrated with the repository program if authorized by Congress; and (3) a study of alternative institutional mechanisms for financing and managing the radioactive waste system, including the option of establishing an independent waste management organization outside of DOE. The Mission Plan and the report on alternative institutional mechanisms were submitted to the 99th US Congress in 1985. The MRS proposal is to be submitted in early 1986. Each of these documents is discussed following an overview of the Nuclear Waste Policy Act of 1982

  19. Radiation effects on medium active waste forms. Annual report - 1988

    International Nuclear Information System (INIS)

    Johnson, D.; Wilding, C.; Lyon, C.

    1989-01-01

    Work has continued on measurements of dimensional changes, strength, and gas evolution on samples of several simulated waste forms under accelerated γ and α irradiation conditions. Samples of RMA5 (mixed ion exchangers in modified vinylester polymer) and RMA10 (incinerated PCM materials in cement) maintain their integrity during irradiation but samples of RMA3 (organic ion exchangers in cement) and RMA11.1 (mixed PCM materials in cement) swell and eventually disintegrate under some γ irradiation conditions. Disintegration of RMA3 samples occurred when samples were γ irradiated whilst immersed in water. Samples of RMA11.1 which cannot rapidly dry out swell, sometimes substantially, during γ irradiation. The principal gases of interest in gas evolution experiments are hydrogen and oxygen. Hydrogen is evolved under all circumstances but oxygen evolution does not always occur. Samples of RMA10 evolve oxygen when α irradiated in an inert atmosphere but oxygen concentration initially falls during α irradiation in air atmosphere. Samples of RMA11.1 absorb oxygen from an air atmosphere during both α and γ irradiation. A comparison has been carried out of the effects of γ and α irradiation on identical cement grouts using BFS/OPC mixes produced under high shear mixing conditions. In contrast to earlier results on such systems, no γ irradiated samples showed physical deterioration after irradiation to 9 MGy but the a irradiated samples all showed surface cracks after about 1 MGy. The gas evolution measurements showed that during α irradiation oxygen evolution commenced after a dose of ∼ 1 MGy whereas oxygen was completely removed from the atmosphere γ irradiation. Hydrogen was evolved under all conditions and the rate of production was found to be dependent upon the dose rate. More hydrogen was evolved during α irradiations than during γ irradiation. A technique for the measurement of hydrogen permeability through cement systems has been further developed

  20. Activation analyses updating the ITER radioactive waste assessment

    International Nuclear Information System (INIS)

    Pampin, R.; Zheng, S.; Lilley, S.; Na, B.C.; Loughlin, M.J.; Taylor, N.P.

    2012-01-01

    Highlights: ► Comprehensive updated of ITER radwaste assessment. ► Latest coupled neutronics and activation methods. ► Type A waste at shutdown decays to TFA within 100 years. ► Most type B waste at shutdown is still type B after 100 years. - Abstract: A study is reported which computes the radiation transport and activation response throughout the ITER machine and updates the ITER radioactive waste assessment using modern 3D models and up-to-date methods. The latest information on component design, maintenance, replacement schedules and materials is adopted. The radwaste classification is revised for all the major components of ITER, as well as several representative port plugs. Results include categorisation snapshots at different decay times, time histories of radiological quantities throughout the machine, and guidelines on interim decay times for components. All plasma-facing materials except tungsten are found to classify as type B due to the transmutation of their main constituents. Major contributors to the IRAS index of all materials are reported. Elemental concentration limits for type A classification of first wall and divertor materials are obtained; for the steels, only a reduction in service lifetime can reduce the waste class. Comparison of total waste amounts with earlier assessments is limited by the fact that analyses of some components are still preliminary; the trend, however, indicates a potential reduction in the total amount of waste if component segregation is demonstrated.

  1. TWRS retrieval and storage mission. Immobilized low-activity waste disposal plan

    International Nuclear Information System (INIS)

    Shade, J.W.

    1998-01-01

    The TWRS mission is to store, treat, and immobilize highly radioactive Hanford waste (current and future tank waste and the encapsulated cesium and strontium) in a safe, environmentally sound, and cost-effective manner (TWRS JMN Justification for mission need). The mission includes retrieval, pretreatment, immobilization, interim storage and disposal, and tank closure. As part of this mission, DOE has established the TWRS Office to manage all Hanford Site tank waste activities. The TWRS program has identified the need to store, treat, immobilize, and dispose of the highly radioactive Hanford Site tank waste and encapsulated cesium and strontium materials in an environmentally sound, safe, and cost-effective manner. To support environmental remediation and restoration at the Hanford Site a two-phase approach to using private contractors to treat and immobilize the low-activity and high-level waste currently stored in underground tanks is planned. The request for proposals (RFP) for the first phase of waste treatment and immobilization was issued in February 1996 (Wagoner 1996) and initial contracts for two private contractor teams led by British Nuclear Fuels Ltd. and Lockheed-Martin Advanced Environmental Services were signed in September 1996. Phase 1 is a proof-of-concept and commercial demonstration effort to demonstrate the technical and business feasibility of using private facilities to treat Hanford Site waste, maintain radiological, nuclear, process, and occupational safety; and maintain environmental protection and compliance while reducing lifecycle costs and waste treatment times. Phase 1 production of ILAW is planned to begin in June 2002 and could treat up to about 13 percent of the waste. Phase 1 production is expected to be completed in 2007 for minimum order quantities or 2011 for maximum order quantities. Phase 2 is a full-scale production effort that will begin after Phase 1 and treat and immobilize most of the waste. Phase 2 production is

  2. 1993 annual report of hazardous waste activities for the Oak Ridge K-25 site

    Energy Technology Data Exchange (ETDEWEB)

    1994-02-01

    This report is a detailed listing of all of the Hazardous Waste activities occurring at Martin Marietta`s K-25 site. Contained herein are hazardous waste notification forms, waste stream reports, generator fee forms and various TSDR reports.

  3. 1993 annual report of hazardous waste activities for the Oak Ridge K-25 site

    International Nuclear Information System (INIS)

    1994-02-01

    This report is a detailed listing of all of the Hazardous Waste activities occurring at Martin Marietta's K-25 site. Contained herein are hazardous waste notification forms, waste stream reports, generator fee forms and various TSDR reports

  4. Production of highly efficient activated carbons from industrial wastes for the removal of pharmaceuticals from water-A full factorial design.

    Science.gov (United States)

    Jaria, Guilaine; Silva, Carla Patrícia; Oliveira, João A B P; Santos, Sérgio M; Gil, María Victoria; Otero, Marta; Calisto, Vânia; Esteves, Valdemar I

    2018-02-26

    The wide occurrence of pharmaceuticals in aquatic environments urges the development of cost-effective solutions for their removal from water. In a circular economy context, primary paper mill sludge (PS) was used to produce activated carbon (AC) aiming the adsorptive removal of these contaminants. The use of low-cost precursors for the preparation of ACs capable of competing with commercial ACs continues to be a challenge. A full factorial design of four factors (pyrolysis temperature, residence time, precursor/activating agent ratio, and type of activating agent) at two levels was applied to the production of AC using PS as precursor. The responses analysed were the yield of production, percentage of adsorption for three pharmaceuticals (sulfamethoxazole, carbamazepine, and paroxetine), specific surface area (S BET ), and total organic carbon (TOC). Statistical analysis was performed to evaluate influencing factors in the responses and to determine the most favourable production conditions. Four ACs presented very good responses, namely on the adsorption of the pharmaceuticals under study (average adsorption percentage around 78%, which is above that of commercial AC), and S BET between 1389 and 1627 m 2  g -1 . A desirability analysis pointed out 800 °C for 60 min and a precursor/KOH ratio of 1:1 (w/w) as the optimal production conditions. Copyright © 2018 Elsevier B.V. All rights reserved.

  5. High level waste vitrification at the SRP [Savannah River Plant] (DWPF [Defense Waste Processing Facility] summary)

    International Nuclear Information System (INIS)

    Weisman, A.F.; Knight, J.R.; McIntosh, D.L.; Papouchado, L.M.

    1988-01-01

    The Savannah River Plant has been operating a nuclear fuel cycle since the early 1950's. Fuel and target elements are fabricated and irradiated to produce nuclear materials. After removal from the reactors, the fuel elements are processed to extract the products, and waste is stored. During the thirty years of operation including evaporation, about 30 million gallons of high level radioactive waste has accumulated. The Defense Waste Processing Facility (DWPF) under construction at Savannah River will process this waste into a borosilicate glass for long-term geologic disposal. The construction of the DWPF is about 70% complete; this paper will describe the status of the project, including design demonstrations, with an emphasis on the melter system. 9 figs

  6. The management of high level waste and its environmental impact

    International Nuclear Information System (INIS)

    Saunders, P.A.H.

    1981-01-01

    This paper summarises the techniques that are used for the management of the radioactive wastes that result from the nuclear generation of electricity and that cannot be released directly into the environment. The quantities and characteristics of the wastes are outlined and a description is given of current and probable future stores and their environmental impact. The research and development programme that is being undertaken to establish the safety and environmental impact of an ultimate repository is discussed and a comparison is made between the activities and toxic potentials of the wastes and those of naturally occurring materials, fossil fuels and fertilisers. It is concluded that the wastes can be managed without undue risk to man or to the environment. (author)

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

    International Nuclear Information System (INIS)

    Decker, L.A.

    1998-04-01

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

  8. 226Ra adsorption on active coals from waste waters

    International Nuclear Information System (INIS)

    Panturu, E.; Georgescu, D.P.; Serban, N.; Filip, D.; Radulescu, R.

    2000-01-01

    During the mining and extraction of uranium, the principle means of protection measurement is to prevent uranium and its products diffusing into the environment. The main carriers of radioactive elements in the environment are air and water. Therefore, reduction of the pollution at a uranium mine can be achieved by the treatment of waste waters contaminated with 226 Ra Radium contaminated waste waters represent a major biological risk. This paper presents the results of the study of the sorption of 226 Ra on active coal mechanisme and the influence of the physical and chemical characteristics of fluid. The 226 Ra removal from the residue pond water at the uranium ore processing plant was studied using eight types of indigenous active coals. The experimental results for each type of active coal and their effect on removal of 226 Ra from waste waters are presented in this paper. (author)

  9. Status report on the Nuclear Regulatory Commission regulations for land disposal of low-level radioactive wastes and geologic repository disposal of high-level wastes

    International Nuclear Information System (INIS)

    Browning, R.E.; Bell, M.J.; Dragonette, K.S.; Johnson, T.C.; Roles, G.W.; Lohaus, P.H.; Regnier, E.P.

    1984-01-01

    On 27 December 1982, the United States Nuclear Regulatory Commission (NRC) amended its regulations to provide specific requirements for licensing the land disposal of low-level radioactive wastes. The regulations establish performance objectives for land disposal of waste; technical requirements for the siting, design, operations, and closure activities for a near-surface disposal facility; technical requirements concerning waste form and classification that waste generators must meet for the land disposal of waste; institutional requirements; financial assurance requirements; and administrative and procedural requirements for licensing a disposal facility. Waste generators must comply with the waste form and classification provisions of the new rule, on 27 December 1983, one year later. During this implementation period, licensees must develop programmes to ensure compliance with the new waste form and classification provisions. The NRC is also promulgating regulations specifying the technical criteria for disposal of high-level radioactive wastes in geological repositories. The proposed rule was published for public comment in July 1981. Public comments have been received and considered by the Commission staff. The Commission will soon approve and publish a revised final rule. While the final rule being considered by the Commission is fundamentally the same as the proposed rule, provisions have been added to permit flexibility in the application of numerical criteria, some detailed design requirements have been deleted, and other changes have been made in response to comments. The rule is consistent with the recently enacted Nuclear Waste Policy Act of 1982. (author)

  10. Solidification of Savannah River Plant high level waste

    International Nuclear Information System (INIS)

    Maher, R.; Shafranek, L.F.; Kelley, J.A.; Zeyfang, R.W.

    1981-11-01

    Authorization for construction of the Defense Waste Processing Facility (DWPF) is expected in FY 83. The optimum time for stage 2 authorization is about three years later. Detailed design and construction will require approximately five years for stage 1, with stage 2 construction completed about two to three years later. Production of canisters of waste glass would begin in 1988, and the existing backlog of high level waste sludge stored at SRP would be worked off by about the year 2000. Stage 2 operation could begin in 1990. The technology and engineering are ready for construction and eventual operation of the DWPF for immobilizing high level radioactive waste at Savannah River Plant (SRP). Proceeding with this project will provide the public, and the leadership of this country, with a crucial demonstration that a major quantity of existing high level nuclear wastes can be safely and permanently immobilized. Early demonstration will both expedite and facilitate rational decision making on this aspect of the nuclear program. Delay in providing these facilities will result in significant DOE expenditures at SRP for new tanks just for continued temporary storage of wastes, and would probably result in dissipation of the intellectual and planning momentum that has built up in developing the project

  11. Application of SYNROC to high-level defense wastes

    International Nuclear Information System (INIS)

    Tewhey, J.D.; Hoenig, C.L.; Newkirk, H.W.; Rozsa, R.B.; Coles, D.G.; Ryerson, F.J.

    1981-01-01

    The SYNROC method for immobilization of high-level nuclear reactor wastes is currently being applied to US defense wastes in tank storage at Savannah River, South Carolina. The minerals zirconolite, perovskite, and hollandite are used in SYNROC D formulations to immobilize fission products and actinides that comprise up to 10% of defense waste sludges and coexisting solutions. Additional phase in SYNROC D are nepheline, the host phase for sodium; and spinel, the host for excess aluminum and iron. Up to 70 wt % of calcined sludge can be incorporated with 30 wt % of SYNROC additives to produce a waste form consisting of 10% nepheline, 30% spinel, and approximately 20% each of the radioactive waste-bearing phases. Urea coprecipitation and spray drying/calcining methods have been used in the laboratory to produce homogeneous, reactive ceramic powders. Hot pressing and sintering at temperatures from 1000 to 1100 0 C result in waste form products with greater than 97% of theoretical density. Hot isostatic pressing has recently been implemented as a processing alternative. Characterization of waste-form mineralogy has been done by means of XRD, SEM, and electron microprobe. Leaching of SYNROC D samples is currently being carried out. Assessment of radiation damage effects and physical properties of SYNROC D will commence in FY81

  12. Potential host media for a high-level waste repository

    Energy Technology Data Exchange (ETDEWEB)

    Hustrulid, W

    1982-01-01

    Earlier studies of burial of radioactive wastes in geologic repositories had concentrated on salt formations for well-publicized reasons. However, under the Carter administration, significant changes were made in the US nuclear waste management program. Changes which were made were: (1) expansion of the number of rock types under consideration; (2) adoption of the multiple-barrier approach to waste containment; (3) additional requirements for waste retrieval; and (4) new criteria proposed by the Nuclear Regulatory Commission for the isolation of high-level waste in geologic repositories. Results of the studies of different types of rocks as repository sites are summarized herein. It is concluded that each generic rock type has certain advantages and disadvantages when considered from various aspects of the waste disposal problem and that characteristics of rocks are so varied that a most favorable or least favorable rock type cannot be easily identified. This lack of definitive characteristics of rocks makes site selection and good engineering barriers very important for containment of the wastes. (BLM)

  13. Generation and management of solid waste resulting from tourist activities of the Porto de Galinhas - P

    Directory of Open Access Journals (Sweden)

    Jaqueline Guimarães Santos

    2015-04-01

    Full Text Available The significant solid waste generation, coupled with the lack of proper management of the waste generated, has been one of the issues of concern and conducting research on the part of scholars in the field. Tourism as an activity that positively impacts and negativity a given location, has emerged as an activity that can generate a lot of waste, especially in periods of high season, considering the increase of people moving to the tourist destinations. Accordingly, this study aims to analyze the generation and management of solid waste resulting from tourism in Porto de Galinhas, PE. We performed an exploratory, descriptive, qualitative study, conducted in the form of a case study in Porto de Galinhas, PE. The data collection was done interviews together social actors, as well as non-participant observation during data collection. The results showed that tourism activities in Porto de Galinhas result in a high amount of solid waste, and these are directed to inappropriate places. Although fate presents a combination of recyclable materials, RECYCLE, reuses this not a significant amount, given the proportion of waste generated.

  14. Pilot plant experience on high-level waste solidification and design of the engineering prototype VERA

    Energy Technology Data Exchange (ETDEWEB)

    Guber, W; Diefenbacher, W; Hild, W; Krause, H; Schneider, E; Schubert, G

    1972-11-01

    In the present paper the solidification process for highly active waste solutions as developed in the Karlsruhe Nuclear Research Center is presented. Its principal steps are: denitration, calcination in a spray calciner operated with superheated steam, melting of the calcine with appropriate additives to borosilicate glass in an induction-heated melting furnace. The operational experiences gained so far in the inactive 1:1 pilot plant are reported. Furthermore, a description is given of the projected multi-purpose experimental facility VERA 2 which is provided for processing the highly active waste solutions from the first German reprocessing plant WAK.

  15. Analysis of alternatives for immobilized low activity waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Burbank, D.A.

    1997-10-28

    This report presents a study of alternative disposal system architectures and implementation strategies to provide onsite near-surface disposal capacity to receive the immobilized low-activity waste produced by the private vendors. The analysis shows that a flexible unit strategy that provides a suite of design solutions tailored to the characteristics of the immobilized low-activity waste will provide a disposal system that best meets the program goals of reducing the environmental, health, and safety impacts; meeting the schedule milestones; and minimizing the life-cycle cost of the program.

  16. Analysis of alternatives for immobilized low-activity waste disposal

    International Nuclear Information System (INIS)

    Burbank, D.A.

    1997-01-01

    This report presents a study of alternative disposal system architectures and implementation strategies to provide onsite near-surface disposal capacity to receive the immobilized low-activity waste produced by the private vendors. The analysis shows that a flexible unit strategy that provides a suite of design solutions tailored to the characteristics of the immobilized low-activity waste will provide a disposal system that best meets the program goals of reducing the environmental, health, and safety impacts; meeting the schedule milestones; and minimizing the life-cycle cost of the program

  17. Predisposal management of high level radioactive waste. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    Radioactive waste is generated in the generation of electricity in nuclear power plants and in the use of radioactive material in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized. The principles and requirements that govern the safety of the management of radioactive waste are presented in 'The Principles of Radioactive Waste Management', 'Legal and Governmental Infrastructure for Nuclear, Radiation, Radioactive Waste and Transport Safety' and 'Predisposal Management of Radioactive Waste, Including Decommissioning'. The objective of this Safety Guide is to provide regulatory bodies and the operators that generate and manage radioactive waste with recommendations on how to meet the principles and requirements established in Refs for the predisposal management of HLW. This Safety Guide applies to the predisposal management of HLW. For liquid HLW arising from the reprocessing of spent fuel the recommendations of this Safety Guide apply from when liquid waste from the first extraction process is collected for storage and subsequent processing. Recommendations and guidance on the storage of spent fuel, whether or not declared as waste, subsequent to its removal from the storage facility of a reactor are provided in Refs. For spent fuel declared as waste this Safety Guide applies to all activities subsequent to its removal from the storage facility of a reactor and prior to its disposal. Requirements pertaining to the transport of spent fuel, whether or not declared as waste, and of all forms of HLW are established. This Safety Guide provides recommendations on the safety aspects of managing HLW, including the planning, design, construction, commissioning, operation and decommissioning of equipment or facilities for the predisposal management of HLW. It addresses the following elements: (a) The characterization and processing (i.e. pretreatment

  18. Characterization and application of municipal solid waste incineration (MSWI) bottom ash and waste granite powder in alkali activated slag

    NARCIS (Netherlands)

    Gao, X.; Yuan, B.; Yu, Q. L.; Brouwers, H. J.H.

    2017-01-01

    In this paper, the feasibility of using two solid wastes in alkali activated slag composites as construction and building materials is evaluated. One waste is the municipal solid waste incineration (MSWI) bottom ash, and the other one is fine granite powder from aggregate manufacturing. These two

  19. New insights into co-digestion of activated sludge and food waste: Biogas versus biofertilizer.

    Science.gov (United States)

    Ma, Yingqun; Yin, Yao; Liu, Yu

    2017-10-01

    This study explored two holistic approaches for co-digestion of activated sludge and food waste. In Approach 1, mixed activated sludge and food waste were first hydrolyzed with fungal mash, and produced hydrolysate without separation was directly subject to anaerobic digestion. In Approach 2, solid generated after hydrolysis of food waste by fungal mash was directly converted to biofertilizer, while separated liquid with high soluble COD concentration was further co-digested with activated sludge for biomethane production. Although the potential energy produced from Approach 1 was about 1.8-time higher than that from Approach 2, the total economic revenue generated from Approach 2 was about 1.9-fold of that from Approach 1 due to high market value of biofertilizer. It is expected that this study may lead to a paradigm shift in biosolid management towards environmental and economic sustainability. Copyright © 2017 Elsevier Ltd. All rights reserved.

  20. Conceptual process for conversion of high level waste to glass

    International Nuclear Information System (INIS)

    1975-01-01

    During a ten-year period highly radioactive wastes amounting to 22 million gallons of salt cake and 5 million gallons of wet sludge are to be converted to 1.2 million gallons of glass and 24 million gallons of decontaminated salt cake and placed in the new storage facilities which will provide high assurance of containment with minimal reliance on maintenance and surveillance. The glass will contain nearly all of the radioactivity in a form that is highly resistant to leaching and dispersion. The salt cake will contain a small amount of residual radioactivity. The process is shown in Figure 1 and the facilities may be arranged in seven modules to accomplish seven tasks, (1) remove wastes from tanks, (2) separate sludge and salt, (3) decontaminate salt, (4) solidify and package sludge and 137 Cs, (5) solidify and package decontaminated salt, (6) store high level waste, and (7) store decontaminated salt cake

  1. Development of high-performance supercapacitor electrode derived from sugar industry spent wash waste.

    Science.gov (United States)

    Mahto, Ashesh; Gupta, Rajeev; Ghara, Krishna Kanta; Srivastava, Divesh N; Maiti, Pratyush; D, Kalpana; Rivera, Paul-Zavala; Meena, R; Nataraj, S K

    2017-10-15

    This study aims at developing supercapacitor materials from sugar and distillery industry wastes, thereby mediating waste disposal problem through reuse. In a two-step process, biomethanated spent wash (BMSW) was acid treated to produce solid waste sludge and waste water with significantly reduced total organic carbon (TOC) and biological oxygen demand (BOD) content. Further, waste sludge was directly calcined in presence of activating agent ZnCl 2 in inert atmosphere resulting in high surface area (730-900m 2 g -1 ) carbon of unique hexagonal morphology. Present technique resulted in achieving two-faceted target of liquid-solid waste remediation and production of high-performance carbon material. The resulted high surface area carbon was tested in both three and two electrode systems. Electrochemical tests viz. cyclic voltammetry, galvanostatic charge-discharge and impedance measurement were carried out in aqueous KOH electrolyte yielding specific capacitance as high as 120Fg -1 , whereas all solid supercapacitor devised using PVA/H 3 PO 4 polyelectrolyte showed stable capacitance of 105Fg -1 at 0.2Ag -1 . The presence of transition metal particles and hetero-atoms on carbon surface were confirmed by XPS, EDX and TEM analysis which enhanced the conductivity and imparted pseudocapacitance to some extent into the working electrode. The present study successfully demonstrated production of high-performance electrode material from dirtiest wastewater making process green, sustainable and economically viable. Copyright © 2017. Published by Elsevier B.V.

  2. Safe immobilization of high-level nuclear reactor wastes

    International Nuclear Information System (INIS)

    Ringwood, A.; Kesson, S.; Ware, N.; Hibberson, W.; Major, A.

    1979-01-01

    The advantages and disadvantages of methods of immobilizing high-level radioactive wastes are discussed. Problems include the devitrification of glasses and the occurrence of radiation damage. An alternative method of radioctive waste immobilization is described in which the waste is incorporated in the constituent minerals of a synthetic rock, Synroc. Synroc is immune from devitrification and is composed of phases which possess crystal structures identical to those of minerals which are known to have retained radioactive elements in geological environments at elevated pressures and tempertures for long periods. The composition and mineralogy of Synroc is given and the process of immobilizing wastes in Synroc is described. Accelerated leaching tests at elevated pressures and temperatures are also described

  3. High-level waste description, inventory and hazard

    International Nuclear Information System (INIS)

    Crandall, J.; Hennelly, E.J.; McElroy, J.L.

    1983-01-01

    High-level nuclear waste (HLW), including its origin, is described and the current differences in definitions discussed. Quantities of defense and commercial radioactive HLW, both volume and curie content, are given. Current waste handling, which is interimin nature, is described for the several sites. The HLW hazard is defined by the times during which various radionuclides are the dominant contributors. The hazard is also compared to that of the ore. Using ICRP-2, which is the legal reference in the US, the hazard of the waste reduces to a level equal to the ore in about 300 years. The disposal plans are summarized and it is shown that regulatory requirements will probably govern disposal operations in such a conservative manner that the risk (product of hazard times probability of release) may well be lower than for any other wastes in existence or perhaps lower than those for any other human endeavor

  4. Glass formulation for phase 1 high-level waste vitrification

    International Nuclear Information System (INIS)

    Vienna, J.D.; Hrma, P.R.

    1996-04-01

    The purpose of this study is to provide potential glass formulations for prospective Phase 1 High-Level Waste (HLW) vitrification at Hanford. The results reported here will be used to aid in developing a Phase 1 HLW vitrification request for proposal (RFP) and facilitate the evaluation of ensuing proposals. The following factors were considered in the glass formulation effort: impact on total glass volume of requiring the vendor to process each of the tank compositions independently versus as a blend; effects of imposing typical values of B 2 O 3 content and waste loading in HLW borosilicate glasses as restrictions on the vendors (according to WAPS 1995, the typical values are 5--10 wt% B 2 O 3 and 20--40 wt% waste oxide loading); impacts of restricting the processing temperature to 1,150 C on eventual glass volume; and effects of caustic washing on any of the selected tank wastes relative to glass volume

  5. Biotransformation of Organic Waste into High Quality Fertilizer

    DEFF Research Database (Denmark)

    Bryndum, Sofie

    Agriculture faces several challenges of future provision of nutrients such as limited P reserves and increasing prices of synthetic fertilizers and recycling of nutrients from organic waste can be an important strategy for the long-term sustainability of the agricultural systems. Organically...... and S, is often low; and (3) the unbalanced composition of nutrients rarely matches crop demands. Therefore the objective of this project was to investigate the potential for (1) recycling nutrients from agro-industrial wastes and (2) compost biotransformation into high-quality organic fertilizers...... other uses into fertilizer use would be unlikely. An estimated ~50 % of the total organic waste pool, primarily consisting of animal manure and waste from the processing of sugar cane, coffee, oil palm and oranges, is currently being re-used as “fertilizers”, meaning it is eventually returned...

  6. Immobilization of high-level wastes into sintered glass: 1

    International Nuclear Information System (INIS)

    Russo, D.O.; Messi de Bernasconi, N.; Audero, M.A.

    1987-01-01

    In order to immobilize the high-level radioactive wastes from fuel elements reprocessing, borosilicate glass was adopted. Sintering experiments are described with the variety VG 98/12 (SiO 2 , TiO 2 , Al 2 O 3 , B 2 O 3 , MgO, CaO and Na 2 O) (which does not present devitrification problems) mixed with simulated calcinated wastes. The hot pressing line (sintering under pressure) was explored in two variants 1: In can; 2: In graphite matrix with sintered pellet extraction. With scanning electron microscopy it is observed that the simulated wastes do not disolve in the vitreous matrix, but they remain dispersed in the same. The results obtained point out that the leaching velocities are independent from the density and from the matrix type employed, as well as from the fact that the wastes do no dissolve in the matrix. (M.E.L.) [es

  7. Development of high integrity containers for rad-waste treatment

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yung Chul; Cho, Myung Sug; Jung, Yun Sub [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

    1995-12-31

    Nuclear power plants are generating rad waste such as solid wastes, concentrated liquid wastes, spent resins and spent filters, and various types of imported containers which have different specifications and material properties are employed to handle the rad wastes according to facility characteristics of the plants or the type of wastes. These containers are stored at the intermediate storage facilities at the plant site due to the construction delay of permanent disposal site, and the additional construction of storage and disposal sites become more difficult with increase of the numbers and the operation time of the plants. In order to solve these difficulties, rad wastes volume reduction facilities such as High Pressure Compression Facility or Drying Facility are being installed and use of High Integrity Containers(HIC) are increasing. Therefore, we decide quality and technology standards required for the HIC, and then develop the HIC which satisfies the standards with new composite material called Steel Fiber Polymer Impregnated Concrete(SFPIC) (author). 84 refs., 118 figs.

  8. Development of high integrity containers for rad-waste treatment

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yung Chul; Cho, Myung Sug; Jung, Yun Sub [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

    1996-12-31

    Nuclear power plants are generating rad waste such as solid wastes, concentrated liquid wastes, spent resins and spent filters, and various types of imported containers which have different specifications and material properties are employed to handle the rad wastes according to facility characteristics of the plants or the type of wastes. These containers are stored at the intermediate storage facilities at the plant site due to the construction delay of permanent disposal site, and the additional construction of storage and disposal sites become more difficult with increase of the numbers and the operation time of the plants. In order to solve these difficulties, rad wastes volume reduction facilities such as High Pressure Compression Facility or Drying Facility are being installed and use of High Integrity Containers(HIC) are increasing. Therefore, we decide quality and technology standards required for the HIC, and then develop the HIC which satisfies the standards with new composite material called Steel Fiber Polymer Impregnated Concrete(SFPIC) (author). 84 refs., 118 figs.

  9. Transportation packagings for high-level wastes and unprocessed transuranic wastes

    International Nuclear Information System (INIS)

    Wilmot, E.L.; Romesberg, L.E.

    1982-01-01

    Packagings used for nuclear waste transport are varied in size, shape, and weight because they must accommodate a wide variety of waste forms and types. However, this paper will discuss the common characteristics among the packagings in order to provide a broad understanding of packaging designs. The paper then discusses, in some detail, a design that has been under development recently at Sandia National Laboratories (SNL) for handling unprocessed, contact-handled transuranic (CHTRU) wastes as well as a cask design for defense high-level wastes (HLW). As presently conceived, the design of the transuranic package transporter (TRUPACT) calls for inner and outer boxes that are separated by a rigid polyurethane foam. The inner box has a steel frame with stainless steel surfaces; the outer box is similarly constructed except that carbon steel is used for the outside surfaces. The access to each box is through hinged doors that are sealed after loading. To meet another waste management need, a cask is being developed to transport defense HLW. The cask, which is at the preliminary design stage, is being developed by General Atomic under the direction of the TTC. The cask design relies heavily on state-of-the-art spent-fuel cask designs though it can be much simpler due to the characteristics of the HLW. A primary purpose of this paper is to show that CHTRU waste and defense HLW currently are and will be transported in packagings designed to meet the hazards of transportation that are present in general commerce

  10. Adsorption of Safranin-T from wastewater using waste materials- activated carbon and activated rice husks.

    Science.gov (United States)

    Gupta, Vinod K; Mittal, Alok; Jain, Rajeev; Mathur, Megha; Sikarwar, Shalini

    2006-11-01

    Textile effluents are major industrial polluters because of high color content, about 15% unfixed dyes and salts. The present paper is aimed to investigate and develop cheap adsorption methods for color removal from wastewater using waste materials activated carbon and activated rice husk-as adsorbents. The method was employed for the removal of Safranin-T and the influence of various factors such as adsorbent dose, adsorbate concentration, particle size, temperature, contact time, and pH was studied. The adsorption of the dye over both the adsorbents was found to follow Langmuir and Freundlich adsorption isotherm models. Based on these models, different useful thermodynamic parameters have been evaluated for both the adsorption processes. The adsorption of Safranin-T over activated carbon and activated rice husks follows first-order kinetics and the rate constants for the adsorption processes decrease with increase in temperature.

  11. Overview of the US program for developing a waste disposal system for spent nuclear fuel and high-level waste

    International Nuclear Information System (INIS)

    Kay, C.E.

    1988-01-01

    Safe disposal of spent nuclear fuel and radioactive high-level waste (HLW) has been a matter of national concern ever since the first US civilian nuclear reactor began generating electricity in 1957. Based on current projections of commercial generating capacity, by the turn of the century, there will be >40,000 tonne of spent fuel in the Untied States. In addition to commercial spent fuel, defense HLW is generated in the United States and currently stored at three US Department of Energy (DOE) sites: The Nuclear Waste Policy Amendments Act of 1987 provided for financial incentives to host a repository or a monitored retrievable storage (MRS) facility; mandated the areas in which DOE's siting efforts should concentrate (Yucca Mountain, Nevada); required termination of site-specific activities at other sites; required a resisting process for an MRS facility, which DOE had proposed as an integral part of the waste disposal system; terminated all activities for identifying candidates for a second repository; established an 11-member Nuclear Waste Technical Review Board; established a three-member MRS commission to be appointed by heads of the US Senate and House; directed the President to appoint a negotiator to seek a state or Indian tribe willing to host a repository or MRS facility at a suitable site and to negotiate terms and conditions under which the state or tribe would be willing to host such a facility; and amended, adjusted, or established other requirements contained in the 1982 law

  12. Wastes from former mining and milling activities in Tajikistan

    International Nuclear Information System (INIS)

    Mirsaidov, U.M.

    2012-01-01

    This article is devoted to wastes from former mining and milling activities in Tajikistan. Currently, the serious radiological and ecological problems in Tajikistan are uranium mining and milling activities consequences overcoming which intensively developed during the soviet period. After the collapse of USSR, the uranic ores extraction in Tajikistan stopped due to deposit's output completion on the territory of the republic. Remediation of mining and milling activities' sites became the most urgent once all mines were closed.

  13. High-level radioactive waste in Canada. Background paper

    International Nuclear Information System (INIS)

    Fawcett, R.

    1993-11-01

    The disposal of radioactive waste is one of the most challenging environmental problems facing Canada today. Since the Second World War, when Canadian scientists first started to investigate nuclear reactions, there has been a steady accumulation of such waste. Research reactors built in the early postwar years produced small amounts of radioactive material but the volume grew steadily as the nuclear power reactors constructed during the 1960s and 1970s began to spawn used fuel bundles. Although this radioactive refuse has been safely stored for the short term, no permanent disposal system has yet been fully developed and implemented. Canada is not alone in this regard. A large number of countries use nuclear power reactors but none has yet put in place a method for the long-term disposal of the radioactive waste. Scientists and engineers throughout the world are investigating different possibilities; however, enormous difficulties remain. In Canada, used fuel bundles from nuclear reactors are defined as high-level waste; all other waste created at different stages in the nuclear fuel cycle is classified as low-level. Although disposal of low-level waste is an important issue, it is a more tractable problem than the disposal of high-level waste, on which this paper will concentrate. The paper discusses the nuclear fuel waste management program in Canada, where a long-term disposal plan has been under development by scientists and engineers over the past 15 years, but will not be completed for some time. Also discussed are responses to the program by parliamentary committees and aboriginal and environmental groups, and the work in the area being conducted in other countries. (author). 1 tab

  14. High-level radioactive waste in Canada. Background paper

    Energy Technology Data Exchange (ETDEWEB)

    Fawcett, R [Library of Parliament, Ottawa, ON (Canada). Science and Technology Div.

    1993-11-01

    The disposal of radioactive waste is one of the most challenging environmental problems facing Canada today. Since the Second World War, when Canadian scientists first started to investigate nuclear reactions, there has been a steady accumulation of such waste. Research reactors built in the early postwar years produced small amounts of radioactive material but the volume grew steadily as the nuclear power reactors constructed during the 1960s and 1970s began to spawn used fuel bundles. Although this radioactive refuse has been safely stored for the short term, no permanent disposal system has yet been fully developed and implemented. Canada is not alone in this regard. A large number of countries use nuclear power reactors but none has yet put in place a method for the long-term disposal of the radioactive waste. Scientists and engineers throughout the world are investigating different possibilities; however, enormous difficulties remain. In Canada, used fuel bundles from nuclear reactors are defined as high-level waste; all other waste created at different stages in the nuclear fuel cycle is classified as low-level. Although disposal of low-level waste is an important issue, it is a more tractable problem than the disposal of high-level waste, on which this paper will concentrate. The paper discusses the nuclear fuel waste management program in Canada, where a long-term disposal plan has been under development by scientists and engineers over the past 15 years, but will not be completed for some time. Also discussed are responses to the program by parliamentary committees and aboriginal and environmental groups, and the work in the area being conducted in other countries. (author). 1 tab.

  15. CIGeO geological disposal for high-level radioactive waste in France

    International Nuclear Information System (INIS)

    Ouzounian, Gerald; Bolia, Jelana

    2014-01-01

    Andra is the sole French organization responsible for the radioactive waste management in the country. Its work relies extensively on the legal basis provided by several major laws (Waste Act of 1991 and the Planning Act of 2006), which shaped the main principles of the waste management strategy and determined the corresponding implementation tools. Andra's industrial activities are essentially based around three of its national disposal facilities. Two of these operational facilities, by their design and comprehensive monitoring system, are considered worldwide as solid and proven reference solutions for the concerned types of radioactive waste. Andra is also charged with designing a future deep geological repository for intermediate-level long-lived and high-level waste and researching potential management and disposal solutions for the graphite and radium-bearing waste. The purpose of this article is to update the information to the readers about the Cigeo geological disposal project for high-level radioactive waste in France (authors)

  16. Waste management assessment and technical review programme. WATRP. An international peer review service for radioactive waste management activities

    International Nuclear Information System (INIS)

    1994-09-01

    International Atomic Energy Agency provides international peer review services in radioactive waste management to those Member States that have established radioactive waste management programmes. Such services are provided within Waste Management Assessment and Technical Review Programme (WATRP). The main objective of WATRP is to provide international expertise and information on a requested subject in the field of radioactive waste management and to validate that programmes and activities are sound and performing well. Refs, figs and tabs

  17. Characterization of selected waste tanks from the active LLLW system

    International Nuclear Information System (INIS)

    Keller, J.M.; Giaquinto, J.M.; Griest, W.H.

    1996-08-01

    From September 1989 through January of 1990, there was a major effort to sample and analyze the Active Liquid-Low Level Waste (LLLW) tanks at ORNL which include the Melton Valley Storage Tanks (MVST) and the Bethel Valley Evaporator Service Tanks (BVEST). The purpose of this report is to summarize additional analytical data collected from some of the active waste tanks from November 1993 through February 1996. The analytical data for this report was collected for several unrelated projects which had different data requirements. The overall analyte list was similar for these projects and the level of quality assurance was the same for all work reported. the new data includes isotopic ratios for uranium and plutonium and an evaluation of the denature ratios to address criticality concerns. Also, radionuclides not previously measured in these waste tanks, including 99Tc and 237Np, are provided in this report

  18. Final repositories for high-level radioactive waste; Endlagerung hochradioaktiver Abfaelle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-10-15

    The brochure on final repositories for high-level radioactive waste covers the following issues: What is the origin of radioactive wastes? How large are the waste amounts? What is going to happen with the wastes? What is the solution for the Waste disposal? A new site search is started. Safety requirements for the final disposal of high-level radioactive wastes. Comparison of host rocks. Who is responsible and who will pay? Final disposal of high-level radioactive wastes worldwide. Short summary: History of the search for a final repository for high-level radioactive wastes in Germany.

  19. Spent fuel and high-level radioactive waste storage

    International Nuclear Information System (INIS)

    Trigerman, S.

    1988-06-01

    The subject of spent fuel and high-level radioactive waste storage, is bibliographically reviewed. The review shows that in the majority of the countries, spent fuels and high-level radioactive wastes are planned to be stored for tens of years. Sites for final disposal of high-level radioactive wastes have not yet been found. A first final disposal facility is expected to come into operation in the United States of America by the year 2010. Other final disposal facilities are expected to come into operation in Germany, Sweden, Switzerland and Japan by the year 2020. Meanwhile , stress is placed upon the 'dry storage' method which is carried out successfully in a number of countries (Britain and France). In the United States of America spent fuels are stored in water pools while the 'dry storage' method is still being investigated. (Author)

  20. Production and properties of solidified high-level waste

    International Nuclear Information System (INIS)

    Brodersen, K.

    1980-08-01

    Available information on production and properties of solidified high-level waste are presented. The review includes literature up to the end of 1979. The feasibility of production of various types of solidified high-level wast is investigated. The main emphasis is on borosilicate glass but other options are also mentioned. The expected long-term behaviour of the materials are discussed on the basis of available results from laboratory experiments. Examples of the use of the information in safety analysis of disposal in salt formations are given. The work has been made on behalf of the Danish utilities investigation of the possibilities of disposal of high-level waste in salt domes in Jutland. (author)