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Sample records for liquid radioactive releases

  1. PWR-GALE, Radioactive Gaseous Release and Liquid Release from PWR

    International Nuclear Information System (INIS)

    Chandrasekaran, T.; Lee, J.Y.; Willis, C.A.

    1988-01-01

    1 - Description of program or function: The PWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from pressurized water reactors (PWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment. 2 - Method of solution: GALE calculates expected releases based on 1) standardized coolant activities derived from ANS Standards 18.1 Working Group recommendations, 2) release and transport mechanisms that result in the appearance of radioactive material in liquid and gaseous waste streams, 3) plant-specific design features used to reduce the quantities of radioactive materials ultimately released to the environs, and 4) information received on the operation of nuclear power plants. 3 - Restrictions on the complexity of the problem: The liquid release portion of GALE uses subroutines taken from the ORIGEN (CCC-217) to calculate radionuclide buildup and decay during collection, processing, and storage of liquid radwaste. Memory requirements for this part of the program are determined by the large nuclear data base accessed by these subroutines

  2. Monitoring of released radioactive gaseous and liquid effluent at Rokkasho Reprocessing Plant

    International Nuclear Information System (INIS)

    Oka, M.; Keta, S.; Nagai, S.; Kano, M.; Ishihara, N.; Moriyama, T.; Ogaki, K.; Noda, K.

    2009-01-01

    Rokkasho Reprocessing Plant (RRP) Rokkasho Reprocessing Plant started its active tests with spent fuel at the end of March 2006. When spent fuels are sheared and dissolved, radioactive gaseous effluent and radioactive liquid effluent such as krypton-85, tritium, etc. are released into the environment. In order to limit the public dose as low as reasonably achievable in an efficient way, RRP removes radioactive material by evaporation, rinsing, filtering, etc., and then releases it through the main stack and the sea discharge pipeline that allow to make dispersion and dilution very efficiently. Also, concerning the radioactive gaseous and liquid effluent to be released into the environment, the target values of annual release have been defined in the Safety Rule based on the estimated annual release evaluated at the safety review of RRP. By monitoring the radioactive material in gaseous exhaust and liquid effluent RRP controls it not to exceed the target values. RRP reprocessed 430 tUpr of spent fuel during Active Test (March 2006 to October 2008). In this report, we report about: The outline of gaseous and liquid effluent monitoring. The amount of radioactive gaseous and liquid effluent during the active test. The performance of removal of radioactive materials in gaseous and liquid effluents. The impact on the public from radioactive effluents during the active test. (author)

  3. Study on radioactive release of gaseous and liquid effluents during normal operation of AP1000

    International Nuclear Information System (INIS)

    Gong Quan; Zhou Jing; Liu Yu

    2014-01-01

    The gaseous and liquid radioactive releases of pressurized water reactors plant during normal operation are an important content of environmental impact assessment and play a significant role in the design of nuclear power plant. According to the design characters of AP1OOO radioactive waste management system and the study on the calculation method and the release pathways, the calculation model of the gaseous and liquid radioactive releases during normal operation for AP1OOO are established. Base on the established calculation model and the design parameters of AP1000, the expected value of gaseous and liquid radioactive releases of AP1OOO is calculated. The results of calculation are compared with the limits in GB 6249-2011 and explain the adder that is included tu account for anticipated operational occurrences, providing a reference for environmental impact assessment of pressurized water reactor. (authors)

  4. Environmental system applied to radioactive liquid effluent release

    International Nuclear Information System (INIS)

    Nisti, Marcelo Bessa

    2009-01-01

    The current environmental administration considers the productive activity as an environmental system, defined as a group of processes, interactions, parameters and factors involved in the production. This mastering dissertation evaluated the release of the liquid radioactive effluents at Instituto de Pesquisas Energeticas e Nucleares (IPEN), under a systemic environmental study. The study evaluated the source term at IPEN in the period from 2004 to 2008, making use of gamma-ray and alpha spectrometry, instrumental neutron activation analysis, liquid phase scintillation and atomic absorption spectrometry. The employed methodologies were verified using samples from the Intercomparison National Program - PNI/IRD and Reference Materials. The facilities that contributed the most in these releases were the Radiopharmaceutical Center (CR) and the Research Reactor Center (CRPq) with an average of 11,4% and 87,4%, respectively, relative to the present radioactive activity. The sewer system releases were within the radioactive protection regulations, showing the effectiveness of IPEN's Radioactive Effluents Monitoring Program. The concentration of the stable elements Ag, Cd, Cr, Fe, Mn, Ni, Pb and Zn was determined in the liquid effluent in ali the samples from the storage tanks TR1 and CR in the period from 2004 to 2008 and in some of the samples of other IPEN's facilities in the period from 2004 to 2007. Among the analyzed effluents, two samples were higher than the stable elements discharge standards established in the state of Sao Paulo, one sample was higher than the required value of the element cadmium and the other higher than required value of the element zinco The storage tank TR1 discharge flow was estimated in 10,9 ± 0,9 m3 h -1 . The dilution factor at discharge point E1 was estimated using a radiotracers the isotopes 3 H, 137 CS, 60 Co, 54 Mn and 65 Zn, which are commonly released into IPEN's sewer system. The executed radiotracer study was carried out

  5. Assessment of Radioactive Liquid Effluents Release at IPEN-CNEN/SP

    International Nuclear Information System (INIS)

    Bessa Nisti, Marcelo; Godoy dos Santos, Adir Janete

    2008-01-01

    A continuous effluent monitoring program has been established at IPEN's plant in order to allow an environmental impact assessment due to radioactive liquid effluent discharge to sanitary system. Representative samples of radioactive liquid effluents are analyzed by using high resolution gamma spectroscopy and instrumental neutron activation analysis, facing to Brazilian radioprotection regulatory rules. The results are consolidating yearly in the Institute source-term. In this paper, results of the source-term are presented, concerning to years 2004, 2005 and 2006. The total activity discharged was 8.5xl0 8 Bq, 5.7x10 8 Bq and 2.7xl0 8 Bq, respectively. As the release is strongly dependent on the total amount of the effluent and on the dilution factor, special attention is needed in order to obtain the correct value of that last one. The estimated inside plant dilution factor, considering the recent facilities and the reshaping of the sewerage system was 80, 180 and 130, for period of 2004, 2005 and 2006 discharged liquid radioactive effluent

  6. Calculation of releases of radioactive materials in gaseous and liquid effluents from boiling water reactors (BWR-GALE Code)

    International Nuclear Information System (INIS)

    Bangart, R.L.; Bell, L.G.; Boegli, J.S.; Burke, W.C.; Lee, J.Y.; Minns, J.L.; Stoddart, P.G.; Weller, R.A.; Collins, J.T.

    1978-12-01

    The calculational procedures described in the report reflect current NRC staff practice. The methods described will be used in the evaluation of applications for construction permits and operating licenses docketed after January 1, 1979, until this NUREG is revised as a result of additional staff review. The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment

  7. Processing method for radioactive liquid waste

    International Nuclear Information System (INIS)

    Yasumura, Keijiro

    1991-01-01

    Drainages, such as water after used for washing operators' clothes and water used for washing hands and for showers have such features that the radioactive concentration is extremely low and detergent ingredients and insoluble ingredients such as waste threads, hairs and dirts are contained. At present, waste threads are removed by a strainer. Then, after measuring the radioactivity and determining that the radioactivity is less than a predetermined concentration, they are released to circumstances. However, various organic ingredients such as detergents and dirts in the liquid wastes are released as they are and it is not preferred in respect of environmental protection. Then, in the present invention, activated carbon is filled in a container orderly so that the diameter of the particles of the activated carbon is increased in the upper layer and decreased in the lower layer, and radioactive liquid wastes are passed through the container. With such a constitution. Both of soluble substances and insoluble substances can be removed efficiently without causing cloggings. (T.M.)

  8. Decree No 74-1181 of 31 December 1974 concerning liquid radioactive effluent releases from nuclear installations

    International Nuclear Information System (INIS)

    1975-01-01

    This Decree prescribes the licensing procedure for the release of liquid wastes from nuclear installations as well as the technical supervision of such operations. It does not apply to the transport of radioactive effluents which is governed by the regulations on the transport of dangerous goods. (NEA) [fr

  9. A high-performance liquid chromatography method for the serotonin release assay is equivalent to the radioactive method.

    Science.gov (United States)

    Sono-Koree, N K; Crist, R A; Frank, E L; Rodgers, G M; Smock, K J

    2016-02-01

    The serotonin release assay (SRA) is considered the gold standard laboratory test for heparin-induced thrombocytopenia (HIT). The historic SRA method uses platelets loaded with radiolabeled serotonin to evaluate platelet activation by HIT immune complexes. However, a nonradioactive method is desirable. We report the performance characteristics of a high-performance liquid chromatography (HPLC) SRA method. We validated the performance characteristics of an HPLC-SRA method, including correlation with a reference laboratory using the radioactive method. Serotonin released from reagent platelets was quantified by HPLC using fluorescent detection. Results were expressed as % release and classified as positive, negative, or indeterminate based on previously published cutoffs. Serum samples from 250 subjects with suspected HIT were tested in the HPLC-SRA and with the radioactive method. Concordant classifications were observed in 230 samples (92%). Sera from 41 healthy individuals tested negative. Between-run imprecision studies showed standard deviation of performance characteristics, equivalent to the historic radioactive method, but avoids the complexities of working with radioactivity. © 2015 John Wiley & Sons Ltd.

  10. Monitoring for radioactive materials releasing to environment in M310 reformatived nuclear power plant

    International Nuclear Information System (INIS)

    Yin Zhenyu; Yang Guangli; Xu Guang

    2012-01-01

    Airborne radioactive materials of nuclear power plant (NPP) releases to the environment from the stack of NPP. Radioactive liquid waste releases of the ocean, the fluvial and the lake through the liquid waste letdyke of NPP. Further more, a few radioactive waste may be taken out of the NPP by vehicle or personnel. For the purpose of strict management and control above-mentioned waste, we use detect equipment monitoring radioactive waste of NPP. Management and control for the releasing of radioactive material to the environment in M310 reformatived NPP is strict and safety. (authors)

  11. Sampling and characterization of radioactive liquid wastes

    International Nuclear Information System (INIS)

    Zepeda R, C.; Monroy G, F.; Reyes A, T.; Lizcano, D.; Cruz C, A. C.

    2017-09-01

    To define the management of radioactive liquid wastes stored in 200 L drums, its isotope and physicochemical characterization is essential. An adequate sampling, that is, representative and homogeneous, is fundamental to obtain reliable analytical results, therefore, in this work, the use of a sampling mechanism that allows collecting homogenous aliquots, in a safe way and minimizing the generation of secondary waste is proposed. With this mechanism, 56 drums of radioactive liquid wastes were sampled, which were characterized by gamma spectrometry, liquid scintillation, and determined the following physicochemical properties: ph, conductivity, viscosity, density and chemical composition by gas chromatography. 67.86% of the radioactive liquid wastes contains H-3 and of these, 47.36% can be released unconditionally, since it presents activities lower than 100 Bq/g. 94% of the wastes are acidic and 48% have viscosities <50 MPa s. (Author)

  12. Membrane Treatment of Liquid Salt Bearing Radioactive Wastes

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  13. Storage of solid and liquid radioactive material

    International Nuclear Information System (INIS)

    Matijasic, A.; Gacinovic, O.

    1961-01-01

    Solid radioactive waste collected during 1961 from the laboratories of the Institute amounted to 22.5 m 3 . This report contains data about activity of the waste collected from january to November 1961. About 70% of the waste are short lived radioactive material. Material was packed in metal barrels and stored in the radioactive storage in the Institute. There was no contamination of the personnel involved in these actions. Liquid radioactive wastes come from the Isotope production laboratory, laboratories using tracer techniques, reactor cooling; decontamination of the equipment. Liquid wastes from isotope production were collected in plastic bottles and stored. Waste water from the RA reactor were collected in special containers. After activity measurements this water was released into the sewage system since no activity was found. Table containing data on quantities and activity of radioactive effluents is included in this report

  14. Radioactive materials released from nuclear power plants

    International Nuclear Information System (INIS)

    Tichler, J.; Norden, K.; Congemi, J.

    1991-05-01

    Releases of radioactive materials in airborne and liquid effluents from commercial light water reactors during 1988 have been compiled and reported. Data on solid waste shipments as well as selected operating information have been included. This report supplements earlier annual reports issued by the former Atomic Energy Commission and the Nuclear Regulatory Commission. The 1988 release data are summarized in tabular form. Data covering specific radionuclides are summarized. 16 tabs

  15. Radioactive materials released from nuclear power plants (1976)

    International Nuclear Information System (INIS)

    Decker, T.R.

    1978-11-01

    Releases of radioactive materials in airborne and liquid effluents from commercial light water reactors during 1976 have been compiled and reported. Data on solid waste shipments as well as selected operating information have been included. This report supplements earlier annual reports issued by the former Atomic Energy Commission and the Nuclear Regulatory Commission. The 1976 release data are compared with previous year releases in tabular form. Data covering specific radionuclides are summarized

  16. Prevention and mitigation of groundwater contamination from radioactive releases

    International Nuclear Information System (INIS)

    1988-10-01

    This document gives basic information on potential pathways and mechanisms, by which radioactive materials from releases can reach man, and on modelling considerations to predict the behaviour of radioactive materials in the ground. The main objective is to present an overview of existing techniques for preventing the offsite releases of contaminants into the groundwater systems and techniques for mitigation of effects of such releases should they occur. The recommended techniques are fully applicable to any hazardous materials, such as organic liquids, and toxic materials or otherwise dangerous materials, the presence of which in the accessible biosphere can represent health risks as well as economic losses to the general public. 11 refs, 2 figs, 8 tabs

  17. Radioactive liquid waste filtering device

    International Nuclear Information System (INIS)

    Inami, Ichiro; Tabata, Masayuki; Kubo, Koji.

    1988-01-01

    Purpose: To prevent clogging in filter materials and improve the filtration performance for radioactive liquid wastes without increasing the amount of radioactive wastes. Constitution: In a radioactive waste filtering device, a liquid waste recycling pipe and a liquid recycling pump are disposed for recycling the radioactive liquid wastes in a liquid wastes vessel. In this case, the recycling pipe and the recycling pump are properly selected so as to satisfy the conditions capable of making the radioactive liquid wastes flowing through the pipe to have the Reynolds number of 10 4 - 10 5 . By repeating the transportation of radioactive liquid wastes in the liquid waste vessel through the liquid waste recycling pipe by the liquid waste recycling pump and then returning them to the liquid waste vessel again, particles of fine grain size in the suspended liquids are coagulated with each other upon collision to increase the grain size of the suspended particles. In this way, clogging of the filter materials caused by the particles of fine grain size can be prevented, thereby enabling to prevent the increase in the rising rate of the filtration differential pressure, reduce the frequency for the occurrence of radioactive wastes such as filter sludges and improve the processing performance. (Kamimura, M.)

  18. Release procedures and disposal of radioactive residual substances at the Medical University Hanover (MHH)

    International Nuclear Information System (INIS)

    Scheller, F.; Behrendt, R.; Harke, H.

    2005-01-01

    The disposal of all radioactive residual packages of the MHH is regulated by the German radiation protection ordinance from 26 th July 2001; shown in appendix III, table 1, column 5: unrestricted release of solid materials and liquids. All radioactive waste packages are collected and handled by the central MHH department for radiation protection. They are sorted for type, nuclide and specific radioactivity. A few packages can directly be released as conventional waste after performing incoming measurements showing very low activity concentrations. Longer living radionuclides with specific activities above the release limits have to be delivered to the national waste disposal. For direct measurements of gamma-emitting radionuclides we use two in-situ-measuring units (ISOCS, Canberra) and a special release unit with 9 plastic scintillators in a fixed geometry (FR-9 PVT, MED Medizintechnik). Fluid beta-emitting radionuclides are measured by taking a fraction and using liquid scintillation counting (LSC). Solid beta-emitting radionuclides activity are calculated. (orig.)

  19. Order of the 10 january 2003 authorizing the national agency for the radioactive wastes management to follow the gaseous and liquid effluents release for the exploitation of the radioactive wastes storage center of the Manche

    International Nuclear Information System (INIS)

    2003-01-01

    This document, took out from the Official Journal, is the law text relative to the order of the 10 january 2003 authorizing the national agency for the radioactive wastes management to follow the gaseous and liquid effluents release for the exploitation of the radioactive wastes storage center of the Manche. (A.L.B.)

  20. Radioactive materials released from nuclear power plants. Annual report 1977

    International Nuclear Information System (INIS)

    Decker, T.R.

    1978-11-01

    Releases of radioactive materials in airborne and liquid effluents from commercial light water reactors during 1977 have been compiled and reported. Data on solid waste shipments as well as selected operating information have been included. This report supplements earlier annual reports issued by the former Atomic Energy Commission and the Nuclear Regulatory Commission. The 1977 release data are compared with previous years releases in tabular form. Data covering specific radionuclides are summarized

  1. Radioactive materials released from nuclear power plants: Annual report, 1984

    Energy Technology Data Exchange (ETDEWEB)

    Tichler, J.; Norden, K.; Congemi, J.

    1987-08-01

    Releases of radioactive materials in airborne and liquid effluents from commercial light water reactors during 1984 have been compiled and reported. Data on solid waste shipments as well as selected operating information have been included. This report supplements earlier annual reports issued by the former Atomic Energy Commission and the Nuclear Regulatory Commission. The 1984 release data are summarized in tabular form. Data covering specific radionuclides are summarized.

  2. Releases of radioactivity at the Savannah River Plant, 1954-1980

    Energy Technology Data Exchange (ETDEWEB)

    Ashley, C.; Zeigler, C. C.; Culp, P. A.

    1982-01-01

    This report contains summaries of releases of radioactivity to onsite seepage basins and to plant effluents for each year since plant startup (1954 through 1980). Releases for the years 1954 through 1959 were reassessed in 1974 and assigned release values for specific long-lived radionuclides. These long-lived radionuclides (half-lives greater than 1 year) are the only radionuclides included for the years 1954 through 1970. Since 1970 all detectable radionuclides have been included. Measured migration o radionuclides from F-, H-, and K-Area seepage basins and desorption of /sup 137/Cs from the Four Mile Creek stream bed are included in Summary B of this report. Summaries of monthly releases for 1979 and 1980 and total SRP releases by radionuclide in three categories: liquid to seepage basins, liquid to streams, and atmospheric are also included. Monthly releases for 1979 are included because the 1979 issue of this report was not published. All 1979 data are included in this report.

  3. Releases of radioactivity from uranium mills and effluent treatment costs

    International Nuclear Information System (INIS)

    Witherspoon, J.P.; Sears, M.B.; Blanco, R.E.

    1977-01-01

    Airborne releases of radioactive materials from uranium milling to the environment consist of ore dust, yellowcake dust, tailings dust, and radon gas while the mill is active. After a mill has ceased operations, tailings may be stabilized to minimize or prevent airborne releases of radioactive particulates. However, radon gas will continue to be released in amounts inversely proportional to the degree of stabilization treatment (and expense). Liquid waste disposal is by evaporation and natural seepage to the ground beneath the tailings impoundment area. The release of radioactive materials (and potential radiation exposures) determines the majority of costs associated with minimizing the environmental impact of uranium milling. Radwaste treatments to reduce estimated radiation doses to individuals to 3 to 5% of those received with current milling practices are equivalent to $0.66 per pounds of U 3 O 8 and 0.032 mill per kWhr of electricity. This cost would cover a high efficiency reverse jet bag filter and high energy venturi scrubbers for dusts, neutralization of liquids, and an asphalt-lined tailings basin with a clay core dam to reduce seepage. In addition, this increased cost would cover stabilization of tailings, after mill closure, with a 1-in. asphalt membrane topped by 2 ft of earth and 0.5 ft of crushed rock to provide protection against future leaching and wind erosion. The cost of reducing the radiological hazards associated with uranium milling to this degree would contribute about 0.4% to the current total cost of nuclear power

  4. Releases of radioactive substances from Swedish nuclear power plants (RAKU)

    Energy Technology Data Exchange (ETDEWEB)

    Ingemansson, T.; Bergstroem, C. [ALARA Engineering AB, Skultuna (Sweden)

    1997-04-01

    Releases of radioactivity to air and water from Swedish nuclear power plants have been studied and compared with those from foreign reactors. Averaged over the years from commissioning of the reactors to the last year data are available, the release of radioactive noble gas from the Swedish BWRs has been about the same as from comparable foreign reactors. The oldest Swedish BWRs, Oskarshamn 1 and 2 (O1 and O2) and Ringhals 1 (R1), have simple off-gas systems with only one delay volume. All BWRs in US, Germany, Japan and Switzerland are equipped with more sophisticated off-gas systems. It can be expected that O1, O2 and R1 therefore will have the highest release of noble gas activity at an international comparison if they do not modernize their off-gas system. BWRs in US, Germany and Japan are today equipped with recombiners and with one exception also charcoal columns. Japanese BWRs report zero releases to air. Releases of radioactivity to water after commissioning was about the same for most of the studied reactors. Some of the newest German plants have had low annual releases already at commissioning. Improvements of the treatment systems at old German, Swiss and US reactors have significantly lowered the releases. For most of the Swedish plants the annual releases to water have remained at the initial level. Forsmark 3 has succeeded in decreasing the release of radionuclides to water by a factor of almost one hundred compared to other Swedish reactors. Also O3 has managed to decrease the liquid effluents. Japanese plants have zero release of radioactivity excluding tritium to water. The release of tritium is about the same for all reactors of the same type in the world. 35 refs, 31 figs, 24 tabs.

  5. Sellafield (release of radioactivity)

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, J; Goodlad, A; Morris, M

    1986-02-06

    A government statement is reported, about the release of plutonium nitrate at the Sellafield site of British Nuclear Fuels plc on 5 February 1986. Matters raised included: details of accident; personnel monitoring; whether radioactive material was released from the site; need for public acceptance of BNFL activities; whether plant should be closed; need to reduce level of radioactive effluent; number of incidents at the plant.

  6. Liquid pathway generic study. Impacts of accidental radioactive releases to the hydrosphere from floating and land-based nuclear power plants

    International Nuclear Information System (INIS)

    1978-02-01

    The staff concludes that for representative sites, there are differences in the impacts of accidental radioactive releases to the hydrosphere between floating nuclear plants (FNPs) and land-based plants (LBPs) of the ice-condenser type. The consequences of releases from design basis accidents are found to be lower for an FNP than for an LBP. For core-melt events which we consider to be of very low probability, the staff results indicate that the expected liquid pathway consequences are higher at an FNP than at an LBP, and that interdiction at the site is not likely for an FNP. The staff considers this combination of differences in release magnitude and interdiction potential to be significant. An assessment of the overall significance of the total risk associated with potential releases to liquid and airborne pathways will be considered in a forthcoming environmental impact statement for the FNP. This environmental impact statement will include a consideration of the environmental, social, and economic impacts of the operation of the FNP, as well as a value-impact analysis of alternatives which may avoid or mitigate radiological impacts to the environment

  7. Radioactive liquid containing vessel

    International Nuclear Information System (INIS)

    Sakurada, Tetsuo; Kawamura, Hironobu.

    1993-01-01

    Cooling jackets are coiled around the outer circumference of a container vessel, and the outer circumference thereof is covered with a surrounding plate. A liquid of good conductivity (for example, water) is filled between the cooling jackets and the surrounding plate. A radioactive liquid is supplied to the container vessel passing through a supply pipe and discharged passing through a discharge pipe. Cooling water at high pressure is passed through the cooling water jackets in order to remove the heat generated from the radioactive liquid. Since cooling water at high pressure is thus passed through the coiled pipes, the wall thickness of the container vessel and the cooling water jackets can be reduced, thereby enabling to reduce the cost. Further, even if the radioactive liquid is leaked, there is no worry of contaminating cooling water, to prevent contamination. (I.N.)

  8. Radioactive liquid waste processing system

    International Nuclear Information System (INIS)

    Noda, Tetsuya; Kuramitsu, Kiminori; Ishii, Tomoharu.

    1997-01-01

    The present invention provides a system for processing radioactive liquid wastes containing laundry liquid wastes, shower drains or radioactive liquid wastes containing chemical oxygen demand (COD) ingredients and oil content generated from a nuclear power plant. Namely, a collecting tank collects radioactive liquid wastes. A filtering device is connected to the exit of the collective tank. A sump tank is connected to the exit of the filtering device. A powdery active carbon supplying device is connected to the collecting tank. A chemical fluid tank is connected to the collecting tank and the filtering device by way of chemical fluid injection lines. Backwarding pipelines connect a filtered water flowing exit of the filtering device and the collecting tank. The chemical solution is stored in the chemical solution tank. Then, radioactive materials in radioactive liquid wastes generated from a nuclear power plant are removed by the filtering device. The water quality standard specified in environmental influence reports can be satisfied. In the filtering device, when the filtering flow rate is reduced, the chemical fluid is supplied from the chemical fluid tank to the filtering device to recover the filtering flow rate. (I.S.)

  9. Radioactive materials released from nuclear power plants. Annual report, 1982. Volume 3

    International Nuclear Information System (INIS)

    Tichler, J.; Norden, K.

    1986-02-01

    Releases of radioactive materials in airborne and liquid effluents from commercial light water reactors during 1982 have been compiled and reported. Data on solid waste shipments as well as selected operating information have been included. This report supplements earlier annual reports issued by the former Atomic Energy Commission and the Nuclear Regulatory Commission. The 1982 release data are summarized in tabular form. Data covering specific radionuclides are summarized

  10. Method of processing radioactive liquid waste

    International Nuclear Information System (INIS)

    Motojima, Kenji; Kawamura, Fumio.

    1981-01-01

    Purpose: To increase the efficiency of removing radioactive cesium from radioactive liquid waste by employing zeolite affixed to metallic compound ferrocyanide as an adsorbent. Method: Regenerated liquid waste of a reactor condensation desalting unit, floor drain and so forth are collected through respective supply tubes to a liquid waste tank, and the liquid waste is fed by a pump to a column filled with zeolite containing a metallic compound ferrocyanide, such as with copper, zinc, manganese, iron, cobalt, nickel or the like. The liquid waste from which radioactive cesium is removed is dried and pelletized by volume reducing and solidifying means. (Yoshino, Y.)

  11. Effective equivalent dose in the critical group due to release of radioactive effluents

    International Nuclear Information System (INIS)

    Santos, John W.A. dos; Varandas, Luciana R.; Souza, Denise N.; Souza, Cristiano B.F.; Lima, Sandro Leonardo N.; Mattos, Marcos Fernando M.; Moraes, Jose Adenildo T.

    2005-01-01

    To ensure that the emissions of radioactive material by liquid and gaseous pathways are below applicable limits it is necessary to evaluate the effective equivalent dose in the critical group, which is a magnitude that takes into consideration the modeling used and the terms radioactive activity source. The calculation of this dose considers each radionuclide released by the activity of Nuclear plant, liquid and gaseous by, and the sum of the values obtained is controlled so that this dose does not exceed the goals of the regulatory body, the CNEN and the goals established by the Nuclear power plant. To hit these targets various controls are used such as: controls for effluent monitors instrumentation, environmental monitoring programs, effluent release controls and dose calculation in the environment. According to the findings, it is concluded that during the period of operation of the plants, this dose is below of the required limits

  12. Method of processing radioactive liquid wastes

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Y; Kikuchi, M; Funabashi, K; Yusa, H; Horiuchi, S

    1978-12-21

    Purpose: To decrease the volume of radioactive liquid wastes essentially consisting of sodium hydroxide and boric acid. Method: The concentration ratio of sodium hydroxide to boric acid by weight in radioactive liquid wastes essentially consisting of sodium hydroxide and boric acid is adjusted in the range of 0.28 - 0.4 by means of a pH detector and a sodium concentration detector. Thereafter, the radioactive liquid wastes are dried into powder and then discharged.

  13. International comparison on radioactive releases of PWR between 1975 and 1988

    International Nuclear Information System (INIS)

    Benedittini, M.; Tabare, M.

    1990-02-01

    This report presents the main results of a comparison on radioactive releases from 135 PWRs in the following countries: Belgium, Federal Republic of Germany, Finland, France, Japan, Sweden, Switzerland and the United States. For liquid releases, distinctions have been undertaken between tritium and no tritium activities and for gaseous releases between noble gases and halogens-aerosols. Data has been collected from literature published or by contacting utilities administrative staff or authorities in order to constitute a data base. As regards this comparison however, reactors, the commercial operation of which started prior to July 1974, have been excluded [fr

  14. Sampling and characterization of radioactive liquid wastes; Muestreo y caracterizacion de desechos liquidos radiactivos

    Energy Technology Data Exchange (ETDEWEB)

    Zepeda R, C.; Monroy G, F.; Reyes A, T.; Lizcano, D. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Cruz C, A. C., E-mail: carla.zepeda@inin.gob.mx [SEP, Instituto Tecnologico de Orizaba, Av. Oriente 9, Col. Emiliano Zapata, 94320 Orizaba, Veracruz (Mexico)

    2017-09-15

    To define the management of radioactive liquid wastes stored in 200 L drums, its isotope and physicochemical characterization is essential. An adequate sampling, that is, representative and homogeneous, is fundamental to obtain reliable analytical results, therefore, in this work, the use of a sampling mechanism that allows collecting homogenous aliquots, in a safe way and minimizing the generation of secondary waste is proposed. With this mechanism, 56 drums of radioactive liquid wastes were sampled, which were characterized by gamma spectrometry, liquid scintillation, and determined the following physicochemical properties: ph, conductivity, viscosity, density and chemical composition by gas chromatography. 67.86% of the radioactive liquid wastes contains H-3 and of these, 47.36% can be released unconditionally, since it presents activities lower than 100 Bq/g. 94% of the wastes are acidic and 48% have viscosities <50 MPa s. (Author)

  15. The regulation of radioactive effluent release in France (mainly from large nuclear installations)

    International Nuclear Information System (INIS)

    Hebert, Jean.

    1978-01-01

    In parallel with the licensing system for construction and operation of classified or so-called large nuclear installations (INB) there are in France regulations for the release of radioactive effuents from such installations. The regulations applicable to installations other than INBs are not specifically of a nuclear nature, while those covering INBs, which are analysed in this study, in particular, cover effluent release in liquid or gaseous form. The licensing and control procedures for such release are analysed in detail. (NEA) [fr

  16. Method of processing radioactive liquid waste

    International Nuclear Information System (INIS)

    Hasegawa, Akira; Kuribayashi, Hiroshi; Soda, Kenzo; Mihara, Shigeru.

    1988-01-01

    Purpose: To obtain satisfactory plastic solidification products rapidly and smoothly by adding oxidizers to radioactive liquid wastes. Method: Sulfuric acid, etc. are added to radioactive liquid wastes to adjust the pH value of the liquid wastes to less than 3.0. Then, ferrous sulfates are added such that the iron concentration in the liquid wastes is 100 mg/l. Then, after adjusting pH suitably to the drying powderization by adding alkali such as hydroxide, the liquid wastes are dried and powderized. The resultant powder is subjected to plastic solidification by using polymerizable liquid unsaturated polyester resins as the solidifying agent. The thus obtained solidification products are stable in view of the physical property such as strength or water proofness, as well as stable operation is possible even for those radioactive liquid wastes in which the content ingredients are unknown. (Takahashi, M.)

  17. The Comparison on Treatment Method of Liquid Radioactive Waste in Yonggwang No 3 and 4 and No 5 and 6

    International Nuclear Information System (INIS)

    Yeom, Yu Sun; Kim, Soong Pyung; Lee, Seung Jin

    2004-01-01

    Most of the low-level liquid radioactive wastes generated from PWR plants are classified into high or low total suspended solid(HTDS or LTDS), and into radiochemical and radioactive laundry waste. Although the evaporation process has a high decontamination ability, it has several problems such as corrosion, foam, and congestion. A new liquid waste disposal process using the ion-exchange demineralizer(IED), instead of the current evaporation process, has been introduced into the Yonggwang NPP No 5 and 6. These two methods have been compared to understand the differences in this study. Aspects compared here were the released radioactivity amount of the liquid radioactive wastes, the dose of off-site residents, the decontamination factor, and the amount of the solid radioactive wastes. The IED system is designed to discharge higher radioactivity about 20% than the evaporating system, and the actual radioactivity released from the evaporating and IED system were 0.473 mCi and 1.098 mCi, respectively. The radioactivity released from the IED was 2.32 times higher than that of the evaporating system. The dose of off-site residents was 2.97 x 10 -6 mSv for the evaporating system, and 6.47 x 10 -6 mSv for IED. The decontamination factor(DF) of the evaporator is, in most cases, far lower than the lower limits of detection(LLD) with the Ge-Li detector. Due to the low concentration of the liquid wastes collected from the liquid waste system, the decontamination factor of IED is very low. Since there is not enough data on the amount of solid radioactive wastes generated by the evaporation system, the comparison on these two systems has been conducted on the basis of the design, and the comparison result was that the evaporating system generated more wastes about 40% than IED.

  18. The Comparison on Treatment Method of Liquid Radioactive Waste in Yonggwang No 3 and 4 and No 5 and 6

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Yu Sun; Kim, Soong Pyung [Chosun University, Gwangju (Korea, Republic of); Lee, Seung Jin [RedTek CO., LTD., Daejeon (Korea, Republic of)

    2004-09-15

    Most of the low-level liquid radioactive wastes generated from PWR plants are classified into high or low total suspended solid(HTDS or LTDS), and into radiochemical and radioactive laundry waste. Although the evaporation process has a high decontamination ability, it has several problems such as corrosion, foam, and congestion. A new liquid waste disposal process using the ion-exchange demineralizer(IED), instead of the current evaporation process, has been introduced into the Yonggwang NPP No 5 and 6. These two methods have been compared to understand the differences in this study. Aspects compared here were the released radioactivity amount of the liquid radioactive wastes, the dose of off-site residents, the decontamination factor, and the amount of the solid radioactive wastes. The IED system is designed to discharge higher radioactivity about 20% than the evaporating system, and the actual radioactivity released from the evaporating and IED system were 0.473 mCi and 1.098 mCi, respectively. The radioactivity released from the IED was 2.32 times higher than that of the evaporating system. The dose of off-site residents was 2.97 x 10{sup -6} mSv for the evaporating system, and 6.47 x 10{sup -6} mSv for IED. The decontamination factor(DF) of the evaporator is, in most cases, far lower than the lower limits of detection(LLD) with the Ge-Li detector. Due to the low concentration of the liquid wastes collected from the liquid waste system, the decontamination factor of IED is very low. Since there is not enough data on the amount of solid radioactive wastes generated by the evaporation system, the comparison on these two systems has been conducted on the basis of the design, and the comparison result was that the evaporating system generated more wastes about 40% than IED.

  19. Method of processing low-level radioactive liquid wastes

    International Nuclear Information System (INIS)

    Matsunaga, Ichiro; Sugai, Hiroshi.

    1984-01-01

    Purpose: To effectively reduce the radioactivity density of low-level radioactive liquid wastes discharged from enriched uranium conversion processing steps or the likes. Method: Hydrazin is added to low-level radioactive liquid wastes, which are in contact with iron hydroxide-cation exchange resins prepared by processing strongly acidic-cation exchange resins with ferric chloride and aqueous ammonia to form hydrorizates of ferric ions in the resin. Hydrazine added herein may be any of hydrazine hydrate, hydrazine hydrochloride and hydranine sulfate. The preferred addition amount is more than 100 mg per one liter of the liquid wastes. If it is less than 100 mg, the reduction rate for the radioactivety density (procession liquid density/original liquid density) is decreased. This method enables to effectively reduce the radioactivity density of the low-level radioactive liquid wastes containing a trace amount of radioactive nucleides. (Yoshihara, H.)

  20. Survey of radioactive effluent releases from byproduct material facilities. Technical report

    International Nuclear Information System (INIS)

    Cook, J.R.

    1981-08-01

    A survey of over 3,000 NRC byproduct material licensees was conducted in late 1980 to collect data on annual effluent releases of radioactivity. The survey was conducted through a questionnaire, which was sent to NRC licensees who handle radioactive material in unsealed form, i.e., research, medical, and industrial institutions. Principal findings from the survey analysis are as follows: More than 98% of the reported annual releases to air (484 to 490) yield calculated average concentrations at the boundary of the unrestricted area that were at 1% or less than the maximum permissible concentration (MPC) of Appendix B, Table II, Column 1 of 10 CFR 20. The largest reported annual release was estimated to yield a concentration that was approximately 12% of MPC, the 5 other releases ranged from 1 to 10% of MPC. All reported annual releases of liquid waste were within the limits specified by NRC with most facilities reporting annual releases of only a fraction of a curie. Based on the data provided by licensees and analyzed in this report, it appears that in general the environmental impacts from research, medical and industrial institutions and organizations licensed by the NRC to possess and use byproduct materials are minimal and correspond to a small fraction of that from natural background

  1. Radioactive liquid waste processing device

    International Nuclear Information System (INIS)

    Murakami, Susumu; Kuroda, Noriko; Matsumoto, Hiroyo.

    1991-01-01

    The present device comprises a radioactive liquid wastes concentration means for circulating radioactive liquid wastes between each of the tank, a pump and a film evaporator thereby obtaining liquid concentrates and a distilled water recovery means for condensing steams separated by the film evaporator by means of a condenser. It further comprises a cyclizing means for circulating the resultant distilled water to the upstream after the concentration of the liquid concentrates exceeds a predetermined value or the quality of the distilled water reaches a predetermined level. Further, a film evaporator having hydrophilic and homogeneous films is used as a film evaporator. Then, the quality of the distilled water discharged from the present device to the downstream can always satisfy the predetermined conditions. Further, by conducting operation at high concentration while interrupting the supply of the processing liquids, high concentration up to the aimed concentration can be attained. Further, since the hydrophilic homogeneous films are used, carry over of the radioactive material accompanying the evaporation is eliminated to reduce the working ratio of the vacuum pump. (T.M.)

  2. Releases of radioactivity at the Savannah River Plant, 1954--1985

    International Nuclear Information System (INIS)

    Zeigler, C.C.; Lawrimore, I.B.

    1988-07-01

    Radioactive releases from Savannah River Plant (SRP) facilities to air, water and earthen seepage basins have been monitored and tabulated throughout the history of the site. The purpose of this report is to provide a source of data on routine releases of radioactivity to air, water and seepage basins that can be used for analyses of trends, environmental impact, etc. As used in this report, routine radioactive releases means radioactive materials that are released through established effluents from process facilities. This report provides a summary of radioactive releases that inflects the release values contained m records and documents from startup through 1985

  3. Type A radioactive liquid sample packaging family

    International Nuclear Information System (INIS)

    Edwards, W.S.

    1995-11-01

    Westinghouse Hanford Company (WHC) has developed two packagings that can be used to ship Type A quantities of radioactive liquids. WHC designed these packagings to take advantage of commercially available items where feasible to reduce the overall packaging cost. The Hedgehog packaging can ship up to one liter of Type A radioactive liquid with no shielding and 15 cm of distance between the liquid and the package exterior, or 30 ml of liquid with 3.8 cm of stainless steel shielding and 19 cm of distance between the liquid and the package exterior. The One Liter Shipper can ship up to one liter of Type A radioactive liquid that does not require shielding

  4. The minimization of radioactive releases to the sea from the Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

    Sugiyama, T.; Yoshikawa, K.; Ishii, T.

    1996-01-01

    The Tokai Reprocessing Plant (TRP) started hot operation in September 1977. The total amount of about 790 tU of spent fuel, generated in Japan, has been successfully reprocessed as of December 1994. Low-level liquid wastes have been treated safely with the low-level waste treatment process. The design of TRP was based on foreign technology. In the early stage of designing, the radioactivity released to the sea was estimated at approximately 2.6 TBq/day (70 Ci/day) for beta activity (except for tritium). Later, PNC added an evaporator to the process to reduce the level down to 1/100, i.e. 9.6 TBq/year (260 Ci/year) or 2.6 x 10 -2 TBq/day (0.7 Ci/day). In addition, under the supervision of the government, PNC started R and D to further decrease the radioactivity released to the sea in terms of ALARA. Aiming at reducing the activity from 9.6 TBq/year (260 Ci/year) to 1/10 of that value (i.e. 26 Ci/year), the release reduction technology development facility was added. This facility was incorporated into the low-level waste treatment process in 1980, before starting the regular operation of TRP. Since the fuel reprocessing commenced, total radioactivity discharged to the sea has been 1.9 x 10 -2 TBq (0.51 Ci) for beta activity, as of December 1994. Before incorporating the release reduction technology development facility, the yearly level was 3.7 x 10 -3 - 7.4 x 10 -3 TBq (0.1 - 0.2 Ci). After incorporation of the facility, radioactivity released to the sea was greatly decreased to non-detection levels in recent years, in spite of increasing annual reprocessing amounts. Although serious equipment failures have occurred such as the acid recovery evaporator and the dissolvers, there was no influence on radioactivity released to the sea. (author)

  5. Experimentation of nuclear weapons, releases and storages of radioactive wastes in the Kara sea and in New Zemble

    International Nuclear Information System (INIS)

    Charmasson, S.

    1996-01-01

    132 nuclear weapons were tested from 1955 to 1990 in New Zemble. From 1959 to 1993, low level liquid radioactive wastes, low and medium level solid radioactive wastes, reactor core and fuel of submarine and nuclear propelled ships were released in the Kara and the Barentz seas. For these two seas, a recapitulation of the different radioactive sources and the found level of radioactivity of the marine environment are presented. (A.B.). 22 refs. 4 figs., 6 tabs

  6. Radioactive iodine releases from nuclear power plant, (2)

    International Nuclear Information System (INIS)

    Naritomi, Mitsuo

    1974-01-01

    Internal radiation dose through the respiratory intake of fission products is predominantly due to radioactive iodine not only at the time of reactor accidents but also in normal operation of nuclear facilities. Technological studies in this field have thus been quite active to this day. With the rapid advance of nuclear power generation in recent years, the efforts to reduce environmental release of radioactive iodine and to enhance environmental safety are all the more emphasized. Experiences in the Japan Atomic Energy Research Institute during past about six years are described concerning the radioactive iodine release to the atmosphere in 131 I production and the measures taken to reduce the release. Then, problems are expounded regarding the radioactive iodine release at the time of reactor accidents and in spent fuel reprocessing. (Mori, K.)

  7. Measurement of liquid radioactive materials for monitoring radioactive emissions

    International Nuclear Information System (INIS)

    1977-10-01

    This draft regulation applies to measuring equipment for liquid radioactive materials for the monitoring of the radioactive discharges from stationary nuclear power plants with LWR and HTR reactors. Demands made on the measuring procedure, methods of concentration determination, balancing, indication of limiting values, and inspections are layed down. The draft regulation deals with: 1) Monitoring liquid radioactive discharges: Water and similar systems; radionuclides and their detection limits, radioactively contaminated water (waste water); secondary cooling water; power house cooling water; primary cooling water; flooding water; 2) Layout of the measuring and sampling equipment and demands made on continuous and discontinuous measuring equipment; demands made on discontinuous α and β measuring equipment; 3) Maintenance and repair work; inspections; repair of defects; 4) Demands made on documentation; reports to authorities; 5) Supplement: List of general and reference regulations. (orig./HP) [de

  8. Reduction of releases of radioactive effluents from light-water-power-reactors in Japan

    International Nuclear Information System (INIS)

    Yoshida, Y.; Itakura, T.; Kanai, T.

    1977-01-01

    Japan Atomic Energy Commission established the dose objectives to the population around the light-water-reactors in May, 1975, based on the ''ALAP'' concept. These values are respectively, 5 mrems per year for total body and 15 mrems per year for thyroid of an individual in the critical group in the environs, due to both gaseous and liquid effluents from LWRs in one site. The present paper describes the implications of the dose objective values, control measures which have been adopted to reduce releases of radioactive materials and related technical developments in Japan. The main control measures for reduction of radioactive gaseous effluents are an installation of a charcoal gas holdup system for decay of noble gases and a supply of clean steam for the gland seal of a turbine in BWR, and a storage tank system allowing decay of noble gases in PWR. For liquid effluents are taken measures to re-use them as the primary coolant. Consequently, the amounts of radioactivity released to the environment from any LWR during normal operation have been maintained under the level to meet the above dose objective values. For research reactors, reduction of release of effluents has also been carried out in a similar way to LWRs. In order to establish the techniques applicable for further reduction, studies are being made on the control measures to reduce leakage of radioiodine, an apparatus for removal of krypton, the treatment of laundry waste and measures to remove the crud in the primary coolant. Presentation is also made on the energy-integrated gas monitor for gaseous effluent and systems of measuring γ dose from radioactive cloud descriminating from natural background, which have been developed for effective monitoring thus reduced environmental dose

  9. Dry washing: the solution for contaminated liquid effluent releases

    International Nuclear Information System (INIS)

    L'homme, D.; Trambouze, P.

    1998-01-01

    The release of wash water used for contaminated garments poses an ever-increasing problem on nuclear sites. Even though the radioactivity is low, it mixes with organic compounds, thus polluting a large quantity of liquid effluents. In many cases, several thousands of m 3 /year per nuclear site are produced, which at times represents more than 30% of the volume of total releases. The conventional dry cleaning process is not a viable option, given that repeated washing cause clothes to fade and the odors are rot removed completely. In order to eliminate releases, STMI has developed, after several years of research with the Technological University of Compiegne, France, a solvent dry washing process for garments used in the nuclear industry. (author)

  10. Radioactivity of French coast of the Channel due to the release of technectium 99 and iodine 129: modelisation and measurements

    International Nuclear Information System (INIS)

    Robeau, D.; Patti, F.; Charmasson, S.

    1988-01-01

    Radioactive releases of Iodine 129 are controlled by measurements of the radioactivity in the liquid effluents before it is released in to the sea from the outlet of the reprocessing plant of La Hague. The effects on the marine environment are examined by a radioactive survey of Technecium 99 and Iodine 129 in Fucus (common seaweed). This radioactivity is measured along the north coast of France from Roscoff in the west of Brittany to Wimereux close to the Belgian frontier. The theoretical study of dispersion of radionuclides in the Channel has permitted a simulation model of the transfer of pollutants and particularly Technecium 99 and Iodine 129 to be formulated. (author)

  11. Impacts of accidental radioactive releases to the hydrosphere from floating and land-based nuclear power plants

    International Nuclear Information System (INIS)

    Vollmer, R.H.

    1978-01-01

    To assess the impacts of accidental radioactive releases to the hydrosphere from floating and land-based nuclear power plants, the Liquid Pathway Generic Study was initiated. The objective of this study was to compare the risks associated with releases through the liquid pathway from accidents that may reasonably be expected and those that are very improbable. Consequences were estimated in terms of radiation dose to man from drinking water ingestion, shell-fish and fish flesh consumption, and direct exposure (swimming, beach) as well as long-term effects, such as genetic effects or aquatic species degradation. (author)

  12. Liquid filter for liquids containing radioactive materials

    International Nuclear Information System (INIS)

    Rohleder, N.; Schwarz, F.

    1986-01-01

    A device for filtering radioactive liquids loaded with solids is described, which has a pressure-resistant housing with a lid and an incomer for the turbid liquid and a collecting space and drain for the filtrate at the bottom of the housing. A filter cartridge is present in this housing. Such a filtering device must be suitable for use in nuclear plants, must be easy to replace by remote control and must minimise the carrying over of radioactive particles. This problem should be solved by the filter cartridge consisting of a large number of horizontal filter plates stacked above one another, which carry a deep layer filter material acting in the sub-micron range. The turbid liquid runs into the centre of the stack of filter plates via a vertical central duct. The intermediate spaces between the filter places are connected to this central duct via the layer of filter material. The filter plates are sealed against one another on the outer circumference and have radial drain openings for the filtrate on the outside. The central duct is sealed at the lower end by a plate. When the filter cartridge is replaced, the radioactive waste in the filter cartridge remains safely enclosed and can be conditioned in suitable containers. (orig.) [de

  13. Radioactive waste management in a fuel reprocessing facility in fiscal 1982

    International Nuclear Information System (INIS)

    1984-01-01

    In the fuel reprocessing facility of the Power Reactor and Nuclear Fuel Development Corporation, radioactive gaseous and liquid waste are released not exceeding the respective permissible levels. Radioactive concentrated solutions are stored at the site. Radioactive solid waste are stored appropriately at the site. In fiscal 1982, the released quantities of radioactive gaseous and liquid waste were both below the permissible levels. The results of radioactive waste management in the fuel reprocessing facility in fiscal 1982 are given in the tables: the released quantities of radioactive gaseous and liquid waste, the produced quantities of radioactive solid waste, and the stored quantities of radioactive concentrated solutions and of radioactive solid waste as of the end of fiscal 1982. (Mori, K.)

  14. Radioactive release during nuclear accidents in Chernobyl and Fukushima

    Science.gov (United States)

    Nur Ain Sulaiman, Siti; Mohamed, Faizal; Rahim, Ahmad Nabil Ab

    2018-01-01

    Nuclear accidents that occurred in Chernobyl and Fukushima have initiated many research interests to understand the cause and mechanism of radioactive release within reactor compound and to the environment. Common types of radionuclide release are the fission products from the irradiated fuel rod itself. In case of nuclear accident, the focus of monitoring will be mostly on the release of noble gases, I-131 and Cs-137. As these are the only accidents have been rated within International Nuclear Events Scale (INES) Level 7, the radioactive release to the environment was one of the critical insights to be monitored. It was estimated that the release of radioactive material to the atmosphere due to Fukushima accident was approximately 10% of the Chernobyl accident. By referring to the previous reports using computational code systems to model the release rate, the release activity of I-131 and Cs-137 in Chernobyl was significantly higher compare to Fukushima. The simulation code also showed that Chernobyl had higher release rate of both radionuclides on the day of accident. Other factors affecting the radioactive release for Fukushima and Chernobyl accidents such as the current reactor technology and safety measures are also compared for discussion.

  15. Method of concentrating radioactive liquid waste

    International Nuclear Information System (INIS)

    Yasumura, Keijiro

    1990-01-01

    Radioactive liquid wastes generated from nuclear power facilities are caused to flow into a vessel incorporated with first hydrophobic porous membranes. Then, the radioactive liquid wastes are passed through the first hydrophobic porous membranes under an elevated or reduced pressure to remove fine particles contained in the liquid wastes. The radioactive liquid wastes passed through the first membranes are stored in a temporary store a vessel and steams generated under heating are passed through the second hydrophobic porous membranes and then cooled and concentrated as condensates. In this case, the first and the second hydrophobic porous membranes have a property of passing steams but not water and, for example, are made of tetrafluoroethylen resin type thin membranes. Accordingly, since the fine particles can be removed by the first hydrophobic porous membranes, lowering of the concentration rate due to the deposition of solid contents to the membranes upon concentration can be prevented. (I.S.)

  16. The situation of radioactive waste management in the fuel reprocessing facility (for fiscal 1979)

    International Nuclear Information System (INIS)

    1981-01-01

    In the fuel reprocessing facility of Power Reactor and Nuclear Fuel Development Corporation (PNC), the release of radioactive gaseous and liquid wastes was so controlled as not to exceed the set standards. Of the radioactive liquid wastes, concentrated wastes and sludge are stored in tanks. Radioactive solid wastes are suitably stored in containers. The situation of radioactive waste management in the fuel reprocessing facility in fiscal 1979 (from April, 1979, to March, 1980) is presented on the basis of the radiation control report made by PNC. The release of radioactive gaseous and liquid wastes was below the set standards. The following data are given in tables: the released quantity of radioactive gaseous and liquid wastes, the cumulative stored amount of radioactive liquid wastes, the produced quantity and cumulative stored amount of radioactive solid wastes; (for reference) the released quantity of radioactive gaseous and liquid wastes in fiscal 1977, 1978 and 1979. (J.P.N.)

  17. Method of processing nitrate-containing radioactive liquid wastes

    International Nuclear Information System (INIS)

    Ogawa, Norito; Nagase, Kiyoharu; Otsuka, Katsuyuki; Ouchi, Jin.

    1983-01-01

    Purpose: To efficiently concentrate nitrate-containing low level radioactive liquid wastes by electrolytically dialyzing radioactive liquid wastes to decompose the nitrate salt by using an electrolytic cell comprising three chambers having ion exchange membranes and anodes made of special materials. Method: Nitrate-containing low level radioactive liquid wastes are supplied to and electrolytically dialyzed in a central chamber of an electrolytic cell comprising three chambers having cationic exchange membranes and anionic exchange membranes made of flouro-polymer as partition membranes, whereby the nitrate is decomposed to form nitric acid in the anode chamber and alkali hydroxide compound or ammonium hydroxide in the cathode chamber, as well as concentrate the radioactive substance in the central chamber. Coated metals of at least one type of platinum metal is used as the anode for the electrolytic cell. This enables efficient industrial concentration of nitrate-containing low level radioactive liquid wastes. (Yoshihara, H.)

  18. Generation projection of solid and liquid radioactive wastes and spent radioactive sources in Mexico

    International Nuclear Information System (INIS)

    Garcia A, E.; Hernandez F, I. Y.; Fernandez R, E.; Monroy G, F.; Lizcano C, D.

    2014-10-01

    This work is focused to project the volumes of radioactive aqueous liquid wastes and spent radioactive sources that will be generated in our country in next 15 years, solids compaction and radioactive organic liquids in 10 years starting from the 2014; with the purpose of knowing the technological needs that will be required for their administration. The methodology involves six aspects to develop: the definition of general objectives, to specify the temporary horizon of projection, data collection, selection of the prospecting model and the model application. This approach was applied to the inventory of aqueous liquid wastes, as well as radioactive compaction organic and solids generated in Mexico by non energy applications from the 2001 to 2014, and of the year 1997 at 2014 for spent sources. The applied projection models were: Double exponential smoothing associating the tendency, Simple Smoothing and Lineal Regression. For this study was elected the first forecast model and its application suggests that: the volume of the compaction solid wastes, aqueous liquids and spent radioactive sources will increase respectively in 152%, 49.8% and 55.7%, while the radioactive organic liquid wastes will diminish in 13.15%. (Author)

  19. Method of processing radioactive nuclide-containing liquids

    International Nuclear Information System (INIS)

    Hirai, Masahide; Tomoshige, Shozo; Kondo, Kozo; Suzuki, Kazunori; Todo, Fukuzo; Yamanaka, Akihiro.

    1985-01-01

    Purpose: To solidify radioactive nuclides in to a much compact state and facilitate the storage. Method: Liquid wastes such as drain liquids generated from a nuclear power plant at a low density of 1 x 10 -6 - 10 -4 μCi/ml are previously brought into contact with a chelate type ion exchange resin such as of phenolic resin or ion exchange resin to adsorb the radioactive nuclides on the resin and the nuclides are eluted with sulfuric acid or the like to obtain liquid concentrates. The liquid concentrates are electrolyzed in an ordinary electrolytic facility using platinum or the like as the anode, Al or the like as the cathode, under the presence of 1 - 20 g/l of non-radioactive heavy metals such as Co and Ni in the liquid and while adjusting pH to 2 - 8. The electrolysis liquid residue is returned again to the electrolysis tank as it is or in the form of precipitates coagulated with a polymeric floculant. The supernatant liquid upon floculating treatment is processed with the chelate type ion exchange resin into hazardless liquid. (Sekiya, K.)

  20. Solid and liquid radioactive wastes

    International Nuclear Information System (INIS)

    Cluchet, J.; Desroches, J.

    1977-01-01

    The problems raised by the solid and liquid radioactive wastes from the CEA nuclear centres are briefly exposed. The processing methods developed at the Saclay centre are described together with the methods for the wastes from nuclear power plants and reprocessing plants. The different storage techniques used at the La Hague centre are presented. The production of radioactive wastes by laboratories, hospitals and private industry is studied for the sealed sources and the various radioactive substances used in these plants. The cost of the radioactive wastes is analysed: processing, transport, long term storage [fr

  1. Radioactive liquid waste processing system

    International Nuclear Information System (INIS)

    Inakuma, Masahiko; Takahara, Nobuaki; Hara, Satomi.

    1996-01-01

    Laundry liquid wastes and shower drains containing radioactive materials generated in a nuclear power plant are removed with radioactive materials by a fiber filtration device and an activated carbon filtration device to satisfy standers of water quality described in the environmental effect investigation report. Spent activated carbon is dehydrated together with the back-wash liquid from the fiber filtration device and the activated carbon filtration device using a Nutsche-type filtration dryer. With such procedures, the scale of the facility is minimized, space for devices, maintenance for equipments and radiation dose rate are reduced. (T.M.)

  2. Radioactive Liquid Waste Treatment Facility: Environmental Information Document

    Energy Technology Data Exchange (ETDEWEB)

    Haagenstad, H.T.; Gonzales, G.; Suazo, I.L. [Los Alamos National Lab., NM (United States)

    1993-11-01

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

  3. Radioactive Liquid Waste Treatment Facility: Environmental Information Document

    International Nuclear Information System (INIS)

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

    1993-11-01

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

  4. Radioactive wastes management in fiscal year 1983 in the fuel reprocessing plant

    International Nuclear Information System (INIS)

    1985-01-01

    In the nuclear fuel reprocessing plant of Power Reactor and Nuclear Fuel Development Corporation, the releases of radioactive gaseous and liquid wastes are so managed not to exceed the respective objective release levels. Of the radioactive liquid wastes, the high level concentrated wastes are stored in tanks and the low level wastes are stored in tanks or asphalt solidified. For radioactive solid wastes, high level solid wastes are stored in casks, low level solid wastes and asphalt solids in drums etc. The releases of radioactive gaseous and liquid wastes in the fiscal year 1983 were below the objective release levels. The radioactive wastes management in the fuel reprocessing plant in fiscal year 1983 is given in tables, the released quantities, the stored quantities, etc. (Mori, K.)

  5. Method and apparatus for glass solidification porcessing for radioactive liquid waste

    International Nuclear Information System (INIS)

    Torada, Shin-ichiro; Masaki, Toshio; Sakai, Akira.

    1989-01-01

    Glass material supplied to a glass melting furnace is made in the form of a glass container. Then, radioactive liquid wastes are directly injected into the glass vessel and the glass vessel injected with the radioactive liquid wastes is charged into the glass melting furnace. The glass material and the radioactive liquid wastes are supplied simultaneously to the glass melting furnace. Then, corresponding to the amount of the glass material used for the glass vessel, the amount of the radioactive liquid wastes injected to the inside thereof is controlled to thereby set the mixing ratio between the glass material and the radioactive liquid wastes. Further, by controlling the number of the glass vessels injected with the radioactive liquid wastes to be charged into the glass melting furnace, the amount of supplying the radioactive liquid wastes and the glass material is controlled. This can easily maintain constant the amount of the glass material and the radioacative liquid wastes supplied to the glass melting furnace and the mixing ratio thereof. (T.M.)

  6. Historical releases of radioactivity to the environment from ORNL

    International Nuclear Information System (INIS)

    Ohnesorge, W.F.

    1986-05-01

    This report gives a brief history and assessment of ORNL radionuclide releases to the environment. A short history and an inventory of radioactivity disposed of by shallow land burial and hydrofracture techniques are given along with a brief discussion of the potential environmental impact of these disposed materials. The data available for this report varied greatly in quality. Data on contaminated liquid waste and liquids discharged to the environment are much more reliable than early data on contaminated air discharges and contaminated solid waste. Data for more recent years are more complete and reliable than data obtained from records dating back to the early history of the laboratory. The data presented here do not include materials retrievably stored (with the exception of uranium solid waste data) or inventories of materials stored in operating or surplus facilities since these materials presently have no contact with or loss to the environment

  7. Methodological studies for deriving release criteria for liquid effluents from medical installations

    International Nuclear Information System (INIS)

    Shu, Jane; Comissao Nacional de Energia Nuclear; Rochedo, Elaine R.R.; Heilbron Filho, Paulo F.L.; Crispim, Verginia R.

    2009-01-01

    This work aims to develop a methodology for the assessment of clearance limits for the release of liquid waste arising from medical installations using radionuclides for medical diagnostic purposes in the town of Rio de Janeiro. The results will be used to assess the need to justify or to revise the current clearance values as specified in regulation CNEN-NE-6.05 - Radioactive Waste Management in Radioactive Facilities. The proposed methodology is based on the mathematical model recommended by the International Atomic Energy Agency, adapted to the observed release conditions in the study area. In order to turn the assessment as realistic as possible, two scenarios are simulated. The first scenario simulates the release to the sewage system with access to a sewer treatment stations. The second scenario simulates the releases without passing a treatment station, with direct outflow to surface water. Probabilistic assessments were performed using the Crystal Ball software. Distributions were than compared to current IAEA clearance criteria that include specific values for average and low probability scenarios. The results will be used to derive adequate clearance levels for radionuclides used in nuclear diagnostic medicine in Brazil according to the specific relevant release scenario. (author)

  8. Method for treating radioactive liquids

    International Nuclear Information System (INIS)

    Komrow, R.R.; Pritchard, J.F.

    1980-01-01

    A process for treating and handling radioactive liquids and rendering such liquids safe for handling is disclosed. Transportation and disposal, the process comprises adding thereto a small amount of a water-insoluble alkali salt of an aqueous alkali saponified gelatinized-starch-polyacrylonitrile graft polymer, to form a solid, semi-solid or gel product

  9. Radioactive releases into the environment under accidental conditions

    International Nuclear Information System (INIS)

    Beninson, D.

    1976-01-01

    Although accidents involving the release of radioactive materials and the unplanned exposure of people can occur at any stage of the nuclear fuel cycle, most attention has been focused on reactor accidents. Although no power reactor accidents involving exposure of the public have yet occured, it should be recognized that the probability of such accidental releases cannot be reduced to zero. Since the inventory of radioactive materials in power reactors is very large, it is usual to postulate, for safety assessments, that a release of fission products takes place in spite of all protective measures. This postulated release is of importance for reactor siting and for preparing emergency plans. (HP) [de

  10. Radioactive liquid wastes processing device

    International Nuclear Information System (INIS)

    Sauda, Kenzo; Koshiba, Yukihiko; Yagi, Takuro; Yamazaki, Hideki.

    1985-01-01

    Purpose: To carry out optimum photooxidizing procession following after the fluctuation in the density of organic materials in radioactive liquid wastes to thereby realize automatic remote procession. Constitution: A reaction tank is equipped with an ultraviolet lamp and an ozone dispersing means for the oxidizing treatment of organic materials in liquid wastes under the irradiation of UV rays. There are also provided organic material density measuring devices to the inlet and outlet of the reaction tank, and a control device for controlling the UV lamp power adjusting depending on the measured density. The output of the UV lamp is most conveniently adjusted by changing the applied voltage. The liquid wastes in which the radioactivity dose is reduced to a predetermined level are returned to the reaction tank by the operation of a switching valve for reprocession. The amount of the liquid wastes at the inlet is controlled depending on the measured ozone density by the adjusting valve. In this way, the amount of organic materials to be subjected to photolysis can be kept within a certain limit. (Kamimura, M.)

  11. 10 CFR 61.41 - Protection of the general population from releases of radioactivity.

    Science.gov (United States)

    2010-01-01

    ... radioactivity. 61.41 Section 61.41 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR... from releases of radioactivity. Concentrations of radioactive material which may be released to the... maintain releases of radioactivity in effluents to the general environment as low as is reasonably...

  12. Discussion on the methods for calculation release limits for low-level radioactive waste

    International Nuclear Information System (INIS)

    Cao Fengbo; Liu Xiaochao

    2012-01-01

    The release request for low-level radioactive waste are briefly described in this paper. Associating with the conditions of low-level radioactive waste of some radioactive waste processing station, the methods and gist for calculating release limits for low-level radioactive waste with national release limits and annual effective dose limit for the public or the occupation are discussed. Then release limits for the low-level radioactive waste are also proposed. (authors)

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

    International Nuclear Information System (INIS)

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

    1998-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

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

  15. Developing technologies for conditioning the liquid organic radioactive wastes from Cernavoda NPP

    International Nuclear Information System (INIS)

    Deneanu, N.; Popescu, I. V.; Teoreanu, I.

    2004-01-01

    The Institute for Nuclear Research (INR)-Pitesti has developed technologies for conditioning liquid organic radioactive wastes (oils, miscellaneous solvent and liquid scintillation cocktail) for Cernavoda NPP. This paper describes the new and viable solidification technology to convert liquid organic radioactive wastes into a stable monolithic form, which minimizes the probability to release tritium in the environment during interim storage, transportation and final disposal. These are normally LLW containing only relatively small quantities of beta/gamma emitting radionuclides and variable amounts of tritium with activity below E+08Bq/l. The INR research staff in the radwaste area developed treatment/conditioning techniques and also designed and tested the containers for the final disposal, following the approach in the management of radwaste related to the nuclear fuel cycle. Thus, the INR focused this type of activity on treating and conditioning the wastes generated at Cernavoda Nuclear Power Plant consisting of lubricants from primary fuelling machines and turbine, the miscellaneous solvent from decontamination operation and the liquid scintillation cocktail used in radiochemical analysis. Laboratory studies on cementation of liquid organic radioactive wastes have been undertaken at INR Pitesti. One simple system, similar to a conventional cement solidification unit, can treat radioactive liquid wastes, which are the major components of low- and medium-level radioactive wastes generated by a Nuclear Power Plant. It was proved that the solidified waste could meet the Waste Acceptance Criteria of the disposal site, in this case Baita-Bihor National Repository, as follows: - The wastes are deposited in type A packages; - The maximum expected quantities of this waste stream that will be produced in the future are 50 drums per year. The maximum specific tritium activity per drum is 10 9 Bq/m 3 ; - Compressive strengths of the samples should be greater than 50 MPa

  16. Device for the disposal of radioactive liquid wastes

    International Nuclear Information System (INIS)

    Tomizawa, Toshi; Inoue, Tadashi.

    1976-01-01

    Object: To adsorb and collect radioactive nuclide ions contained in the radioactive liquid waste to select and separate thereof. Structure: A unitary disposing tank comprises an insulative cylindrical tank, an unsoluble cathode plate positioned thereunder and formed with a number of liquid inlet holes, an adsorbent layer filled with unsoluble electrically conductive substances having a large surface area in contact with the cathode plate, and an unsoluble anode plate positioned at the upper part of the cylindrical disposing tank so as not to come into contact with the adsorbent layer and formed with a number of liquid inlets, whereby one or more disposing tanks are stacked in a layer fashion, and a DC voltage is applied between the anode and cathode plates to flow a liquid to be disposed into the disposing tanks so that the radioactive metal ion nuclide in the liquid may be adsorbed and collected by the cathode and the adsorbent layer for selection and separation. (Ohara, T.)

  17. Released radioactivity reducing facility

    International Nuclear Information System (INIS)

    Tanaka, Takeaki.

    1992-01-01

    Upon occurrence of a reactor accident, penetration portions of a reactor container, as a main leakage source from a reactor container, are surrounded by a plurality of gas-tight chambers, the outside of which is surrounded by highly gas-tightly buildings. Branched pipelines of an emergency gas processing system are introduced to each of the gas-tight chambers and they are joined and in communication with an emergency gas processing device. With such a constitution, radioactive materials are prevented from leaking directly from the buildings. Further, pipeline openings of the emergency gas processing facility are disposed in the plurality highly gas-tight penetration chambers. If the radioactive materials are leaked from the reactor to elevate the pressure in the penetration chambers, the radioactive materials are introduced to a filter device in the emergency gas processing facility by way of the branched pipelines, filtered and then released to the atmosphere. Accordingly, the reliability and safety of the system can be improved. (T.M.)

  18. Impact on the marine environment of radioactive releases resulting from the Fukushima-Daiichi nuclear accident. April 4, 2011

    International Nuclear Information System (INIS)

    2011-01-01

    After having outlined that measurements taken over several days in the sea water in the vicinity of the power station have revealed severe contamination of the marine environment by various radionuclides released as a result of the accident at the Fukushima-Daiichi nuclear power station, this report describes the origins of the contamination of the marine environment (release of liquid effluents directly into the sea in the vicinity of the damaged reactors, atmospheric fallout onto the surface of the sea, conveyance of radioactive pollution by rainout of contaminated ground). Then, while proposing several maps, the authors analyse and comment the dispersion in the sea of radioactive pollutants by addressing the following issues: topography of the sea bed and sea currents off the Japanese coast, immediate or short term dispersion (a few days), mid-term dispersion (weeks and months), long term and large scale future of the radioactive pollutants. Finally, the report briefly discusses the impact of radioactive pollution on living species

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

    Energy Technology Data Exchange (ETDEWEB)

    Osmanlioglu, Ahmet Erdal [Radioactive Waste Management Unit (RWMU), Turkish Atomic Energy Authority, Cekmece Nuclear Research and Training Center, Altinsehir Yolu 5 km. Halkali, 34303K Cekmece, Istanbul (Turkey)]. E-mail: Erdal.Osmanlioglu@taek.gov.tr

    2007-01-15

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

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

    International Nuclear Information System (INIS)

    Osmanlioglu, Ahmet Erdal

    2007-01-01

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

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

    Science.gov (United States)

    Osmanlioglu, Ahmet Erdal

    2007-01-01

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

  2. Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

    International Nuclear Information System (INIS)

    1996-01-01

    The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour

  3. Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour.

  4. Order of the 10 january 2003 authorizing the national agency for the radioactive wastes management to follow the gaseous and liquid effluents release for the exploitation of the radioactive wastes storage center of the Manche; Arrete du 10 janvier 2003 autorisant l'Agence nationale pour la gestion des dechets radioactifs a poursuivre les rejets d'effluents gazeux et liquides pour l'exploitation du centre de stockage de dechets radioactifs de la Manche

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-01-01

    This document, took out from the Official Journal, is the law text relative to the order of the 10 january 2003 authorizing the national agency for the radioactive wastes management to follow the gaseous and liquid effluents release for the exploitation of the radioactive wastes storage center of the Manche. (A.L.B.)

  5. Final treatment of liquid radioactive wastes

    International Nuclear Information System (INIS)

    Svolik, S.

    2004-01-01

    Final treatment of liquid radioactive wastes which are produced by 1 st and 2 nd bloc of the Mochovce NPP, prepares the NPP in its natural range. The purpose of the equipment is liquidation of wastes, which are formed at production. Wastes are warehoused in the building of active auxiliary plants in the present time, where are reservoirs in which they are deposited. Because they are already feeling and in 2006 year they should be filled definitely, it is necessary to treat them in that manner, so as they may be liquidated. Therefore the Board of directors of the Slovenske elektrarne has disposed about construction of final treatment of liquid radioactive wastes in the Mochovce NPP. Because of transport the wastes have to be treated in the locality of power plant. Technically, the final treatment of the wastes will be interconnected with building of active operation by bridges. These bridges will transport the wastes for treatment into processing centre

  6. Boiling water reactor liquid radioactive waste processing system

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The standard sets forth minimum design, construction and performance requirements with due consideration for operation of the liquid radioactive waste processing system for boiling water reactor plants for routine operation including design basis fuel leakage and design basis occurrences. For the purpose of this standard, the liquid radioactive waste processing system begins at the interfaces with the reactor coolant pressure boundary, at the interface valve(s) in lines from other systems and at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material. The system terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system and at the point of recycle back to storage for reuse. The standard does not include the reactor coolant clean-up system, fuel pool clean-up system, sanitary waste system, any nonaqueous liquid system or controlled area storm drains

  7. Devoluming method of acidic radioactive liquid waste and processing system therefor

    International Nuclear Information System (INIS)

    Shirai, Takamori; Honda, Tadahiro

    1998-01-01

    Radioactive liquid wastes such as liquid wastes discharged from chemical decontamination (containing free acids, metal salts dissolved in acids, not-dissolved iron rust and radioactive metals) are introduced to an acid recovering device using a diffusion permeation membrane and separated to a deacidified liquid and separated acid liquid. The separated acid liquid mainly comprising free acids is recovered to a tank for recovered acids, and used repeatedly for removing crud. The deacidified liquid mainly comprising salts is concentrated in a reverse osmosis membrane (RO) concentration device. RO concentrated liquid containing radioactive metals is dried, and salts are decomposed in a drying/salt-decomposing device and separated into metal oxides and a mixed gas of an acidic gas and steams. The gas is cooled in an acid absorbing device and recovered as free acids. The metal oxides containing radioactive metals are solidified. (I.N.)

  8. Radioactive waste control at the reprocessing facility in fiscal 1980

    International Nuclear Information System (INIS)

    1982-01-01

    At the fuel reprocessing facility of the Power Reactor and Nuclear Fuel Development Corporation (PNC), the release of radioactive gaseous and liquid wastes are controlled so as not to exceed the specific levels. Concentrated low and high level liquid wastes, sludge, etc. are contained in storage tanks. Low and high level solid wastes are stored in appropriate containers. In fiscal 1980 (April to March), the release of gaseous and liquid wastes was below the specific levels (as in the previous years). Based on the report made by PNC in accordance with the law concerning the regulation of reactors, etc., the following data are presented in tables: the released quantity of radioactive gaseous and liquid wastes in fiscal 1980, the cumulative stored quantity of radioactive liquid wastes up to fiscal 1980; the cumulative stored quantity of radioactive solid wastes up to fiscal 1980 and the quantity of the same stored in fiscal 1980. (J.P.N.)

  9. The liquidation of liquid radioactive waste on nuclear medicine departments

    International Nuclear Information System (INIS)

    Fueriova, A.

    1995-01-01

    The most serious problems for Clinic of Nuclear Medicine of National Oncological Institute, Bratislava (CNM) is the localization of CNM in the downtown, inside the hospital area with the dilution water deficit. This department is the only one in Slovak Republic performing therapeutical applications. To be able to perform the necessary amount of therapies and also to introduce a new therapeutical methods, in 1992-1994 the old liquidation waste disposal station (LWDS) was reconstructed with the aim to satisfy the newest requirements of radiation hygiene. LWDS is the 5-floor object partly underground which satisfied the requirements for liquidation of radioactive liquid waste from diagnostic procedures(annually 5000 patients) and also from 200 therapeutical applications annually (15 beds, 720 GBq iodine-131). The capacity of LWDS is able to store about 90 m 3 liquid radioactive waste. Part of the underground spaces are used for the storage of solid radioactive trash. The liquid waste from CNM is collected through isolated metal sewage system to the storage with continuous observation of water specific activity. According to the activity, the liquid waste is placed to the 5 decay storages with the volume about 15 m 3 . The six one serves for the case of technical accident. When the activity declines, the liquid waste is diluted with non active medical trash to the level which is acceptable by low about radiation hygiene protection. The storage walls are made from barium-concrete 25-50 cm thick which is enough for sufficient protection of operation staff and also for walking around persons. Double-layer high quality chemical material prevents the water leak and diffusion of radionuclides into the concrete. Technology consists of cast-iron drains, powerful slush pumps, operation valves, regulation technology from dosimetric system for continuous monitoring of specific activity, for managing system with powerful industrial computer

  10. The liquidation of liquid radioactive waste on nuclear medicine departments

    Energy Technology Data Exchange (ETDEWEB)

    Fueriova, A [National Oncological Institue, Bratislava (Slovakia). Hospital St. Elis, Clinic of Nuclear Medicine

    1996-12-31

    The most serious problems for Clinic of Nuclear Medicine of National Oncological Institute, Bratislava (CNM) is the localization of CNM in the downtown, inside the hospital area with the dilution water deficit. This department is the only one in Slovak Republic performing therapeutical applications. To be able to perform the necessary amount of therapies and also to introduce a new therapeutical methods, in 1992-1994 the old liquidation waste disposal station (LWDS) was reconstructed with the aim to satisfy the newest requirements of radiation hygiene. LWDS is the 5-floor object partly underground which satisfied the requirements for liquidation of radioactive liquid waste from diagnostic procedures(annually 5000 patients) and also from 200 therapeutical applications annually (15 beds, 720 GBq iodine-131). The capacity of LWDS is able to store about 90 m{sup 3} liquid radioactive waste. Part of the underground spaces are used for the storage of solid radioactive trash. The liquid waste from CNM is collected through isolated metal sewage system to the storage with continuous observation of water specific activity. According to the activity, the liquid waste is placed to the 5 decay storages with the volume about 15 m{sup 3}. The six one serves for the case of technical accident. When the activity declines, the liquid waste is diluted with non active medical trash to the level which is acceptable by low about radiation hygiene protection. The storage walls are made from barium-concrete 25-50 cm thick which is enough for sufficient protection of operation staff and also for walking around persons. Double-layer high quality chemical material prevents the water leak and diffusion of radionuclides into the concrete. Technology consists of cast-iron drains, powerful slush pumps, operation valves, regulation technology from dosimetric system for continuous monitoring of specific activity, for managing system with powerful industrial computer.

  11. Potential of membrane processes in management of radioactive liquid waste

    International Nuclear Information System (INIS)

    Kumar, Surender; Jain, Savita; Raj, Kanwar

    2010-01-01

    Various categories of radioactive liquid waste are generated during operations and maintenance of nuclear installations. The potential of membrane processes for the treatment of low-level radioactive liquids is discussed in this paper

  12. Removal of Radioactive Pollutants by Liquid Emulsion Membrane From Liquid Waste

    International Nuclear Information System (INIS)

    Yossef, Y.A.A.

    2013-01-01

    Radioactive liquid waste should be safely managed because it is potentially hazardous to human health and the environment. Several methods were used for treatment of liquid waste, such as liquid emulsion membrane (LEM). In this work, liquid emulsion membrane using Tri-butyl phosphate (TBP) plus Bis (2-ethylhexyl) phosphate (HDEHP) as mobile carriers, hydrochloric acid (HCl) as stripping agents and an emulsifying agent (span 80) was used for the extraction of uranium ions from radioactive liquid waste. Various parameters influencing the permeation of uranium ions through the membrane have been optimized to separate uranium ions from radioactive liquid waste such as: the effects of membrane material, carrier concentration, operating conditions, etc. were examined; moreover, the transport mechanism of this uranium was also studied. The internal mass transfer in the water/oil (W/O) emulsion drop, the external mass transfer around the drop, the rates of formation, and the decomposition of the complex at the external aqueous-organic interface were considered. The results show that, the liquid emulsion membrane which consists of (25% by volume HDEHP, 0.005 M + 75% by volume TBP, 0.01 M) as extractant (carrier), span 80, 4% (v/v) (sorbitan monooleate) as surfactant agent, hydrochloric acid (HCl), (1.0 M) as stripping agent. From the results, the maximum extraction percent of uranium ions (nearly about of 100%) occurred at the operating conditions: stirring speed =500 rpm, the ratio between LEM and feed phase (liquid waste) = 20 ml: 100 ml, the ratio between organic phase (membrane phase) to internal aqueous phase (stripping phase) = 1.0 and the ph value of the external aqueous phase equal to 5.0.

  13. Spray drying of liquid radioactive wastes

    International Nuclear Information System (INIS)

    Abrams, R.F.; Monat, J.P.

    1984-01-01

    Full scale performance tests of a Koch spray dryer were conducted on simulated liquid radioactive waste streams. The liquid feeds simulated the solutions that result from radwaste incineration of DAW an ion exchange resins, as well as evaporator bottoms. The integration of the spray dryer into a complete system is discussed

  14. Process for treatment of detergent-containing radioactive liquid wastes

    International Nuclear Information System (INIS)

    Kamiya, K.; Chino, K.; Funabashi, K.; Horiuchi, S.; Motojima, K.

    1984-01-01

    A detergent-containing radioactive liquid waste originating from atomic power plants is concentrated to have about 10 wt. % detergent concentration, then dried in a thin film evaporator, and converted into powder. Powdered activated carbon is added to the radioactive waste in advance to prevent the liquid waste from foaming in the evaporator by the action of surface active agents contained in the detergent. The activated carbon is added in accordance with the COD concentration of the radioactive liquid waste to be treated, and usually at a concentration 2-4 times as large as the COD concentration of the liquid waste to be treated. A powdery product having a moisture content of not more than 15 wt. % is obtained from the evaporator, and pelletized and then packed into drums to be stored for a predetermined period

  15. Deep-well injection of liquid radioactive waste in Russia. Present situation

    International Nuclear Information System (INIS)

    Rybalchenko, A.

    1998-01-01

    At present there are 3 facilities (polygons) for the deep-well injection of liquid radioactive waste in Russia, all of which were constructed in the mid60's. These facilities are operating successfully, and activities have started in preparation for decommissioning. Liquid radioactive waste is injected into deep porous horizons which act as 'collector-layers', isolated from the surface and from groundwaters by a relatively thick sequence of rock of low permeability. The collector-layers (also collector-horizons) contain salt waters or fresh waters of no practical application, lying beneath the main horizons containing potable waters. Construction of facilities for the deep-well injection of liquid radioactive waste was preceded by geological surveys and investigations which were able to substantiate the feasibility and safety of radioactive waste injection, and to obtain initial data for facility design. Operation of the facilities was accompanied by monitoring which confirmed that the main safety requirement was satisfied i.e. localisation of radioactive waste within specified boundaries of the geologic medium. The opinion of most specialists in the atomic power industry in Russia favours deep-well injection as a solution to the problem of liquid radioactive waste management; during the period of active operation of defence facilities (atomic power industry of the former U.S.S.R.), this disposal method prevented the impact of radioactive waste on man and the environment. The experience accumulated concerning the injection of liquid radioactive waste in Russia is of interest to scientists and engineers engaged in problems of protection and remediation of the environment in the vicinity of nuclear industry facilities; an example of the utilisation of the deep subsurface for solidified radioactive waste and the disposal of different types of nuclear materials. Information on the scientific principles and background for the development of facilities for the injection

  16. Pretreatment method for radioactive iodine-containing liquid wastes and pretreatment device

    International Nuclear Information System (INIS)

    Wakaida, Yasuo.

    1996-01-01

    Heretofore, radioactive iodine-containing liquid wastes have been discharged directly to a storing and decaying storage vessel to conduct a water draining treatment. In the present invention, the radioactive iodine-containing liquid wastes to be discharged are not discharged to the storage vessel directly but injected to a filling tank, as a pretreatment, to distinguish whether proteins are mixed in the liquid wastes or not. When proteins are mixed, miscellaneous materials such as proteins are recovered and removed by a protein processing system. When proteins are not mixed, radioactive iodine is recovered and removed directly by an iodine processing system. With such procedures, water draining treatment in the storing and decaying storage vessel is mitigated, and even when the amount of the radioactive iodine-containing liquid wastes is increased, the existent maintaining and decaying storage vessel can be used as it is. Accordingly, a safe water draining treatment with good efficiency can be conducted relative to radioactive iodine-containing liquid wastes at a reduced cost. (T.M.)

  17. Proposal for radioactive liquid waste management in a brachytherapy sealed sources development laboratory

    International Nuclear Information System (INIS)

    Souza, C.D.; Peleias Jr, F.S.; Rostelato, M.E.C.M.; Zeituni, C.A.; Benega, M.A.G.; Tiezzi, R.; Mattos, F.R.; Rodrigues, B.T.; Oliveira, T.B.; Feher, A.; Moura, J.A.; Costa, O.L.

    2014-01-01

    The radioactive waste management is addressed in several regulations. Literature survey indicates limited guidance on liquid waste management in Brachytherapy I-125 seeds production. Laboratories for those seeds are under implementation not only in Brazil but in several countries such as Poland, South Korea, Iran, China, and others. This paper may be used as reference to these other groups. For the correct implementation, a plan for radiological protection that has the management of radioactive waste fully specified is necessary. The proposal is that the waste will be deposited in a 20 L and 60 L containers which will take 2 years to fill. For glove box 1, the final activity of this container is 1.91 x 10 10 Bq (3.19 years to safe release in the environment). For glove box 3, the final activity of this container is 1.28 x 10 10 Bq (2.85 years to safe release in the environment). (authors)

  18. Derived release limits (DRL's) for airborne and liquid effluents from the Chalk River Nuclear Laboratories during normal operations

    International Nuclear Information System (INIS)

    Palmer, J.F.

    1981-02-01

    Derived release limits (DRL's), based on regulatory dose limits, have been calculated for routine discharges of radioactivity in airborne and liquid effluents from the Chalk River Nuclear Laboratories. Three types of sources of airborne effluents were considered: the NRX/NRU stack, the 61 m stack connected to the 99 Mo production facility, and a roof vent typical of those installed on several buildings on the site. Sources of liquid effluents to the Ottawa River were treated as a single source from the site as a whole. Various exposure pathways to workers on the site and to members of the public outside the site boundary were considered in the calculations. The DRL's represent upper limits for routine emissions of radioactivity from the Chalk River Nuclear Laboratories to the surrounding environment. Actual releases are regulated by Administrative Levels, set lower than the DRL's, and are confirmed by monitoring. (author)

  19. Evaluation Of Radioactivity Of Released Air During Postirradiation Examination At The RMI

    International Nuclear Information System (INIS)

    Prayitno, B.; Eko, Pudjadi

    1998-01-01

    Radioactivity evaluation of released air during post-irradiation examination at RMI has been done since january 1993 to december 1996. The released air radioactivity has been observed during post-irradiation examinations of irradiated fuels and structure materials in the hot cell. Analysis method employed has been by air sampling and measurement of the alpha and beta activities by using alpha beta aerosol LB 150 D model. Air release from RMI was sucked by using an air pump having 40 m3/hour capacity which is equipped with a 200 mm diameter filter paper. The filter paper is automatically counted by the detector of the instrument. The average of the daily maximum counting result in a month has been used as monthly data. It has been shown that there have been increase in the released air radioactivity caused by the post-irradiation examination activity. The data of the released air activity obtained have been used to calculate the radioactivity concentration and radioactivity on the soil surface based on Gauss Plumes. The calculation result have shown that the alpha and beta radioactivity concentration at stacks and on the soil surface are less then the set maximum permissible concentration (MPC)

  20. Treatment and conditioning of radioactive organic liquids

    International Nuclear Information System (INIS)

    1992-07-01

    Liquid organic radioactive wastes are generated from the use of radioisotopes in nuclear research centres and in medical and industrial applications. The volume of these wastes is small by comparison with aqueous radioactive wastes, for example; nevertheless, a strategy for the effective management of these wastes is necessary in order to ensure their safe handling, processing, storage and disposal. A aqueous radioactive wastes may be discharged to the environment after the radioactivity has decayed or been removed. By contrast, organic radioactive wastes require management steps that not only take account of their radioactivity, but also of their chemical content. This is because both the radioactivity and the organic chemical nature can have detrimental effects on health and the environment. Liquid radioactive wastes from these applications typically include vacuum pump oil, lubricating oil and hydraulic fluids, scintillation cocktails from analytical laboratories, solvents from solvent extraction research and uranium refining, and miscellaneous organic solvents. The report describes the factors which should be considered in the development of appropriate strategies for managing this class of wastes from generation to final disposal. Waste sources and characterization, treatment and conditioning processes, packaging, interim storage and the required quality assurance are all discussed. The report is intended to provide guidance to developing Member States which do not have nuclear power generation. A range of processes and procedures is presented, though emphasis is given to simple, easy-to-operate processes requiring less sophisticated and relatively inexpensive equipment. 31 refs, 16 figs, 3 tabs

  1. Processing method for liquid waste containing various kinds of radioactive material

    International Nuclear Information System (INIS)

    Toyabe, Keiji; Nabeshima, Masahiro; Ozeki, Noboru; Muraki, Tsutomu.

    1996-01-01

    Various kind of radioactive materials and heavy metal elements dissolved in liquid wastes are removed from the liquid wastes by adsorbing them on chitin or chitosan. In this case, a hydrogen ion concentration in the liquid wastes is adjusted to a pH value of from 1 to 3 depending on the kinds of the radioactive materials and heavy metal elements to be removed. Since chitin or chitosan has a special ion exchange performance or adsorbing performance, chemical species comprising radioactive materials or heavy metals dissolved in the liquid wastes are adsorbed thereto by ion adsorption or physical adsorption. With such procedures, radioactive materials and heavy metal elements are removed from the liquid wastes, and the concentration thereof can be reduced to such a level that they can be discharged into environments. On the other hand, since chitin or chitosan adsorbing the radioactive materials and heavy metal elements has a structure of polysaccharides, it is easily burnt into gaseous carbon dioxide. Accordingly, the amount of secondary wastes can remarkably be reduced. (T.M.)

  2. Methodology development for radioactive waste treatment of CDTN/BR - liquid low-level radioactive wastes

    International Nuclear Information System (INIS)

    Morais, Carlos Antonio de

    1996-01-01

    The radioactive liquid wastes generated in Nuclear Technology Development Centre (CDTN) were initially treated by precipitation/filtration and then the resulting wet solid wastes were incorporated in cement. These wastes were composed of different chemicals and different radioactivities and were generated by different sectors. The objective of the waste treatment method was to obtain minimum wet solid waste volume and decontamination and minimum operational cost. The composition of the solid wastes were taken into consideration for compatible cementation process. Approximately 5,400 litres of liquid radioactive wastes were treated by this process during 1992-1995. The volume reduction was 1/24 th and contained 20% solids. (author)

  3. Principles for limiting releases of radioactive effluents into the environment

    International Nuclear Information System (INIS)

    1986-01-01

    This publication is concerned with the subject of limiting releases of radioactive effluents during normal, controlled operation of nuclear installations. It does not deal with releases from accidents where it is only possible to limit exposures by intervention. In 1978 the IAEA published guidance on the concepts and principles for use by the competent authorities in setting limits for planned releases of radioactive material into the environment (Safety Series No. 45). This publication is a complete revision of Safety Series No. 45 and its Annex

  4. Radioactive liquid water processing method

    International Nuclear Information System (INIS)

    Yasumura, Keijiro; Noda, Tetsuya; Kobayashi, Fumio.

    1993-01-01

    Alkaline earth metals and heavy metals are added to radioactive liquid wastes containing a surface active agent comprising alkali metal salts of higher fatty acids. These metals form metal soaps with the surface active agent dissolved in the liquid wastes and crystallized. The crystallized metal soaps are introduced to a filtering column filled with a burnable polymeric fibrous filtering material. The filtering material is burnt. This can remove the surface active agent to remove COD without using an active carbon. (T.M.)

  5. Monitoring of the release of gaseous and aerosol-bound radioactive materials. Pt. 2

    International Nuclear Information System (INIS)

    1992-01-01

    KTA 1503 contains requirements on technical installations and supplementary organizational measures considered necessary in order to monitor the release of gaseous and aerosol-bound radioactive materials. It consists of part 1: Monitoring of the release of radioactive materials together with stack gas during normal operation; part 2: Monitoring of the release of radioactive materials together with stack gas in the event of incidents; part 3: Monitoring of radioactive materials not released together with stack gas. The concept on which this rule is based is to ensure that in the case of incidents during which the result of effluent monitoring remains meaningful, such monitoring can be reliably performed. (orig./HSCH) [de

  6. Consequences of severe radioactive releases to Nordic Marine environment

    DEFF Research Database (Denmark)

    Iosjpe, M.; Isaksson, M.; Joensen, H.P.

    - or minor – radioactive releases to Nordic marine environment. As a reference, the release amounts from a 3000 MWth reactor size were used. Based on source term analyses, the chosen release fractions in the study were: iodine 20% (of the total core inventory), caesium 10%, tellurium 10%, strontium 0...

  7. Future radioactive liquid waste streams study

    Energy Technology Data Exchange (ETDEWEB)

    Rey, A.S.

    1993-11-01

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

  8. Future radioactive liquid waste streams study

    International Nuclear Information System (INIS)

    Rey, A.S.

    1993-11-01

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

  9. Immobilization of low and intermediate level radioactive liquid wastes using some industrial by-product materials

    International Nuclear Information System (INIS)

    Sami, N.M.; EI-Dessouky, M.I.; Abou EI-Nour, F.H.; Abdel-Khalik, M.

    2006-01-01

    Immobilization of low and intermediate level.radioactive liquid wastes in different matrices: ordinary Portland cement and cement mixed with some industrial byproduct: by-pass kiln cement dust, blast furnace slag and ceramic sludge was studied. The effect of these industrial by-product materials on the compressive strength, water immersion, radiation effect and teachability were investigated. The obtained results showed that, these industrial by-product improve the cement pastes where they increase the compressive strength, decrease the leaching rate for radioactive cesium-137 and cobalt-60 ions through the solidified waste forms and increase resistance for y-radiation. It is found that, solidified waste forms of intermediate level liquid waste (ILLW) had high compressive strength values more than those obtained from low level liquid waste (LLLW). The compressive strength increased after immersion in different leachant for one and three months for samples with LLLW higher than those obtained for ILLW. The cumulative fractions released of cesium-137 and cobalt-60 of solidified waste forms of LLLW was lower than those obtained for ILLW

  10. Applied exposure modeling for residual radioactivity and release criteria

    International Nuclear Information System (INIS)

    Lee, D.W.

    1989-01-01

    The protection of public health and the environment from the release of materials with residual radioactivity for recycle or disposal as wastes without radioactive contents of concern presents a formidable challenge. Existing regulatory criteria are based on technical judgment concerning detectability and simple modeling. Recently, exposure modeling methodologies have been developed to provide a more consistent level of health protection. Release criteria derived from the application of exposure modeling methodologies share the same basic elements of analysis but are developed to serve a variety of purposes. Models for the support of regulations for all applications rely on conservative interpretations of generalized conditions while models developed to show compliance incorporate specific conditions not likely to be duplicated at other sites. Research models represent yet another type of modeling which strives to simulate the actual behavior of released material. In spite of these differing purposes, exposure modeling permits the application of sound and reasoned principles of radiation protection to the release of materials with residual levels of radioactivity. Examples of the similarities and differences of these models are presented and an application to the disposal of materials with residual levels of uranium contamination is discussed. 5 refs., 2 tabs

  11. Method of processing radioactive liquid wastes by solidification with cement

    International Nuclear Information System (INIS)

    Yasumura, Keijiro; Matsuura, Hiroyuki.

    1975-01-01

    Object: To subject radioactive liquid wastes to a cement solidification treatment after heating and drying it by a thin film scrape-off drier to render it into the form of power, and then molding it into pellets for the treatment. Structure: Radioactive liquid wastes discharged from a nuclear power plant or nuclear reactor are supplied through a storage tank into a thin film scrape-off drier. In the drier, the radioactive liquid wastes are heated to separate the liquid, and the residue is taken out as dry powder from the scrape-off apparatus. The powder obtained in this way is molded into pellets of a desired form. These pellets are then packed in a drum can or similar container, into which cement paste is then poured for solidification. (Moriyama, K.)

  12. Cementation of liquid radioactive waste with high content of borate salts

    International Nuclear Information System (INIS)

    Gorbunova, O.

    2015-01-01

    The report reviews the ways of optimization of cementation of boron-containing liquid radioactive waste. The most common way to hardening the low-level liquid radioactive waste (LRW) is the cementation. However, boron-containing liquid radioactive waste with low pH values cannot be cemented without alkaline additives, to neutralize acid forms of borate compounds. Cement setting without additives happens only on 14-56 days, the compounds have low strength, and hence an insufficient reliability of radionuclides fixation in the cement matrix. The alkaline additives increase the volume of the final cement compound which enhances financial and operational costs. In order to control the speed of hardening of cement solution with a boron-containing liquid radioactive waste and to remove the components that prevent hardening of cement solution, it is proposed an electromagnetic treatment of LRW in the vortex layer of ferromagnetic particles. The results of infrared spectroscopy show, that electromagnetic treatment of liquid radioactive waste changes the ionic forms of the borates and raises the pH due to the dissociation of the oxygen and hydrogen bonds in the aqueous solutions of the boron compounds. The various types of ferromagnetic activators of the vortex layer have been investigated, including the highly dispersed nano-powders and the magnetic phases of the iron oxides. It has been determined the technological parameters of the electromagnetic treatment of liquid radioactive waste and the subsequent cementation of this type of LRW. By using the method of scanning electron microscopy it has been shown, that the nano-particles of magnetic phases of the ferric oxides are involved in phase formation of hydro-aluminum-calcium ferrites in the early stages of hardening and improving strength of the cement compounds with liquid radioactive waste. (authors)

  13. Management of radioactive wastes (solids and liquids) of CDTN

    International Nuclear Information System (INIS)

    Prado, M.A.S. do; Reis, L.C.A.

    1984-01-01

    Estimates of solid and liquid radioactive wastes produced in CDTN, the foreseen treatment and the responsibilities of various organs of CDTN involved in radioactive waste management are presented. (C.M.)

  14. Radioactive liquid waste processing method

    International Nuclear Information System (INIS)

    Nishi, Takashi; Baba, Tsutomu; Fukazawa, Tetsuo; Matsuda, Masami; Chino, Koichi; Ikeda, Takashi.

    1993-01-01

    As an adsorbent used for removing radioactive nuclides such as cesium and strontium from radioactive liquid wastes generated from a reprocessing plant, a silicon compound having siloxane bonds constituted by silicon and oxygen and having silanol groups constituted by silicon, oxygen and hydrogen, or an inorganic material mainly comprising aluminosilicate constituted with silicon, oxygen and aluminum is used. In the adsorbent of the present invention, since silica main skeletons are partially decomposed in an aqueous alkaline solution to newly form silanol groups having a cation adsorbing property, pretreatment such as pH adjustment is not necessary. (T.M.)

  15. Response to a radioactive materials release having a transboundary impact

    International Nuclear Information System (INIS)

    1989-01-01

    Compared with an accidental release of radioactive material which is confined to the accident State, a transboundary release has added dimensions which were not fully anticipated in publications dealing with response to accidents at nuclear facilities. The new aspects to the problem may be summarized as follows: (1) A transboundary release of radioactive material, as distinct from a release which affects only the accident State, has international repercussions in the following ways: Potentially at least, the difficulties associated with a transboundary release may be magnified in those States that have no nuclear facilities of their own and may, therefore, have foreseen no need for resources to assess and deal with radioactive contamination of their food supplies, their water and their environment appropriately. International trade, in food commodities particularly, may be severely affected. Issues of compensation may arise for which the dispute settlement mechanisms are weak or non-existent. (2) Many Member States are in such geographic locations that they could be affected by a transboundary release occurring in any of their surrounding neighbour States. Planning for and responding to such an event is necessarily more difficult than planning for an accidental release from a single, identified nuclear facility. (3) Deposits of radioactive material from a distant source are apt to be highly unpredictable. Depending on weather conditions, they may be localized in a random fashion or widespread. Because of the international dimension of the problem and its essentially unpredictable character it is recommended here that planning for such events should be carried at the national or federal government level rather than at provincial government level. 14 refs

  16. Radioactive releases from a thorium-contaminated site in Wayne, New Jersey

    International Nuclear Information System (INIS)

    Wang, J.; Yang, J.; Merry-Libby, P.

    1985-01-01

    Various residues and wastes from the production of thorium and rare earths from monazite ore are buried on a hillside in Wayne, New Jersey. In addition, contaminated materials (primarily soils) from nearby vicinity properties are being consolidated onto the Wayne site. The US Department of Energy plans to stabilize all the contaminated materials on an interim basis (20 years) until funding is available to remove them to another location. In order to evaluate the effectiveness of interim stabilization measures, pre-remedial action radioactive releases are compared to estimated releases under a reference stabilization option (one meter of soil cover). Two potential pathways are examined: (1) airborne radioactive gases (thoron and radon) and particulates, and (2) seepage into the near-surface groundwater. The relative reduction of releases into the air and groundwater for the reference stabilization option is analyzed using mathematical models for radioactive gas fluxes and atmospheric dispersion as well as groundwater transport and dispersion. The consequent health implications for nearby individuals and the general population are also estimated. Health effects due to radioactive releases are estimated to be insignificant

  17. Released radioactivity reducing device

    International Nuclear Information System (INIS)

    Miyamoto, Yumi.

    1995-01-01

    A water scrubber is disposed in a scrubber tank and a stainless steel fiber filter is disposed above the water scrubber. The upper end of the scrubber tank is connected by way of a second bent tube to a capturing vessel incorporating a moisture removing layer and an activated carbon filter. The exit of the capturing vessel is connected to a stack. Upon occurrence of an accident of a BWR-type power plant, gases containing radioactive materials released from a reactor container are discharged into the water scrubber from a first bent tube through a venturi tube nozzle, and water soluble and aerosol-like radioactive materials are captured in the water. Aerosol and splashes of water droplets which can not be captured thoroughly by the water scrubber are captured by the stainless steel fiber filter. Gases passing through the scrubber tank are introduced to a capturing vessel through a second bent tube, and organic iodine is captured by the activated carbon filter. (I.N.)

  18. Screening calculations for radioactive waste releases from non-nuclear facilities

    International Nuclear Information System (INIS)

    Xu, Shulan; Soederman, Ann-Louis

    2009-02-01

    A series of screening calculations have been performed to assess the potential radiological consequences of discharges of radioactive substances to the environment arising from waste from non-nuclear practices. Solid waste, as well as liquids that are not poured to the sewer, are incinerated and ashes from incineration and sludge from waste water treatment plants are disposed or reused at municipal disposal facilities. Airborne discharges refer to releases from an incineration facility and liquid discharges refer both to releases from hospitals and laboratories to the sewage system, as well as leakage from waste disposal facilities. The external exposure of workers is estimated both in the waste water treatment plant and at the disposal facility. The calculations follow the philosophy of the IAEA's safety guidance starting with a simple assessment based on very conservative assumptions which may be iteratively refined using progressively more complex models, with more realistic assumptions, as necessary. In the assessments of these types of disposal, with cautious assumptions, carried out in this report we conclude that the radiological impacts on representative individuals in the public are negligible in that they are small with respect to the target dose of 10 μSv/a. A Gaussian plume model was used to estimate the doses from airborne discharges from the incinerator and left a significant safety margin in the results considering the conservative assumptions in the calculations. For the sewage plant workers the realistic approach included a reduction in working hours and the shorter exposure time resulted in maximum doses around 10 μSv/a. The calculations for the waste disposal facility show that the doses are higher or in the range of the target dose. The excess for public exposure is mainly caused by H-3 and C-14. The assumption used in the calculation is that all of the radioactive substances sent to the incineration facility and waste water treatment plant

  19. Screening calculations for radioactive waste releases from non-nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Shulan Xu; Soederman, Ann-Louis

    2009-02-15

    A series of screening calculations have been performed to assess the potential radiological consequences of discharges of radioactive substances to the environment arising from waste from non-nuclear practices. Solid waste, as well as liquids that are not poured to the sewer, are incinerated and ashes from incineration and sludge from waste water treatment plants are disposed or reused at municipal disposal facilities. Airborne discharges refer to releases from an incineration facility and liquid discharges refer both to releases from hospitals and laboratories to the sewage system, as well as leakage from waste disposal facilities. The external exposure of workers is estimated both in the waste water treatment plant and at the disposal facility. The calculations follow the philosophy of the IAEA's safety guidance starting with a simple assessment based on very conservative assumptions which may be iteratively refined using progressively more complex models, with more realistic assumptions, as necessary. In the assessments of these types of disposal, with cautious assumptions, carried out in this report we conclude that the radiological impacts on representative individuals in the public are negligible in that they are small with respect to the target dose of 10 muSv/a. A Gaussian plume model was used to estimate the doses from airborne discharges from the incinerator and left a significant safety margin in the results considering the conservative assumptions in the calculations. For the sewage plant workers the realistic approach included a reduction in working hours and the shorter exposure time resulted in maximum doses around 10 muSv/a. The calculations for the waste disposal facility show that the doses are higher or in the range of the target dose. The excess for public exposure is mainly caused by H-3 and C-14. The assumption used in the calculation is that all of the radioactive substances sent to the incineration facility and waste water treatment

  20. Liquid radioactive waste processing system for pressurized water reactor plants

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    This Standard sets forth design, construction, and performance requirements, with due consideration for operation, of the Liquid Radioactive Waste Processing System for pressurized water reactor plants for design basis inputs. For the purpose of this Standard, the Liquid Radioactive Waste Processing System begins at the interfaces with the reactor coolant pressure boundary and the interface valve(s) in lines from other systems, or at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material; and it terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system, and at the point of recycle back to storage for reuse

  1. The cleanup of releases of radioactive materials from commercial low-level radioactive waste disposal sites: Whose jurisdiction?

    International Nuclear Information System (INIS)

    Hartnett, C.

    1994-01-01

    There exists an overlap between the Comprehensive Environmental Response, Compensation and Recovery Act (open-quotes CERCLAclose quotes) and the Atomic Energy Act (open-quotes AEAclose quotes) regarding the cleanup of releases of radioactive materials from commercial low-level radioactive waste sites. The Nuclear Regulatory Commission (open-quotes NRCclose quotes) and Agreement States have jurisdiction under the AEA, and the Environmental Protection Agency (open-quotes EPAclose quotes) has jurisdiction pursuant to CERCLA. This overlapping jurisdiction has the effect of imposing CERCLA liability on parties who have complied with AEA regulations. However, CERCLA was not intended to preempt existing legislation. This is evidenced by the federally permitted release exemption, which explicitly exempts releases from CERCLA liability pursuant to an AEA license. With little guidance as to the applicability of this exemption, it is uncertain whether CERCLA's liability is broad enough to supersede the Atomic Energy Act. It is the purpose of this paper to discuss the overlapping jurisdiction for the cleanup of releases of radioactive materials from commercial low-level radioactive waste disposal sites with particular emphasis on the cleanup at the Maxey Flats, West Valley and Sheffield sites

  2. Method of processing radioactive liquid wastes

    International Nuclear Information System (INIS)

    Kurumada, Norimitsu; Shibata, Setsuo; Wakabayashi, Toshikatsu; Kuribayashi, Hiroshi.

    1984-01-01

    Purpose: To facilitate the procession of liquid wastes containing insoluble salts of boric acid and calcium in a process for solidifying under volume reduction of radioactive liquid wastes containing boron. Method: A soluble calcium compound (such as calcium hydroxide, calcium oxide and calcium nitrate) is added to liquid wastes whose pH value is adjusted neutral or alkaline such that the molar ratio of calcium to boron in the liquid wastes is at least 0.2. Then, they are agitated at a temperature between 40 - 70 0 C to form insoluble calcium salt containing boron. Thereafter, the liquid is maintained at a temperature less than the above-mentioned forming temperature to age the products and, thereafter, the liquid is evaporated to condensate into a liquid concentrate containing 30 - 80% by weight of solid components. The concentrated liquid is mixed with cement to solidify. (Ikeda, J.)

  3. Method of processing liquid wastes containing radioactive materials

    International Nuclear Information System (INIS)

    Matsumoto, Kaname; Shirai, Takamori; Nemoto, Kuniyoshi; Yoshikawa, Jun; Matsuda, Takeshi.

    1983-01-01

    Purpose: To reduce the number of solidification products by removing, particularly, Co-60 that is difficult to remove in a radioactive liquid wastes containing a water-soluble chelating agent, by adsorbing Co-60 to a specific chelating agent. Method: Liquid wastes containing radioactive cobalt and water-soluble chelating agent are passed through the layer of less water-soluble chelating agent that forms a complex compound with cobalt in an acidic pH region. Thus, the chelating compound of radioactive cobalt (particularly Co-60) is eliminated by adsorbing the same on a specific chelating agent layer. The chelating agent having Co-60 adsorbed thereon is discarded as it is through the cement- or asphalt-solidification process, whereby the number of solidification products to be generated can significantly be suppressed. (Moriyama, K.)

  4. Radioactivity monitor for high-performance liquid chromatography

    International Nuclear Information System (INIS)

    Reeve, D.R.; Crozier, A.

    1977-01-01

    The coupling of a homogeneous radioactivity monitor to a liquid chromatograph involves compromises between the sensitivity of the monitor and the resolution and speed of analysis of the chromatograph. The theoretical relationships between these parameters are considered and expressions derived which make it possible to calculate suitable monitor operating conditions for most types of high-performance liquid chromatography

  5. Solid and liquid radioactive waste treatment

    International Nuclear Information System (INIS)

    Rzyski, B.M.

    1989-01-01

    The technology for the treatment of low - and intermediate-level radioactive solid and liquid wastes is somewhat extensive. Some main guidance on the treatment methods are shown, based on informations contained in technical reports and complementary documents. (author) [pt

  6. Releases of radioactivity at the Savannah River Plant, 1954--1975

    International Nuclear Information System (INIS)

    Ashley, C.; Zeigler, C.C.

    1976-07-01

    This report contains summaries of releases of radioactivity to the environs from the Savannah River Plant for each year since plant startup (1954-1975). It also contains monthly summaries of 1975 releases from major emission sources (Separations and Reactor Areas). Releases for the years 1954 through 1959 are reassessed and assigned release values for long-lived specific radionuclides. These long-lived radionuclides (half lives greater than one year) are the only radionuclides included for the years 1954 through 1970. Beginning in 1971 all detectable radionuclides are included. The tabulated data, now compiled by computer, will be updated annually. All measured migration of radioactivity from the F- and H-Area seepage basin systems to Four Mile Creek, and from the K-Area containment basin to Pen Branch is listed in the release summaries and will be updated annually

  7. Filters for radioactive liquid wastes

    International Nuclear Information System (INIS)

    Koshiba, Yukihiko; Kawashima, Akio

    1980-01-01

    In the crud generated in the reactor cooling water for nuclear power plants, iron oxides (hematite and magnetite) are contained as the main components, and also Co, Mn, Fe, Cr exist as radioactive nuclides. A new filter to separate these cruds, nuclepore membrane filter (NPMF), was investigated for its adaptability, and has been adopted as a practical filter for radioactive liquid wastes. The NPMF has such features as the possibility of complete automation of operation, no generation of secondary wastes, and easy maintenance, because the NPMF has uniform circular holes in poly-carbonate thin films, and shows the properties of stable filtering of particulates, capability of back washing, and others. The elements mounted in a practical system have such construction that the membrane is cut in the form of doughnut, and sandwiched with 100 mesh polyester nets (spacer); the obtained unit filter (cassette) is mounted on the stackable plate of the same size; and 80 pieces of this cassette are formed in a filter of 4 m 2 filtering area. The performance varies with the properties of suspended matters and the turbidity of wastes. For example, the filtered liquid of 0.1 ppm or less can be obtained when the 1 μm filter material is used to treat the liquid waste containing 1 to 100 ppm suspended matters. Usually back washed water is produced by about 1/100 of treated liquid wastes. The lifetime of the membrane is expected to be 1 or 2 years if crud is the main component. (Wakatsuki, Y.)

  8. The environmental impact of radioactive releases from accidents in nuclear power reactors

    International Nuclear Information System (INIS)

    Beattie, J.R.; Griffiths, R.F.; Kaiser, G.D.; Kinchin, G.H.

    1978-01-01

    A survey of accidental releases of radioactivity from thermal and fast reactors is presented. Following a general discussion on the hazards involved, the nature of the environmental impact of radioactive releases is examined. This includes a brief review of the natural radiation background, the effect on human health of various levels of radiation and radioactivity, permissible and reference levels, and the type of hazards from both passing clouds of airbourne radioactive material and from ground deposited material. The problem of atmospheric dispersion and methods of calculations of radioactive materials in the atmosphere are examined in order for the consequences of accidental release to be analysed. National accidents and their environmental consequences are then examined. Finally there is a review of the risks to which man is always exposed because of his environment. Common and collective risks are also considered. Conclusions are reached as to the acceptibility or otherwise of the environmental impact of reactor accidents. (U.K.)

  9. Low level radioactive liquid waste decontamination by electrochemical way

    International Nuclear Information System (INIS)

    Tronche, E.

    1994-10-01

    As part of the work on decontamination treatments for low level radioactive aqueous liquid wastes, the study of an electro-chemical process has been chosen by the C.E.A. at the Cadarache research centre. The first part of this report describes the main methods used for the decontamination of aqueous solutions. Then an electro-deposition process and an electro-dissolution process are compared on the basis of the decontamination results using genuine radioactive aqueous liquid waste. For ruthenium decontamination, the former process led to very high yields (99.9 percent eliminated). But the elimination of all the other radionuclides (antimony, strontium, cesium, alpha emitters) was only favoured by the latter process (90 percent eliminated). In order to decrease the total radioactivity level of the waste to be treated, we have optimized the electro-dissolution process. That is why the chemical composition of the dissolved anode has been investigated by a mixture experimental design. The radionuclides have been adsorbed on the precipitating products. The separation of the precipitates from the aqueous liquid enabled us to remove the major part of the initial activity. On the overall process some operations have been investigated to minimize waste embedding. Finally, a pilot device (laboratory scale) has been built and tested with genuine radioactive liquid waste. (author). 77 refs., 41 tabs., 55 figs., 4 appendixes

  10. Bituminization of liquid radioactive wastes. Part 1

    International Nuclear Information System (INIS)

    Gradev, G.D.; Ivanov, V.I.; Stefanova, I.G.; Milusheva, A.G.; Guteva, E.S.; Zhelyazkov, V.T.; Stefanov, G.I.; G'oshev, G.S.

    1991-01-01

    Salt-bitumen products are produced by the method of 'hot mixing' of some Bulgarian bitumens (road bitumen PB 66/99 and the hydroinsulating bitumen HB 80/25) and salts (chlorides, sulphates, borates, salt mixtures modelling the liquid waste from nuclear power plants) in different ratios to determine the optimum conditions for bituminization of liquid radioactive waste. The penetration, ductility and softening temperature were determined. The sedimentation properties and the thermal resistance of the various bitumen-salt mixtures were studied. The most suitable bitumen for technological research at the Kozloduy NPP was found to be the road bitumen PB 66/90 with softening temperature at 48 o C. The optimum amount of salts incorporated in the bitumen - about 45% - was found. No exothermal effects were observed in the bituminization process in the temperature range of up to 200 o C. The results obtained may be useful in the elaboration of a technology for bituminization of liquid radioactive wastes in the Kozloduy NPP. 4 tabs., 5 figs., 4 refs

  11. Methodology for estimating accidental radioactive releases in nuclear waste management

    International Nuclear Information System (INIS)

    Levy, H.B.

    1979-01-01

    Estimation of the risks of accidental radioactive releases is necessary in assessing the safety of any nuclear waste management system. The case of a radioactive waste form enclosed in a barrier system is considered. Two test calculations were carried out

  12. Method for the disposal of radioactive waste liquids

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Y; Kamiya, K; Kuriyama, O

    1976-03-19

    A method is presented to solidify radioactive waste liquids such as washing liquids containing radioactive material generated in an atomic power plant to thereby facilitate transport of them. A drum can is inserted into a drum can supporting vessel and carried by a truck toward and under the evaporation chamber. A lifter is upwardly extended by an elevator to provide an intimate contact between the lower end of a steam chamber and the upper end of the drum can through a seal ring. Next, a mixture of a washing waste liquid and a defoaming agent is filled from a supply pipe into the drum can in spraying manner. Into a heater is supplied heated vapor from a heated vapor supply pipe to vaporize and condense the waste liquids. The vaporized vapor passes through a demister and is condensed by a condenser. After the condensed liquids of a predetermined concentration have been obtained, a lifter is retracted to cause the drum can to be moved under a cement mixer to feed cement into the drum can for mixing and solidifying it therein.

  13. Treatment of low- and intermediate-level liquid radioactive wastes

    International Nuclear Information System (INIS)

    1984-01-01

    This report aims at giving the reader details of the experience gained in the treatment of both low- and intermediate-level radioactive liquid wastes. The treatment comprises those operations to remove radioactivity from the wastes and those that change only its chemical composition, so as to permit its discharge. Considerable experience has been accumulated in the satisfactory treatment of such wastes. Although there are no universally accepted definitions for low- and intermediate-level liquid radioactive wastes, the IAEA classification (see section 3.2) is used in this report. The two categories differ from one another in the fact that for low-level liquids the actual radiation does not require shielding during normal handling of the wastes. Liquid wastes which are not considered in this report are those from mining and milling operations and the high-level liquid wastes resulting from fuel reprocessing. These are referred to in separate IAEA reports. Likewise, wastes from decommissioning operations are not within the scope of this report. Apart from the description of existing methods and facilities, this report is intended to provide advice to the reader for the selection of appropriate solutions to waste management problems. In addition, new and promising techniques which are either being investigated or being considered for the future are discussed

  14. Radioactivity measurements by liquid scintillation spectroscopy

    International Nuclear Information System (INIS)

    Cassette, Ph.

    2004-01-01

    The activity measurement techniques by liquid scintillation spectroscopy consist to mix the radioactive solution to measure with a scintillating liquid and to transform the ionizing radiations, resulting from decays, into light, detectable and quantifiable. The main advantages of these techniques are the easiness of preparation of the radioactive sources, the geometric efficiency of detection of 4π and the possibility of detection of low-level energy radiations. There are one of the only methods giving the possibility to measure the activity of pure β radionuclides; indeed, the nuclear disintegration is not accompanied of gamma radiations detectable by other techniques. There are one of the only methods too of measurement of radionuclides which disintegrate by electron capture and especially those leading to the emission of low-level energy ionizing radiations. Liquid scintillation spectroscopy can be used as an absolute method of activity measurement that is to say without the use of a calibration standard. The modern liquid scintillation counting devices can be very sensitive; the measurement of micro-activities being possible. Some of the applications of these activity measurement techniques are the carbon 14 dating and the geological tracing. Their main disadvantage is the global energetic yield which is low and variable in terms of the composition of the scintillation source necessitating to calculate the detection yield for each condition of measurement. (O.M.)

  15. Leak test of the pipe line for radioactive liquid waste

    International Nuclear Information System (INIS)

    Machida, Chuji; Mori, Shoji.

    1976-01-01

    In the Tokai Research Establishment, most of the radioactive liquid waste is transferred to a wastes treatment facility through pipe lines. As part of the pipe lines a cast iron pipe for town gas is used. Leak test has been performed on all joints of the lines. For the joints buried underground, the test was made by radioactivity measurement of the soil; and for the joints in drainage ditch by the pressure and bubble methods. There were no leakage at all, indicating integrity of all the joints. On the other hand, it is also known by the other test that the corrosion of inner surface of the piping due to liquid waste is only slight. The pipe lines for transferring radioactive liquid waste are thus still usable. (auth.)

  16. Generation and release of radioactive gases in LLW disposal facilities

    Energy Technology Data Exchange (ETDEWEB)

    Yim, M.S. [Harvard School Public Health, Boston, MA (United States); Simonson, S.A. [Massachusetts Institute of Technology, Cambridge, MA (United States)

    1995-02-01

    The atmospheric release of radioactive gases from a generic engineered LLW disposal facility and its radiological impacts were examined. To quantify the generation of radioactive gases, detailed characterization of source inventory for carbon-14, tritium, iodine-129, krypton-85, and radon-222, was performed in terms of their activity concentrations; their distribution within different waste classes, waste forms and containers; and their subsequent availability for release in volatile or gaseous form. The generation of gases was investigated for the processes of microbial activity, radiolysis, and corrosion of waste containers and metallic components in wastes. The release of radionuclides within these gases to the atmosphere was analyzed under the influence of atmospheric pressure changes.

  17. Action plan for response to abnormal conditions in Hanford high level radioactive liquid waste storage tanks containing flammable gases

    International Nuclear Information System (INIS)

    Sherwood, D.J.

    1994-03-01

    Radioactive liquid waste tends to produce hydrogen as a result of the interaction of gamma radiation and water. In tanks containing organic chelating agents, additional hydrogen gas as well as nitrous oxide and ammonia can be produced by thermal and radiolytic decomposition of these organics. Several high-level radioactive liquid waste storage tanks, located underground at the Hanford Site, contain waste that retains the gases produced in them until large quantities are released rapidly to the tank vapor space. Tanks filled to near capacity have relatively little vapor space; therefore, if the waste suddenly releases a large amount of hydrogen and nitrous oxide, a flammable gas mixture may result. The most notable waste tank with a flammable gas problem is tank 241-SY-101. Waste in this tank has occasionally released enough flammable gas to burn if an ignition source had been present inside of the tank. Several other waste tanks exhibit similar behavior to a lesser magnitude. Administrative controls have been developed to assure that these Flammable Gas Watch List tanks are safely maintained. Responses have also been developed for off-normal conditions which might develop in these tanks. In addition, scientific and engineering studies are underway to further understand and mitigate the behavior of the Flammable Gas Watch List tanks

  18. Regulatory review of releases from HIFAR of radioactive airborne effluent

    International Nuclear Information System (INIS)

    Westall, D.J.; Macnab, D.I.

    1996-01-01

    Full text: The Nuclear Safety Bureau (NSB) was set up by legislation in 1992 as an independent Commonwealth corporate body reporting to the Minister for Health and Family Services. Its functions include monitoring and reviewing the safety of nuclear plant owned or operated by the Australian Nuclear Science and Technology Organisation (ANSTO). The NSB sets requirements for authorisation of the operation of the HIFAR research reactor, and may impose restrictions and conditions on its operation. The authorisation for the operation of HIFAR includes a requirement for arrangements for the treatment, safe storage and disposal of solid, liquid and gaseous radioactive wastes from the reactor. The objective is to establish conditions which would ensure that radiation exposure to plant personnel and the public from radioactive wastes are within acceptable limits and that releases are maintained as low as reasonably achievable. The NSB has developed expectations based on international best practice, against which to review HIFAR's arrangements for satisfying the requirement and achieving the objective. Arrangements for the release of airborne radioactive effluent from HIFAR were reviewed by the NSB as part of an overall review of the upgrade of safety documentation for HIFAR. The NSB's expectations for the review were drawn from the International Atomic Energy Agency (IAEA) Basic Safety Standards (Safety Series No 115-I) and the National Health and Medical Research Council (NHMRC) Recommendations for Limiting Exposure to Ionizing Radiation (1995). These expectations included a hierarchy of primary dose limits, stack discharge limits and reference levels for HIFAR aimed at ensuring that radiation doses to the public due to airborne effluent are less than the national dose limits and ANSTO's dose constraints, and are as low as reasonably achievable. An approach favoured by the operator is to work directly to a primary dose limit using an airborne dispersion computer program to

  19. Method of processing radioactive liquid wastes

    International Nuclear Information System (INIS)

    Matsuura, Hiroyuki; Kuribayashi, Nobuhide; Minami, Yuji; Kamiyama, Hisashi

    1979-01-01

    Purpose: To greatly reduce the quantity of radioactive liquid wastes by subjecting the same to drying treatment, and to granulate the thus formed dry powders to prevent scattering thereof thereby to fill a storage vessel safely with the powders without contaminating the surroundings. Constitution: Radioactive liquid wastes within a storage tank are supplied to a drier where the wastes are subjected to evaporation treatment, and pulverized. The thus dried powders are temporarily stored in a hopper by means of a screw feeder. The dry powders which have reached a predetermined quantity are supplied to a stirrer-granulator by means of a quantitative screw feeder, and mixed and stirred with a binder sent from a binder storage tank through a binder quantity determining device, whereby the powders are granulated. After the granulation, the granulated powders are extruded by a centrifugal force, and filled in the storage vessel by way of a conduit. (Yoshino, Y.)

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

    International Nuclear Information System (INIS)

    Christensen, B. D.

    1999-01-01

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

  1. Design of mobile receiving and treatment equipment for radioactive liquid waste

    International Nuclear Information System (INIS)

    Kong Jinsong; Guo Weiqun; Lu Jingbin

    2012-01-01

    The advantage and disadvantage of radioactive liquid waste treatment technology are analyzed in this paper. The experimental disposal equipment for radioactive liquid waste with complicated sources is designed by combining the far infrared calcification technology with evaporation technology. It has advantages of low energy consuming and high decontamination efficiency. The frothy and dirt appear rarely in this equipment. (authors)

  2. Liquid Radioactive Wastes Treatment: A Review

    Directory of Open Access Journals (Sweden)

    Yung-Tse Hung

    2011-05-01

    Full Text Available Radioactive wastes are generated during nuclear fuel cycle operation, production and application of radioisotope in medicine, industry, research, and agriculture, and as a byproduct of natural resource exploitation, which includes mining and processing of ores, combustion of fossil fuels, or production of natural gas and oil. To ensure the protection of human health and the environment from the hazard of these wastes, a planned integrated radioactive waste management practice should be applied. This work is directed to review recent published researches that are concerned with testing and application of different treatment options as a part of the integrated radioactive waste management practice. The main aim from this work is to highlight the scientific community interest in important problems that affect different treatment processes. This review is divided into the following sections: advances in conventional treatment of aqueous radioactive wastes, advances in conventional treatment of organic liquid wastes, and emerged technological options.

  3. Aspects of chemistry in management of radioactive liquid wastes from nuclear installations

    International Nuclear Information System (INIS)

    Yeotikar, R.G.

    2007-01-01

    Nuclear energy is the only source available to the mankind to fulfill the continuous and ever increasing demand of energy. The public acceptance and popularity of nuclear energy depends to a large extent on management of radioactive waste. The nuclear waste management demands eco-friendly process/systems. This article highlights the sources of different types of radioactive liquid wastes generated in the nuclear installation and their treatment process. The radioactive liquid waste is classified mainly into three categories based on activity levels e.g. low, intermediate and high level. The management of radioactive liquid waste is very critical because of its 'mobility and liquid' nature. Secondly the liquid wastes have wide range of activity and chemistry spectrum and their volumes are also different. Hence the methods for management of different types of liquid wastes are also different. Mostly the treatment and conditioning processes are chemical processes. The chemistry involved in the treatment and conditioning of these wastes, problems related with chemistry for each processes and efforts to solve these problems, aspects of adoption on plant scale, etc., have been discussed in this article. (author)

  4. Radioactive contamination mapping system detailed design report

    International Nuclear Information System (INIS)

    Bauer, R.G.; O'Callaghan, P.B.

    1996-08-01

    The Hanford Site's 100 Area production reactors released radioactively and chemically contaminated liquids into the soil column. The primary source of the contaminated liquids was reactor coolant and various waste waters released from planned liquid discharges, as well as pipelines, pipe junctions, and retention basins leaking into the disposal sites. Site remediation involves excavating the contaminated soils using conventional earthmoving techniques and equipment, treating as appropriate, transporting the soils, and disposing the soils at ERDF. To support remediation excavation, disposal, and documentation requirements, an automated radiological monitoring system was deemed necessary. The RCMS (Radioactive Contamination Mapping System) was designed to fulfill this need. This Detailed Design Report provides design information for the RCMS in accordance with Bechtel Hanford, Inc. Engineering Design Project Instructions

  5. A method for prevention of radioactive material release

    International Nuclear Information System (INIS)

    Uchida, Shunsuke; Sato, Chikara; Kitamura, Masao.

    1975-01-01

    Object: To provide a method for preventing an underwater radioactive material from being released in a simple and highly reliable manner, which can decrease an amount of radioactive materials discharged into open air from reactor water containing a large amount of radioactive materials such as a reactor core pool. Structure: Pure warm water higher in temperature than that of reactor water is poured from the top of a water surface of a water tank which stores reactor water containing radioactive materials such as radioactive iodine, and water is drawn through an outlet located downwardly of the pure warm water inlet to form a layer of pure warm water at the upper part of the water tank while preventing diffusion of the reactor water into the pure warm water by the difference in density between the reactor water and the pure warm water and downward movement of the pure warm water, thereby preventing contact of the reactor water with the atmosphere and diffusion of the radioactive material into the atmosphere. (Kamimura, M.)

  6. Assessment of consequences from airborne releases of radioactive material

    International Nuclear Information System (INIS)

    McGrath, P.E.; Blond, R.M.

    1976-01-01

    Over the past several years, the manner in which assessments have been made of the consequences of large airborne releases of radioactive material has not changed much conceptually. The models to describe the atmospheric dispersion of the radioactive material have generally been time-invariant, i.e., the meteorological conditions (thermal stability, wind speed, and precipitation) are invariant during release and the subsequent period of radiation exposure of the population to the airborne material. The frequency distribution of the meteorological conditions are determined by analyzing several years of weather data from the appropriate geographical location. In reality, weather is continuously changing over short time periods (hours) following the release. It is to be expected that the changing meteorological conditions would have important effects on the potential consequences of the release. A time-dependent atmospheric dispersion model was developed and implemented in the Reactor Safety Study. This paper provides a description of the model and the nature of the results generated. Emphasis is given to an explanation of how, and why, these results differ from those estimated with time-invariant models

  7. Chemical treatment of radioactive liquid wastes from medical applications

    International Nuclear Information System (INIS)

    Castillo A, J.

    1995-01-01

    This work is a study about the treatment of the most important radioactive liquid wastes from medical usages, generated in medical institutions with nuclear medicine services. The radionuclides take in account are 32 P, 35 S, 125 I. The treatments developed and improved were specific chemical precipitations for each one of the radionuclides. This work involve to precipitate the radionuclide from the liquid waste, making a chemical compound insoluble in the aqueous phase, for this process the radionuclide stay in the precipitate, lifting the aqueous phase with a very low activity than the begin. The 32 P precipitated in form of Ca 3 32 P O 4 and Ca 2 H 32 P O 4 with a value for Decontamination Factor (DF) at the end of the treatment of 32. The 35 S was precipitated in form of Ba 35 SO 4 with a DF of 26. The 125 I was precipitated in Cu 125 I to obtain a DF of 24. The results of the treatments are between the limits given for the International Atomic Energy Agency and the 10 Code of Federal Regulation 20, for the safety release at the environment. (Author)

  8. Radiological assessment and management of radioactive spill in a liquid waste treatment facility - Case study

    International Nuclear Information System (INIS)

    Amer, H.A.; Shawky, S.; Ibrahiem, N.

    2002-01-01

    The radiological assessment and management of radioactive spill from liquid waste treatment facility is presented. The incident contaminated the area surrounding the treatment facility with various radionuclides, which were dispersed into the soil. A method based on the European basic safety standards was used to contain the risks associated with the contaminated site. The introduced case study proceeded up to the stage of simplified risk study, since the site is small and it was relatively easy to remove and store the contaminated soil. According to the obtained results, the removal of the upper 30-cm would be considered as appropriate remedying action to resume background level. One of the most important basic concepts of radiation protection in nuclear facilities is the continuity of monitoring radiological release to the environment. It is known that from nuclear facilities only very small amounts of radioactivity are discharged with the liquid effluents and the exhaust air into the environment. Recent studies screening the natural and artificial radionuclide in soil samples from the investigated area revealed normal background concentrations with no anomalies

  9. Method of cement-solidification of radioactive liquid wastes containing surfactant

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Y; Yusa, H

    1979-04-10

    Purpose: To provide the subject method comprising the steps of adjusting the concentration of the surfactant to a value less than the predetermined value even when the concentration of the surfactant is high, and rendering the uniaxial compression strength of the cement-solidification body into more than the defined fabrication reference value. Method: To radioactive liquid wastes there are applied means for boiling and heating liquid wastes by addition of sulfuric acid, means for cracking surfactants by the addition of oxidants and means for precipitating and arresting surfactants. After suppressing the hindrance of the cement hydration reaction by surfactants, the radioactive liquid wastes are cement-solidified. (Nakamura, S.).

  10. ORNL radioactive waste operations

    International Nuclear Information System (INIS)

    Sease, J.D.; King, E.M.; Coobs, J.H.; Row, T.H.

    1982-01-01

    Since its beginning in 1943, ORNL has generated large amounts of solid, liquid, and gaseous radioactive waste material as a by-product of the basic research and development work carried out at the laboratory. The waste system at ORNL has been continually modified and updated to keep pace with the changing release requirements for radioactive wastes. Major upgrading projects are currently in progress. The operating record of ORNL waste operation has been excellent over many years. Recent surveillance of radioactivity in the Oak Ridge environs indicates that atmospheric concentrations of radioactivity were not significantly different from other areas in East Tennesseee. Concentrations of radioactivity in the Clinch River and in fish collected from the river were less than 4% of the permissible concentration and intake guides for individuals in the offsite environment. While some radioactivity was released to the environment from plant operations, the concentrations in all of the media sampled were well below established standards

  11. Modelling of motorway tunnels scenario for utilization of conditionally released radioactive materials

    International Nuclear Information System (INIS)

    Hrncir, Tomas; Panik, Michal; Necas, Vladimir

    2011-01-01

    Considerable amount of solid radioactive waste with radioactivity slightly above the limits for unconditional release is generated during the decommissioning of the nuclear installations. Conditional release deals with precisely this type of materials with activity slightly above limits in order to save considerable financial resources, which would be otherwise spend on treatment, conditioning and disposal of these materials at appropriate repository. The basic principles of conditional release as well as possibilities of reusing of the conditionally released materials are described. One of these possibilities of the reusing was chosen and application proposal of conditional release of metal waste - steel reinforcement in the concrete, which could be used for construction of motorway tunnels, was created. The computer code Visiplan 4.0 3D ALARA planning tool software was used for the calculation of effective individual dose for personnel constructing the tunnel and for critical group related to scenario. Particular models for individual scenarios of conditional release have been developed within the scope of this software code. The aim of the paper is to determine a level of the radioactivity of conditional released materials to avoid over exceeding the value of annual individual effective dose 10μSv/year established by international recommendations. (author)

  12. Technical report on natural evaporation system for radioactive liquid waste treatment arising from TRIGA research reactors' decontamination and decommissioning activities

    International Nuclear Information System (INIS)

    Moon, J. S.; Jung, K. J.; Baek, S. T.; Jung, U. S.; Park, S. K.; Jung, K. H.

    1999-01-01

    This technical report described that radioactive liquid waste treatment for dismantling/decontamination of TRIGA Mark research reactor in Seoul. That is, we try safety treatment of operation radioactive liquid waste during of operating TRIGA Mark research reactor and dismantling radioactive liquid waste during R and D of research reactor hereafter, and by utilizing of new natural evaporation facility with describing design criteria of new natural evaporation facility. Therefore, this technical report described the quantity of present radioactive liquid waste and dismantling radioactive liquid waste hereafter, analysis the status of radial-rays/radioactivity, and also treatment method of this radioactive liquid waste. Also, we derived the method that the safeguard of outskirts environment and the cost down of radioactive liquid waste treatment by minimize of the radioactive liquid waste quantities, through-out design/operation of new natural evaporation facility for treatment of operation radioactive liquid waste and dismantling radioactive liquid waste. (author). 6 refs., 12 tabs., 5 figs

  13. Massive radioactive releases have a great impact on the accident costs

    International Nuclear Information System (INIS)

    Pascucci-Cahen, L.; Momal, P.

    2013-01-01

    This article investigates the costs of the consequences of a nuclear accident. The importance of the costs is very dependant on the amount of radioactivity released in the environment during the accident. 2 severe accidents are considered, each accident involves the fusion of the core but the first is characterized by a limited amount of radioactivity released in the atmosphere while the second involves massive radioactive releases. The list of consequences is as comprehensive as possible: site decontamination and dismantlement, land decontamination, sanitary impacts, population displacement, agricultural and economical losses, impact on tourism, impact on the production of electricity...In the first case the total cost reaches 120 billion euros which is still manageable at the scale of a country whereas in the second case the bill reaches 430 billion euros which is unbearable for a country. The very slight probability of such events does not compensate for their catastrophic potentials. (A.C.)

  14. Removal of some ions from the radioactive liquid wastes by means of membrane techniques

    International Nuclear Information System (INIS)

    Roman, Gabriela; Garganciuc, Dana; Batrinescu, Gheorghe; Popescu, Georgeta

    2000-01-01

    The radioactive wastes imply important problems in the pollution control. Contrary to the case of other liquid wastes, which are specifically treated depending on the nature of pollutants, the liquid radioactive wastes are treated as a function of their activity (high, medium or low) and not depending on the nature of radioisotopes. The paper presents the advantages of the membrane processes as comparing with the classical processes in the removal of some ions from liquid radioactive waste up to values admissible of the current standards. Two types of radioactive liquid solutions were processed namely: one solution from the decontamination of the parts of an installation and other from the decontamination of primary circuit of the nuclear power plant. The first solution was treated with ultrafiltration and reverse osmosis, the retention for radioactive and toxic elements ranging between 14 - 69% for ultrafiltration and 63 - 99% for reverse osmosis. The second solution was processed only with reverse osmosis, a retention between 64 - 98% being obtained. The tests proved that by reverse osmosis membrane process a good removal efficiency of radioactive elements from liquid waste is obtained, corresponding to the requirements imposed by the current regulations. (author)

  15. Packaging and transportation of radioactive liquid at the U.S. Department of Energy Hanford Site

    International Nuclear Information System (INIS)

    Smith, R.J.

    1995-02-01

    Beginning in the 1940's, radioactive liquid waste has been generated at the US Department of Energy (DOE) Hanford Site as a result of defense material production. The liquid waste is currently stored in 177 underground storage tanks. As part of the tank remediation efforts, Type B quantity packagings for the transport of large volumes of radioactive liquids are required. There are very few Type B liquid packagings in existence because of the rarity of large-volume radioactive liquid payloads in the commercial nuclear industry. Development of aboveground transport systems for large volumes of radioactive liquids involves institutional, economic, and technical issues. Although liquid shipments have taken place under DOE-approved controlled conditions within the boundaries of the Hanford Site for many years, offsite shipment requires compliance with DOE, US Nuclear Regulatory Commission (NRC), and US Department of Transportation (DOT) directives and regulations. At the present time, no domestic DOE nor NRC-certified Type B packagings with the appropriate level of shielding are available for DOT-compliant transport of radioactive liquids in bulk volumes. This paper will provide technical details regarding current methods used to transport such liquids on and off the Hanford Site, and will provide a status of packaging development programs for future liquid shipments

  16. Treatment of radioactive liquid organic waste using bacteria community

    International Nuclear Information System (INIS)

    Rafael Vicente de Padua Ferreira; Solange Kazumi Sakata; Maria Helena Bellini; Julio Takehiro Marumo; Fernando Dutra; Patricia Busko Di Vitta; Maria Helena Tirollo Taddei

    2012-01-01

    Waste management plays an important role in radioactive waste volume reduction as well as lowering disposal costs and minimizing the environment-detrimental impact. The employment of biomass in the removal of heavy metals and radioisotopes has a significant potential in liquid waste treatment. The aim of this study is to evaluate the radioactive waste treatment by using three different bacterial communities (BL, BS, and SS) isolated from impacted areas, removing radioisotopes and organic compounds. The best results were obtained in the BS and BL community, isolated from the soil and a lake of a uranium mine, respectively. BS community was able to remove 92% of the uranium and degraded 80% of tributyl phosphate and 70% of the ethyl acetate in 20 days of experiments. BL community removed 81% of the uranium and degraded nearly 60% of the TBP and 70% of the ethyl acetate. SS community collected from the sediment of Sao Sebastiao channel removed 76% of the uranium and 80% of the TBP and 70% of the ethyl acetate. Both americium and cesium were removed by all communities. In addition, the BS community showed to be more resistant to radioactive liquid waste than the other communities. These results indicated that the BS community is the most viable for the treatment of large volumes of radioactive liquid organic waste. (author)

  17. Method for solidifying liquid radioactive wastes

    International Nuclear Information System (INIS)

    Berreth, J.R.

    1976-01-01

    The quantity of nitrous oxides produced during the solidification of liquid radioactive wastes containing nitrates and nitrites can be substantially reduced by the addition to the wastes of a stoichiometric amount of urea which, upon heating, destroys the nitrates and nitrites, liberating nontoxic N 2 , CO 2 and NH 3 . 5 claims, no drawings

  18. Atmospheric dispersion of radioactive releases: Computer code DIASPORA

    International Nuclear Information System (INIS)

    Synodinou, B.M.; Bartzis, J.M.

    1982-05-01

    The computer code DIASPORA is presented. Air and ground concentrations of an airborne radioactive material released from an elevated continuous point source are calculated using Gaussian plume models. Dry and wet deposition as well as plume rise effects are taken into consideration. (author)

  19. Models for environmental impact assessments of releases of radioactive substances from CERN facilities

    CERN Document Server

    Vojtyla, P

    2005-01-01

    The document describes generic models for environmental impact assessments of releases of radioactive substances from CERN facilities. Except for few models developed in the Safety Commission, the models are based on the 1997 Swiss directive HSK-R-41 and on the 2001 IAEA Safety Report No. 19. The writing style is descriptive, facilitating the practical implementation of the models at CERN. There are four scenarios assumed for airborne releases: (1) short-term releases for release limit calculations, (2) actual short-term releases, (3) short-term releases during incidents/accidents, and (4) chronic long-term releases during the normal operation of a facility. For water releases, two scenarios are considered: (1) a release into a river, and (2) a release into a water treatment plant. The document shall be understood as a reference for specific environmental studies involving radioactive releases and as a recommendation of the Safety Commission.

  20. Method of processing radioactive liquid wastes by using zeolites

    Energy Technology Data Exchange (ETDEWEB)

    Kanno, T; Mimura, H

    1975-09-18

    The object is to processing radioactive liquid waste by zeolites to be fixed to a solidified body having a very small lixiviation property. The nuclide in radioactive liquid waste is exchanged and adsorbed into natural or synthetic zeolites, which are then burnt to a temperature lower than 1000/sup 0/C -- melting point. Thus, the zeolite structure is broken to form fine amorphous silicate aluminate or silicate aluminate of the nuclide exchanged and adsorbed. Both are very hard to be soluble in water. Further, the lixiviation from the solidified body is limited to the surface thereof, and it will no longer be detected in a few days.

  1. Technical report on treatment of radioactive slurry liquid waste

    International Nuclear Information System (INIS)

    Jeong, Gyeong Hwan; Jo, Eun Sung; Park, Seung Kook; Jung, Ki Jung

    1999-06-01

    By literature survey, this report deals with the technology on typical pre-treatment and filtration of radioactive slurry liquid waste, produced during the operation of TRIGA Mark-II, III research reactor, and produced during the decommission/decontamination of TRIGA Mark-II, III research reactor. It is reviewed pre-treatment procedure, both physical and chemical that optimise the dewatering characteristics, and also surveyed types of dewatering devices based on centrifuges, vacuum and pressure filters with particular reference to various combined field approaches using two or more complementary driving forces to achieve better performance. Dewatering operations and devises on filtration of radioactive slurry liquid waste are also analysed. (author)

  2. Study on cementation of simulated radioactive borated liquid wastes

    International Nuclear Information System (INIS)

    Sun Qina; Li Junfeng; Wang Jianlong

    2010-01-01

    To compare sulfoaluminate cement with ordinary Portland cement on their cementation of radioactive borated liquid waste and to provide more data for formula optimization, simulated radioactive borated liquid waste were solidified by the two cements. 28 d compressive strength and strength losses after water/freezing/irradiation resistance tests were investigated. Leaching test and X-ray diffraction analysis were also conducted. The results show that it is feasible to solidify borated liquid wastes with sulfoaluminate cement and ordinary Portland cement with formulas used in the study. The 28 d compressive strengths, strength losses after tests and simulated nuclides leaching rates of the solidified waste forms meet the demand of GB 14569.1-93. The sulfoaluminate cement formula show better retention of Cs + than ordinary Portland cement formula. Boron, in form of B (OH) 4 - , incorporate in ettringite as solid solutions. (authors)

  3. Method for storage of liquid radioactive waste

    International Nuclear Information System (INIS)

    Hesky, H.; Wunderer, A.

    1978-01-01

    When nuclear fuel is reprocessed, apart from liquid radioactive wastes in certain cases also oxyhydrogen, i.e. a mixture of oxygen and hydrogen, is formed by radiolysis. It is proposed to remove the decay heat that will be formed by means of boiling cooling, to condense the steam and to recycle the condensate to the liquid waste store. The oxyhydrogen is to be rarefied by means of the steam and then catalytically recombined. The most advantageous process steps are discussed. (RW) [de

  4. VUJE experience with cementation of liquid and wet radioactive waste

    International Nuclear Information System (INIS)

    Kravarik, Kamil; Holicka, Zuzana; Pekar, Anton; Zatkulak, Milan

    2011-01-01

    Liquid and wet LLW generated during operation as well as decommissioning of NPPs is treated with different methods and fixed in a suitable fixation matrix so that a final product meets required criteria for its disposal in a final repository. Cementation is an important process used for fixation of liquid and wet radioactive waste such as concentrate, spent resins and sludge. Active cement grout is also used for fixation of low level solid radioactive waste loaded in final packing containers. VUJE Inc. has been engaged in research of cementation for long. The laboratory for analyzing radioactive waste properties, prescription of cementation formulation and estimation of final cement product properties has been established. Experimental, semi-production cementation plant has been built to optimize operation parameters of cementation. VUJE experience with cementation of liquid and wet LLW is described in the presented paper. VUJE has assisted in commissioning of Jaslovske Bohunice Treatment Centre. Cement formulations for treatment of concentrate, spent resins and sludge have been developed. Research studies on the stability of a final concrete packaging container for disposal in repository have been performed. Gained experience has been further utilized for design and manufacture of several cementation plants for treatment of various liquid and wet LLW. Their main technological and technical parameters as well as characterization of treated waste are described in the paper. Applications include the Mochovce Final Treatment Centre, Movable Cementation Facility utilizing in-drum mixing for treatment of sludge, Cementation Facility for treatment of tritiated water in Latvia and Cementation Facility for fixation of liquid and solid institutional radioactive waste in Bulgaria, which utilizes lost stirrer mixer. (author)

  5. Consequences of severe radioactive releases to Nordic Marine environment

    Energy Technology Data Exchange (ETDEWEB)

    Iosjpe, M. [Norwegian Radiation Protection Authority (NRPA) (Norway); Isaksson, M. [Univ. of Gothenburg (Sweden); Joensen, H.P. [Froskaparsetur Foeroya. Faroe Islands, Torshavn (Denmark); Lahtinen, J. [Radiation and Nuclear Safety Authority (Finland); Logemann, K. [Univ. of Iceland (Iceland); Palsson, S.E. [Geislavarnir Rikisins (Iceland); Roos, P. [Technical Univ. of Denmark. Risoe DTU, Roskilde (Denmark); Suolanen, V. [Technical Research Centre of Finland (Finland)

    2013-02-15

    In the report, consequences of hypothetical severe nuclear accidents releases to Nordic marine environment are preliminary considered. The considered marine area comprises the Baltic Sea (Sweden, Denmark, Finland) and the North Atlantic (Iceland, Faroes, Norway) areas. The hypothetical severe nuclear accidents can be related to nuclear power plants, nuclear powered submarines or ice-breakers. Quite comprehensive survey on radioactive source terms of extremely severe nuclear power and submarine accidents has been done. This enables to estimate more realistically possible radioactive releases of various elements and nuclides to marine environment. One recent reference is of course the Fukushima accident and estimated releases there. The marine flows and dilution circumstances around the Nordic nuclear power plants and in the Baltic Sea area in general, has been studied. Respectively marine flows related to Iceland and Faroes coasts are considered with measured data and with preliminary 3D-model simulations. The substantial depth of sea water in the North Atlantic affect vertical concentration profiles to some extent. At Icelandic or Faroese waters, a potential submarine accident would likely occur in a well defined water mass, and radioactivity from the accident would be detected and spread with the flow regime of the water mass in the world ocean. Based on hypothetical severe accidents scenarios, preliminary consequence calculations has been done. It should be emphasised that the considered severe accident cases, considered in this study, do not directly attach any specific Nordic nuclear power plant or any specific submarine type. The considered radioactive releases will, however, provide specified references for more extensive consideration of environmental consequences of severe - or minor - radioactive releases to Nordic marine environment. As a reference, the release amounts from a 3000 MW{sub th} reactor size were used. Based on source term analyses, the

  6. Consequences of severe radioactive releases to Nordic Marine environment

    International Nuclear Information System (INIS)

    Iosjpe, M.; Isaksson, M.; Joensen, H.P.; Lahtinen, J.; Logemann, K.; Palsson, S.E.; Roos, P.; Suolanen, V.

    2013-02-01

    In the report, consequences of hypothetical severe nuclear accidents releases to Nordic marine environment are preliminary considered. The considered marine area comprises the Baltic Sea (Sweden, Denmark, Finland) and the North Atlantic (Iceland, Faroes, Norway) areas. The hypothetical severe nuclear accidents can be related to nuclear power plants, nuclear powered submarines or ice-breakers. Quite comprehensive survey on radioactive source terms of extremely severe nuclear power and submarine accidents has been done. This enables to estimate more realistically possible radioactive releases of various elements and nuclides to marine environment. One recent reference is of course the Fukushima accident and estimated releases there. The marine flows and dilution circumstances around the Nordic nuclear power plants and in the Baltic Sea area in general, has been studied. Respectively marine flows related to Iceland and Faroes coasts are considered with measured data and with preliminary 3D-model simulations. The substantial depth of sea water in the North Atlantic affect vertical concentration profiles to some extent. At Icelandic or Faroese waters, a potential submarine accident would likely occur in a well defined water mass, and radioactivity from the accident would be detected and spread with the flow regime of the water mass in the world ocean. Based on hypothetical severe accidents scenarios, preliminary consequence calculations has been done. It should be emphasised that the considered severe accident cases, considered in this study, do not directly attach any specific Nordic nuclear power plant or any specific submarine type. The considered radioactive releases will, however, provide specified references for more extensive consideration of environmental consequences of severe - or minor - radioactive releases to Nordic marine environment. As a reference, the release amounts from a 3000 MW th reactor size were used. Based on source term analyses, the chosen

  7. Bituminization of liquid radioactive waste. Part 3

    International Nuclear Information System (INIS)

    G'oshev, G.S.; Gradev, G.D.; Stefanova, I.G.; Milusheva, A.G.; Guteva, E.S.; Stefanov, G.I.

    1991-01-01

    The elaborated technology for bituminization of liquid radioactive wastes (salt concentrates) is characterized by the fact that the bituminization process takes place in two stages: concentration of the liquid residue and evaporation of the water with simultaneous homogeneous incorporation of the salts in the melted bitumen. An experimental installation for bituminization of salt concentrates was designed on the basis of this technology. The experience accumulated during the design and construction of the installation for bituminization of salt concentrates could be used for designing and constructing an industrial installation for bituminization of the liquid residue of the nuclear power plants. 2 tabs., 3 figs., 3 refs

  8. Simulated atmospheric disperison of radioactive material released in an urban area

    International Nuclear Information System (INIS)

    Akins, R.E.; Church, H.W.; Tierney, M.S.

    1977-01-01

    A combination of Gaussian plume and particle-in-cell techniques is used to simulate the atmospheric transport and dispersion of a puff release of radioactive material. The release is caused by an accident that is assumed to occur during the shipment of the radioactive material through central New York City. The simulation provides estimates of volumetric and surface concentrations of the dispersed material that are used to predict radiation doses incurred by the City's population in the event of an accidental release. In the simulation, the release point is arbitrary and the material is assumed to be either a gas or fine particles. The Gaussian plume model follows cloud concentrations from the release time until times when transport over distances up to 500 m has been achieved. The released cloud may stabilize at street level or above the mean buildings height; at a street intersection or in the middle of the block. The possibility of the formation of multiple clouds, owing to circumstances of wind flow direction and street geometry, is allowed

  9. Atmospheric dispersion models of radioactivity releases

    International Nuclear Information System (INIS)

    Oza, R.B.

    2016-01-01

    In view of the rapid industrialization in recent time, atmospheric dispersion models have become indispensible 'tools' to ensure that the effects of releases are well within the acceptable limits set by the regulatory authority. In the case of radioactive releases from the nuclear facility, though negligible in quantity and many a times not even measurable, it is required to demonstrate the compliance of these releases to the regulatory limits set by the regulatory authority by carrying out radiological impact assessment. During routine operations of nuclear facility, the releases are so low that environmental impact is usually assessed with the help of atmospheric dispersion models as it is difficult to distinguish negligible contribution of nuclear facility to relatively high natural background radiation. The accidental releases from nuclear facility, though with negligible probability of occurrence, cannot be ruled out. In such cases, the atmospheric dispersion models are of great help to emergency planners for deciding the intervention actions to minimize the consequences in public domain and also to workout strategies for the management of situation. In case of accidental conditions, the atmospheric dispersion models are also utilized for the estimation of probable quantities of radionuclides which might have got released to the atmosphere. Thus, atmospheric dispersion models are an essential tool for nuclear facility during routine operation as well as in the case of accidental conditions

  10. Release mechanisms from shallow engineered trenches used as repositories for radioactive wastes

    International Nuclear Information System (INIS)

    Locke, J.; Wood, E.

    1987-05-01

    This report has been written for the Department of the Environment as part of their radioactive waste management research programme. The aim has been to identify release mechanisms of radioactivity from fully engineered trenches of the LAND 2 type and, to identify the data needed for their assessment. No direct experimental work has been involved. The report starts with a brief background to UK strategy and outlines a basic disposal system. It gives reviews of existing experience of low level radioactive waste disposal from LAND 1 trenches and of UK experience of toxic waste disposal to provide a practical basis for the next section which covers the implications of identified release mechanisms on the design requirements for an engineered trench. From these design requirements and their interaction with potential site conditions (both saturated and unsaturated zone sites are considered) an assessment of radionuclide release mechanism is made. (author)

  11. Savannah River Site Ingestion Pathway Methodology Manual for Airborne Radioactive Releases

    Energy Technology Data Exchange (ETDEWEB)

    Vincent, A.W. III

    2001-01-03

    This manual documents a recommended methodology for determining the ingestion pathway consequences of hypothetical accidental airborne radiological releases from facilities at the Savannah River Site. Both particulate and tritiated radioactive contaminants are addressed. Other approaches should be applied for evaluation of routine releases.

  12. Risk assessment methodology for evaluating releases of radioactively contaminated materials

    International Nuclear Information System (INIS)

    Chen, S.Y.

    1993-01-01

    Extensive decontamination and decommissioning (D ampersand D) activities are expected to be required in the near future in association with license termination of nuclear power facilities and cleanup efforts at the U.S. Department of Energy's (DOE's) weapons production facilities. In advance of these D ampersand D activities, it is becoming increasingly urgent that standards be established for the release of materials with residual radioactive contamination. The only standards for unrestricted release that currently exist address surface contamination. The methods used to justify those standards were developed some 20 yr ago and may not satisfy today's criteria. Furthermore, the basis of setting standards has moved away from the traditional open-quotes instrumentation-basedclose quotes concept toward a open-quotes risk-basedclose quotes approach. Therefore, as new release standards are developed, it will be necessary that risk assessment methodology consistent with modern concepts be incorporated into the process. This paper discusses recent developments in risk methodology and issues and concerns regarding the future development of standards for the release of radioactively contaminated materials

  13. Deep injection disposal of liquid radioactive waste in Russia

    International Nuclear Information System (INIS)

    Foley, M.G.; Ballou, L.; Rybal'chenko, A.I.; Pimenov, M.K.; Kostin, P.P.

    1998-01-01

    Originally published in Russian, Deep Injection Disposal is the most comprehensive account available in the West of the Soviet and Russian practice of disposing of radioactive wastes into deep geological formations. It tells the story of the first 40 years of work in the former Soviet Union to devise, test, and execute a program to dispose by deep injection millions of cubic meters of liquid radioactive wastes from nuclear materials processing. The book explains decisions involving safety aspects, research results, and practical experience gained during the creation and operation of disposal systems. Deep Injection Disposal will be useful for studying other problems worldwide involving the economic use of space beneath the earth's surface. The material in the book is presented with an eye toward other possible applications. Because liquid radioactive wastes are so toxic and the decisions made are so vital, information in this book will be of great interest to those involved in the disposal of nonradioactive waste

  14. Analysis and study on generic models for use in assessing the impact of radioactive liquid effluent to the environment

    International Nuclear Information System (INIS)

    Ma Wenlin; Cao Jianzu; Fang Dong

    2008-01-01

    The assessment of the impact of discharges of radioactive substances into surface water under normal condition of nuclear facilities is an important part of the environmental impact analysis. Generic methods for assessing the impact of radioactive liquid effluent release into surface water provided by IAEA Safety Reports Series 19 are studied in this paper, and also an example calculation that assesses the impact of radioactive surface water discharge of HTR-PM ( High Temperature Air-cooled Reactor demonstration unit) in Anhui is presented in this paper to illustrate that a simplified but conservative assessment can be used for the purpose of screening proposed radioactive discharges. If the results meet the relevant requirements specified by the relevant regulatory authority, the further calculations are not needed. If they fails to meet the requirements, the more field data are to be sampled for calculations by more sophisticated mode or otherwise. (authors)

  15. Biodegradation of radioactive organic liquid waste from spent fuel reprocessing

    International Nuclear Information System (INIS)

    Ferreira, Rafael Vicente de Padua

    2008-01-01

    The research and development program in reprocessing of low burn-up spent fuel elements began in Brazil in 70's, originating the lab-scale hot cell, known as Celeste located at Nuclear and Energy Research Institute, IPEN - CNEN/SP. The program was ended at the beginning of 90's, and the laboratory was closed down. Part of the radioactive waste generated mainly from the analytical laboratories is stored waiting for treatment at the Waste Management Laboratory, and it is constituted by mixture of aqueous and organic phases. The most widely used technique for the treatment of radioactive liquid wastes is the solidification in cement matrix, due to the low processing costs and compatibility with a wide variety of wastes. However, organics are generally incompatible with cement, interfering with the hydration and setting processes, and requiring pre -treatment with special additives to stabilize or destroy them. The objective of this work can be divided in three parts: organic compounds characterization in the radioactive liquid waste; the occurrence of bacterial consortia from Pocos de Caldas uranium mine soil and Sao Sebastiao estuary sediments that are able to degrade organic compounds; and the development of a methodology to biodegrade organic compounds from the radioactive liquid waste aiming the cementation. From the characterization analysis, TBP and ethyl acetate were chosen to be degraded. The results showed that selected bacterial consortia were efficient for the organic liquid wastes degradation. At the end of the experiments the biodegradation level were 66% for ethyl acetate and 70% for the TBP. (author)

  16. Influence of plume rise on the consequences of radioactive material releases

    International Nuclear Information System (INIS)

    Russo, A.J.; Wayland, J.R.; Ritchie, L.T.

    1977-01-01

    Estimates of health consequences resulting from a postulated nuclear reactor accident can be strongly dependent on the buoyant rise of the plume of released radioactive material. The sensitivity of the consequences of a postulated accident to two different plume rise models has been investigated. The results of these investigations are the subject of this report. One of the models includes the effects of emission angle, momentum, and radioactive heating of the released material. The difference in the consequence estimates from the two models can exceed an order of magnitude under some conditions, but in general the results are similar

  17. Experiment of decontamination of radioactive liquid by a biological method

    International Nuclear Information System (INIS)

    Wormser, G.

    1962-01-01

    The author reports experiments of treatment of radioactive liquid effluents by percolation on a bacterial bed like the one used for the treatment of sewer wastewaters. He also reports results obtained in other countries in terms of reduction of effluent radioactivity for various radioactive ions. The installation is described and results are presented in terms of variation of contamination of an effluent with respect to its recycling on a bacterial bed [fr

  18. Treatment of low-level liquid radioactive wastes by electrodialysis

    International Nuclear Information System (INIS)

    DelDebbio, J.A.; Donovan, R.I.

    1986-01-01

    This paper presents the results of pilot plant studies on the use of electrodialysis (ED) for the removal of radioactive and chemical contaminants from acidic low-level radioactive wastes resulting from nuclear fuel reprocessing operations. Decontamination efficiencies are reported for strontium-90, cesium-137, iodine-129, ruthenium-106 and mercury. Data for contaminant adsorption on ED membranes and liquid waste volumes generated are also presented

  19. Method and apparatus for the purification of a liquid contaminated with radioactive substances

    International Nuclear Information System (INIS)

    Mende, H.

    1976-01-01

    A method of and apparatus for the purification of a liquid contaminated with radioactive substances is described, wherein the liquid is infed to an evaporator in or with which there is connected a column having a multiplicity of superposed plates or floors. The vapor generated in the evaporator is guided through a washing or scrubbing liquid uniformly distributed at the floors and flowing in crosswise counterflow with regard to the vapor. The washing liquid at the floors is deflected a number of times in such a manner that the washing liquid itself together with the droplets entrained by the vapor are uniformly admixed and the washing liquid subjected to a constant intake of the radioactive substance

  20. Method of processing radioactive liquid wastes

    International Nuclear Information System (INIS)

    Kawamura, Fumio; Funabashi, Kiyomi; Matsuda, Masami.

    1984-01-01

    Purpose: To improve the performance of removing metal ions in ion exchange resins for use in clean-up of service water or waste water in BWR type reactors. Method: A column filled with activated carbon is disposed at the pre- or post-stage of a clean-up system using ion exchange resins disposed for the clean-up of service water or waste water of a nuclear reactor so that organics contained in water may be removed through adsorption. Since the organic materials are thus adsorbed and eliminated, various types of radioactive ions contained in radioactive liquid are no more masked and the performance of removing ions in the ion exchanger resins of the clean-up device can be improved. (Moriyama, K.)

  1. PABLM, Doses from Radioactive Releases to Atmosphere and Food Chain

    International Nuclear Information System (INIS)

    Napier, B.A.; Kennedy, W.E. Jr.; Soldat, J.K.

    1983-01-01

    1 - Description of problem or function: PABLM calculates internal radiation doses to man from radionuclides in food products and external radiation doses from radionuclides in the environment. Radiation doses from radionuclides in the environment may be calculated from deposition on the soil or plants during an atmospheric or liquid release, or from exposure to residual radionuclides after the releases have ended. Radioactive decay is considered during the release, after deposition, and during holdup of food after harvest. The radiation dose models consider exposure to radionuclides deposited on the ground or crops from contaminated air or irrigation water, radionuclides in contaminated drinking water, aquatic foods raised in contaminated water, and radionuclides in bodies of water and sediments where people might fish, boat, or swim. For vegetation, the radiation dose model considers both direct deposition and uptake through roots. Doses may be calculated for either a maximum-exposed individual or for a population group. The program is designed to calculate accumulated radiation doses from the chronic ingestion of food products that contain radionuclides and doses from the external exposure to radionuclides in the environment. A first-year committed dose is calculated as well as an integrated dose for a selected number of years. 2 - Method of solution: A chain decay scheme including branching for transitions to and from isomeric states is used for radioactive decay. The equations for calculating internal radiation doses are derived from those given by the International Commission on Radio- logical Protection (ICRP) for body burdens and the maximum possible concentration (MPC) for each radionuclide. These doses are calculated as a function of radionuclide concentration in food products, ingestion rates, and a radionuclide-specific dose- commitment factor. Radiation doses from external exposure to contaminated water and soil are calculated using the basic assumption

  2. Process of liquid radioactive waste treatment in nuclear power plant and development trend

    International Nuclear Information System (INIS)

    Liu Jiean; Wang Xin; Liu Dan; Zhu Laiye; Chen Bin

    2014-01-01

    The popular liquid radioactive waste treatment methods in nuclear power plants (NPP) are Chemical precipitation, evaporation, ion exchange, membrane treatment, chemical coagulation and activated carbon absorption and so on. 'Filter + activated carbon absorption (Chemical coagulation) + ion exchange' has a good prospect for development, as its simple process, high decontamination factor, low energy consumption and smaller secondary wastes. Also the process is used in Sanmen and Haiyang Projects. The severe incident in NPP set an even higher demand on liquid radioactive waste treatment. The new type treatment materials, optimization of the existed treatment, combination of treatment and the mobile treatment facility is the development trend in liquid radioactive waste treatment in NPP. (authors)

  3. The containment and an absorbent evaluation for a package for a liquid radioactive isotope

    Energy Technology Data Exchange (ETDEWEB)

    Bang, K. S.; Lee, J. C.; Kim, D. H.; Hwang, C. S.; Kim, H. J.; Seo, K. S

    2005-03-01

    Radioactive isotopes must be safely transported from the production centre to the point of use. The shipping package to safely transport radioactive isotopes should be able to withstand the conditions prescribed by law. A type a package, which is used to transport liquid radioactive materials, should have a containment system comprising a primary inner and a secondary outer containment or it should be provided with a sufficiently absorbent material to absorb twice the volume of the liquid contents. Accordingly, an absorbent material for use in a Type A package to transport a liquid radioactive isotope was estimated. To estimate the integrity of containment, the leakage tests for a containment system for a Type A package for domestic and abroad expert were conducted.

  4. Palisades Nuclear Plant. Radioactive effluents and environmental monitoring sections to second annual operating report

    International Nuclear Information System (INIS)

    1977-01-01

    A total of 0.435 Ci of radioactive liquid effluent less tritium was released with 19.63 Ci of tritium. Both liquid and gaseous releases were within permissible limits. There were 8 Ci of solid wastes stored on the site as of 12/31/76. Data clearly shows there was no detectable increase in radioactivity levels in the environmental media that can be attributed to plant effluents. Monitoring reports are presented concerning fish, meteorology, noise, and cooling tower drift

  5. ARANO - a computer program for the assessment of radiological consequences of atmospheric radioactive releases

    International Nuclear Information System (INIS)

    Savolainen, I.; Vuori, S.

    1980-09-01

    A short description of the calculation possibilities, methods and of the structure of the computer code system ARANO is given, in addition to the input quide. The code can be employed in the calculation of environmental radiological consequences caused by radioactive materials released to atmosphere. Results can be individual doses for different organs at given distances from the release point, collective doses, numbers of persons exceeding given dose limits, numbers of casualties, areas polluted by deposited activity and losses of investments or production due to radioactive contamination. Both a case with a single release and atmospheric dispersion situation and a group of radioactive release and dispersions with discrete probability distributions can be considered. If the radioactive releases or the dispersion conditions are described by probability distributions, the program assesses the magnitudes of the specified effects in all combinations of the release and dispersion situations and then calculates the expectation values and the cumulative probability distributions of the effects. The vertical mixing in the atmosphere is described with a Ksub(Z)-model. In the lateral direction the plume is assumed to be Gaussian, and the release duration can be taken into account in the σsub(y)-values. External gamma dose from the release plume is calculated on the basis of a data file which has been created by 3-dimensional integration. Dose due to inhalation and due to gamma radiation from the contaminated ground are calculated by using appropriate dose conversion factors, which are collected into two mutually alternative block data subprograms. (author)

  6. Treatment of radioactive organics liquid wastes

    International Nuclear Information System (INIS)

    Morales Galarce, Tania

    1999-01-01

    Because of the danger that radioactive wastes can pose to society and to the environment a viable treatment alternative must be developed to prepare these wastes for final disposal. The waste studied in this work is a liquid organic waste contaminated with the radioisotope tritium. This must be treated and then changed into solid form in a 200 liter container. This study defined an optimum formulation that immobilizes the liquid waste. The organic waste is first submitted to an absorption treatment, with Celite absorbent, which had the best physical characteristics from the point of view of radioactive waste management. Then this was solidified by forming a cement mortar, using a highly resistant local cement, Polpaico 400. Various mixes were tested, with different water/cement, waste/absorbent and absorbed waste/cement ratios, until a mixture that met the quality control requirements was achieved. The optimum mixture obtained has a water/cement ratio of 0.35 (p/p) that is the amount of water needed to make the mixture workable, and minimum water for hydrating the cement; a waste/absorbent ration of 0.5 (v/v), where the organic liquid is totally absorbed, and is incorporated in the solid's crystalline network; and an absorbed waste/cement ratio of 0.8 (p/p), which represents the minimum amount of cement needed to obtain a solid product with the required mechanical resistance. The mixture's components join together with no problem, to produce a good workable mixture. It takes about 10 hours for the mixture to harden. After 14 days, the resulting solid product has a resistance to compression of 52 Kgf/cm2. The formulation contains 22.9% immobilized organic waste, 46.5% cement, 14.3% Celite and 16.3% water. Organic liquid waste can be treated and a solid product obtained, that meets the qualitative and quantitative parameters required for its disposal. (CW)

  7. Application of macrophytes as biosorbents for radioactive liquid waste treatment

    International Nuclear Information System (INIS)

    Vieira, Ludmila Cabreira

    2016-01-01

    Radioactive waste as any other type of waste should be treated and disposed adequately. It is necessary to consider its physical, chemical and radiological characteristics for choosing the appropriate action for the treatment and final disposal. Many treatment techniques currently used are economically costly, often invalidating its use and favoring the study of other treatment techniques. One of these techniques is biosorption, which demonstrates high potential when applied to radioactive waste. This technology uses materials of biological origin for removing metals. Among potential biosorbents found, macrophyte aquatics are useful because they may remove uranium present in the liquid radioactive waste at low cost. This study aims to evaluate the biosorption capacity of macrophyte aquatics Pistia stratiotes, Limnobium laevigatum, Lemna sp and Azolla sp in the treatment of liquid radioactive waste. This study was divided into two stages, the first one is characterization and preparation of biosorption and the other is tests, carried out with uranium solutions and real samples. The biomass was tested in its raw form and biosorption assays were performed in polypropylene vials containing 10 ml of solution of uranium or 10ml of radioactive waste and 0.20g of biomass. The behavior of biomass was evaluated by sorption kinetics and isotherm models. The highest sorption capacities found was 162.1 mg / g for the macrophyte Lemna sp and 161.8 mg / g for the Azolla sp. The equilibrium times obtained were 1 hour for Lemna sp, and 30 minutes for Azolla sp. With the real waste, the macrophyte Azolla sp presented a sorption capacity of 2.6 mg / g. These results suggest that Azolla sp has a larger capacity of biosorption, therefore it is more suitable for more detailed studies of treatment of liquid radioactive waste. (author)

  8. Chemical treatment of liquid radioactive waste at the Boris Kidric Institute

    International Nuclear Information System (INIS)

    Lazic, S.; Vukovic, Z.; Voko, A.

    1989-01-01

    The results of lab-scale experiments on the chemical treatment of radioactive liquid waste collected at the Boris Kidric Institute are presented. The radioactive waste was treated by cobalt hexacyanoferrate precipitation followed by flocculation with polyelectrolyte flocculating agents. The main parameters investigated were standing time, pH and ratio of reagents. The flocculating agents were tested by filtration test and floccule stability test. Satisfactory decontamination factors by precipitation at pH 10 and good separation of solid and liquid phase by applying Praestol polyelectrolytes were obtained (author)

  9. Hazard of radioactive releases resulted from coal burning

    International Nuclear Information System (INIS)

    Gabbard, V.

    1995-01-01

    Consideration is given to the data, pointing to the fact, that coal-burning power plants release of radioactive substances, contained in gaseous wastes, is not less, than the same one of nuclear power plants. The necessity of regulating emission of these substance in atmosphere by analogy with nuclear power industry is shown. 1 fig

  10. Determination of 90Sr in radioactive liquid waste-a comparative study using 'SrCO3 precipitation, extraction chromatography and Cerenkov radiation counting' techniques

    International Nuclear Information System (INIS)

    Mani, A.G.S.; Renganathan, K.; Rao, S.V.S.; Sinha, P.K.

    2007-01-01

    In the wastes generated at nuclear power plants, 90 Sr and 90 Y are responsible for approximately 7% and 38% of the total fission product activity after 1 year and 10 years respectively. Monitoring of the liquid waste produced in nuclear facilities for 90 Sr before and after releasing to the aquatic environment is essential, as it can cause a long term biological hazard due to its chemical similarity with Ca, high fission yield (5.9%) and long radioactive and biological half lives of 28 and 49.3 years respectively. As conventional carbonate precipitation method for 90 Sr- 90 Y estimation is laborious and time consuming, the other methods like extraction chromatography technique using strontium selective crown ether and Cerenkov radiation counting by liquid scintillation analyser (LSA) were also investigated. This paper describe procedures involved in the estimation of strontium in low-level radioactive liquid effluent by using above methods and deals with the comparison of results. Cerenkov radiation counting by LSA has an edge over the other methods in the determination of 90 Sr as it is a simple and rapid technique. Experiences gained in the estimation of 90 Sr in low-level radioactive liquid effluent using Cerenkov radiation counting technique is discussed in detail in this paper. (author)

  11. Real-time alpha monitoring of a radioactive liquid waste stream at Los Alamos National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.D.; Whitley, C.R.; Rawool-Sullivan, M. [Los Alamos National Lab., NM (United States)

    1995-12-31

    This poster display concerns the development, installation, and testing of a real-time radioactive liquid waste monitor at Los Alamos National Laboratory (LANL). The detector system was designed for the LANL Radioactive Liquid Waste Treatment Facility so that influent to the plant could be monitored in real time. By knowing the activity of the influent, plant operators can better monitor treatment, better segregate waste (potentially), and monitor the regulatory compliance of users of the LANL Radioactive Liquid Waste Collection System. The detector system uses long-range alpha detection technology, which is a nonintrusive method of characterization that determines alpha activity on the liquid surface by measuring the ionization of ambient air. Extensive testing has been performed to ensure long-term use with a minimal amount of maintenance. The final design was a simple cost-effective alpha monitor that could be modified for monitoring influent waste streams at various points in the LANL Radioactive Liquid Waste Collection System.

  12. Release process for non-real property containing residual radioactive material

    International Nuclear Information System (INIS)

    Ranek, N.L.; Chen, S.Y.; Kamboj, S.; Hensley, J.; Burns, D.; Fleming, R.; Warren, S.; Wallo, A.

    1997-01-01

    It is DOE's objective to operate its facilities and to conduct its activities so that radiation exposures to members of the public are maintained within acceptable limits and exposures to residual radioactive materials are controlled. To accomplish this, DOE has adopted Order DOE 5400.51 'Radiation Protection of the Public and the Environment', and will be promulgating IO CR Part 834 to codify and clarify the requirements of DOE 5400.5. Under both DOE 5400.5 and 10 CR Part 834, radioactively contaminated DOE property is prohibited from release unless specific actions have been completed prior to the release. This paper outlines a ten-step process that, if followed, will assist DOE Operations and contractor personnel in ensuring that the required actions established by Order DOE 5400.5 and 10 CR Part 834 have been appropriately completed prior to the release for reuse or recycle of non-real property (e.g., office furniture, computers, hand tools, machinery, vehicles and scrap metal). Following the process will assist in ensuring that radiological doses to the public from the released materials will meet applicable regulatory standards and be as low as reasonably achievable (ALARA)

  13. Summary of accidental releases of radioactivity detected off the Nevada Test Site, 1963--1986

    International Nuclear Information System (INIS)

    Patzer, R.G.; Phillips, W.G.; Grossman, R.F.; Black, S.C.; Costa, C.F.

    1988-08-01

    Of the more than 450 underground nuclear explosives tests conducted at the Nevada Test Site from August 1963 (signing of the Limited Test Ban Treaty) through the end of 1986, only 23 accidentally released radioactivity that was detectable beyond the boundary of the NTS. Of these 23, 4 were detectable off the NTS only by aircraft while the remainder were detectable by ground monitoring instruments. Since the Baneberry venting of December 1970, only two tests released radioactivity that was detectable off the NTS, and this was a seepage of radioactive noble gases. None of these releases from underground tests designed for complete containment caused exposure of the population living in the area that exceeded standards recommended by national and international radiation protection agencies. This report summarizes the releases from each of the tests, describes the monitoring that was conducted, and lists the location of the maximum exposure

  14. Radioactive waste handling at the Mochovce NPP, 1998-2008

    International Nuclear Information System (INIS)

    Vasickova, Gabriela

    2009-01-01

    The radioactive waste management system at the Mochovce NPP is described. The system addresses technical aspects as well as administrative provisions related to radioactive waste generated within the controlled area, from the waste generation phase to waste sorting, packaging, storage, recording, measurement, and transportation to the Bohunice waste processing facility or transfer to the Mochovce liquid radioactive waste treatment facility. The article also addresses conditions for release from the controlled area to the environment for radioactive waste which can be exempt from the institutional administrative control system or released to the environment on the basis of a valid permission issued by the relevant regulatory authority

  15. Environmental releases for calendar year 1993

    International Nuclear Information System (INIS)

    Thomas, S.P.; Curn, B.L.

    1994-07-01

    This report presents data on radioactive and nonradioactive materials released into the environment during calendar year 1993 from facilities managed by Westinghouse Hanford Company. As part of this executive summary, comprehensive data summaries of air emissions and liquid effluents in 1993 are displayed in tables. These tables represent the following: radionuclide air emissions data; data on radioactive liquid effluents discharged to the soil; radionuclides discharged to the Columbia River; nonradioactive air emissions data; total volumes and flow rates of 200/600 area liquid effluents. Both summary and detailed presentations of these data are given. When appropriate, comparisons to data from previous years are made

  16. Study of optimal transformation of liquid effluents resulting from the destruction of radioactive sodium by water into ultimate solid wastes

    International Nuclear Information System (INIS)

    Rodriguez, G.; Camaro, S.; Fiquet, O.; Bernard, A.; Le Bescop, P.

    1997-01-01

    In the framework of sodium waste processing, it has been proposed to retain only processes that treat the sodium using water, thus generating the same by-products: hydrogen and sodium hydroxide. As the objective is to minimise radioactive liquid releases and as, moreover, the authorizations with respect to sodium salt releases are highly restrictive, several solutions have been envisaged for transforming the active sodium hydroxide coming from sodium destruction processes into ultimate solid wastes that can be stored on the surface in a storage site approved by the ANDRA (National Radioactive Waste Management Agency): the Aube Storage Site (CSA). Two processes have been considered and compared: immobilisation in concrete (cementation) and immobilisation in ceramic (ceramisation). These two processes are evaluated according to several criteria: the state of advancement of the process, the quantity of sodium hydroxide (and therefore of sodium) that can be treated per package. (author)

  17. Device for treating plastic counting vials containing radioactive liquids

    International Nuclear Information System (INIS)

    Neidhart, B.; Brindoepke, H.W.; Flocke, W.; Kringe, K.P.; Lippmann, C.H.

    1985-01-01

    The treatment consists of separating the radioactive contents of the counting vial from its plastic components. The apparatus consists of a device for continuously supplying the counting vials to be treated, a means for crushing the vials into chips of plastic and a facility by means of which the radioactive contents of the counting vial and the separated plastic chips are collected separately from one another. A stirring assembly with a motor-driven stirrer and an alignment device are also provided. The radioactive substances pass through a sieve while the plastic chips slide down the sieve chute and into another container. All the metal parts of the facility are of stainless steel. The plastic chips collected in the sieve holder are washed and, after drying, are removed as negligibly radioactive solids. The weakly radioactive wash liquid is separated and collected. (orig./PW)

  18. Pyrohydrolytic separation technique for fluoride and chloride from radioactive liquid wastes

    International Nuclear Information System (INIS)

    Sawant, R.M.; Mahajan, M.A.; Shah, D.J.; Thakur, U.K.; Ramakumar, K.L.

    2011-01-01

    A rapid method for simultaneous determination of fluorine and chlorine in radioactive liquid wastes with ion chromatography after pyrohydrolysis separation was proposed for routine analysis. The elements were separated from radioactive liquid wastes by pyrohydrolysis and were subsequently determined with ion chromatography. Total time taken to determine these elements is about 45 min including 30 min for the pyrohydrolysis and 15 min for ion chromatography. The results of recovery tests ranged 95% or above. The limits of detection for F and Cl are 0.5 and 0.8 mg kg -1 , respectively. (author)

  19. Solvent for the simultaneous recovery of radionuclides from liquid radioactive wastes

    Science.gov (United States)

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

    2002-01-01

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

  20. Measurement of gross beta radioactivity in high-level liquid waste

    International Nuclear Information System (INIS)

    Lu Feng; Lin Cansheng; Zhang Xianzi; Chen Guoan; Zhang Chonghai

    1992-01-01

    Using beta plastic scintillation counter of low level background, gross beta radioactivity of twelve samples for high-level liquid waste is determined directly. Beta efficiency curves of plastic scintillation counter for four mass thickness are calibrated in advance. Determining gross beta radioactivity, gross efficiency of the scintillation counter for various energy beta ray is calculated via weighted mean method with the ratio of radioactivity for each nuclide. The ratio of radioactivity for nuclides which have gamma disintegration is determined in terms of the radioactivity measured by gamma spectrometer. The ratio of the radioactivity for 90 Sr which has purity beta disintegration is calculated in terms of half life time approximation. The ratio of the radioactivity for 147 Pm which also has purity disintegration is calculated by means of apparent cooling-time approximation. The uncertainty of results for the present work is about +-15%

  1. Air concentration and ground deposition following radioactive airborne releases

    International Nuclear Information System (INIS)

    Brofferio, C.; Cagnetti, P.; Ferrara, V.

    1985-01-01

    The fundamental aim of this report is to provide the mathematical and physical operational basis for the evaluation of air concentration and ground deposition, following radioactive airborne releases from a nuclear power plant, both during normal operations and in accidental conditions. As far as accidental releases are concerned, the basical assumptions on meteorological and diffusive situation are considered from a safety point of view: namely those pessimistic but realistically representative situation are taken into account which lead to maximum air concentration and ground deposition values, even if characterized by low recurrence probability. Those elements are the inputs for many environmental transfer models of maximum consequence evaluations up to man. As far as routine releases are concerned, it is shown, together with the usual models based on long term averaged meteorological conditions, also models studied to estimate atmospheric diffusion and deposition in low wind situations and in fog conditions, being those latter very frequent in the Po valley. Finally, the main operations and modalities of collecting and elaborating meteorological data for for radioprotection evaluations are also shown. It is to be pointed out that the methods and the models developed and considered in this work are of a more general validity, and can be also used for applications concerning non-radioactive releases, as it is the case when dealing with conventional power plants

  2. Proposed radioactive liquid effluent monitoring requirements at the Savannah River Site

    International Nuclear Information System (INIS)

    Jannik, G.T.; Carlton, W.H.; Blunt, B.C.

    1994-01-01

    Clear regulatory guidance exists for structuring a radiological air monitoring program, however, there is no parallel guidance for radiological liquid monitoring. For Department of Energy (DOE) facilities, there are no existing applicable federal regulations, DOE orders, or DOE guidance documents that specify at what levels continuous monitoring, continuous sampling, or periodic confirmatory measurements of radioactive liquid effluents must be made. In order to bridge this gap and to technically justify and document liquid effluent monitoring decisions at DOE's Savannah River Site, Westinghouse Savannah River Company has proposed that a graded, dose-based approach be established, in conjunction with limits on facility radionuclide inventories, to determine the monitoring and sampling criteria to be applied at each potential liquid radioactive effluent point. The graded approach would be similar to--and a conservative extension of--the existing, agreed-upon SRS/EPA-IV airborne effluent monitoring approach documented in WSRC's NESHAP Quality Assurance Project Plan. The limits on facility radionuclide inventories are based on--and are a conservative extension of--the 10 CFR 834, 10 CFR 20, and SCR 61-63 annual limits on discharges to sanitary sewers. Used in conjunction with each other, the recommended source category criteria levels and facility radionuclide inventories would allow for the best utilization of resources and provide consistent, technically justifiable determinations of radioactive liquid effluent monitoring requirements

  3. Characterization of radioactive organic liquid wastes

    International Nuclear Information System (INIS)

    Hernandez A, I.; Monroy G, F.; Quintero P, E.; Lopez A, E.; Duarte A, C.

    2014-10-01

    With the purpose of defining the treatment and more appropriate conditioning of radioactive organic liquid wastes, generated in medical establishments and research centers of the country (Mexico) and stored in drums of 208 L is necessary to characterize them. This work presents the physical-chemistry and radiological characterization of these wastes. The samples of 36 drums are presented, whose registrations report the presence of H-3, C-14 and S-35. The following physiochemical parameters of each sample were evaluated: ph, conductivity, density and viscosity; and analyzed by means of gamma spectrometry and liquid scintillation, in order to determine those contained radionuclides in the same wastes and their activities. Our results show the presence of H-3 (61%), C-14 (13%) and Na-22 (11%) and in some drums low concentrations of Co-60 (5.5%). In the case of the registered drums with S-35 (8.3%) does not exist presence of radioactive material, so they can be liberated without restriction as conventional chemical wastes. The present activities in these wastes vary among 5.6 and 2312.6 B g/g, their ph between 2 and 13, the conductivities between 0.005 and 15 m S, the densities among 1.05 and 1.14, and the viscosities between 1.1 and 39 MPa. (Author)

  4. Biosorption of uranium in radioactive liquid organic waste by coconut fiber

    International Nuclear Information System (INIS)

    Marumo, Julio Takehiro; Ferreira, Eduardo Gurzoni Alvares; Vieira, Ludmila Cabreira; Ferreira, Rafael Vicente de Padua; Silva, Edson Antonio da

    2013-01-01

    Radioactive liquid organic waste needs special attention because the available treatment processes are often expensive and difficult to be managed. Biosorption is a potential technique since it allies low cost with relatively high efficiency. Biosorption has been defined as the property of certain biomolecules to bind and remove selected ions or other molecules from aqueous solutions. Biosorption using vegetable biomass from agricultural waste has become a very attractive technique because it involves the removal of heavy metal ions by low cost biosorbent. This technique could be employed in the treatment of radioactive liquid wastes. Among the biosorbent reported in the literature, coconut fiber (Cocos nucifera L.) is highlighted due to the large number of functional groups in its composition. The aim of this study was to assess the potential of coconut fiber to remove uranium from radioactive liquid organic waste. This work was divided into three stages: 1) Preparation and activation of the coconut fiber; 2) Physical characterization of the biomass, 3) Batch biosorption experiments. Two forms of coconut fiber were tested, raw and activated. The activation was performed with dilute HNO3 and NaOH solutions. The parameters evaluated for physical characterization of biomass were morphological characteristics of coconut fiber, real and apparent density and surface area. The biomass was suspended in 10 ml of solutions prepared with distillate water and radioactive liquid waste for 2 hours in the proportion of 0.2% w/v. After the contact time, the coconut fiber was removed by filtration and the supernatant, analyzed by inductively coupled plasma optical emission spectrometry (ICP-OES).The results were evaluated using Langmuir and Freundlich isotherms. The maximum capacity for the raw coconut fiber was lower than the activated one, removing only 1.14mg/g against 2.61mg/g. These results suggest that biosorption with coconut fiber in activated form can be applied in the

  5. Biosorption of uranium in radioactive liquid organic waste by coconut fiber

    Energy Technology Data Exchange (ETDEWEB)

    Marumo, Julio Takehiro; Ferreira, Eduardo Gurzoni Alvares; Vieira, Ludmila Cabreira; Ferreira, Rafael Vicente de Padua, E-mail: jtmarumo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Silva, Edson Antonio da, E-mail: edson.silva2@unioeste.br [Universidade Estadual do Oeste do Parana (UNIOESTE), Toledo, PR (Brazil)

    2013-07-01

    Radioactive liquid organic waste needs special attention because the available treatment processes are often expensive and difficult to be managed. Biosorption is a potential technique since it allies low cost with relatively high efficiency. Biosorption has been defined as the property of certain biomolecules to bind and remove selected ions or other molecules from aqueous solutions. Biosorption using vegetable biomass from agricultural waste has become a very attractive technique because it involves the removal of heavy metal ions by low cost biosorbent. This technique could be employed in the treatment of radioactive liquid wastes. Among the biosorbent reported in the literature, coconut fiber (Cocos nucifera L.) is highlighted due to the large number of functional groups in its composition. The aim of this study was to assess the potential of coconut fiber to remove uranium from radioactive liquid organic waste. This work was divided into three stages: 1) Preparation and activation of the coconut fiber; 2) Physical characterization of the biomass, 3) Batch biosorption experiments. Two forms of coconut fiber were tested, raw and activated. The activation was performed with dilute HNO3 and NaOH solutions. The parameters evaluated for physical characterization of biomass were morphological characteristics of coconut fiber, real and apparent density and surface area. The biomass was suspended in 10 ml of solutions prepared with distillate water and radioactive liquid waste for 2 hours in the proportion of 0.2% w/v. After the contact time, the coconut fiber was removed by filtration and the supernatant, analyzed by inductively coupled plasma optical emission spectrometry (ICP-OES).The results were evaluated using Langmuir and Freundlich isotherms. The maximum capacity for the raw coconut fiber was lower than the activated one, removing only 1.14mg/g against 2.61mg/g. These results suggest that biosorption with coconut fiber in activated form can be applied in the

  6. Approach and issues toward development of risk-based release standards for radioactive scrap metal recycle and reuse

    International Nuclear Information System (INIS)

    Chen, S.Y.; Nieves, L.A.; Nabelssi, B.K.; LePoire, D.J.

    1994-01-01

    The decontamination and decommissioning of nuclear facilities is expected to generate large amounts of slightly radioactive scrap metal (RSM). It is likely that some of these materials will be suitable for recycling and reuse. The amount of scrap steel from DOE facilities, for instance, is estimated to be more than one million tons (Hertzler 1993). However, under current practice and without the establishment of acceptable recycling standards, the RSM would be disposed of primarily as radioactive low-level waste (LLW). In the United States, no specific standards have been developed for the unrestricted release of bulk contaminated materials. Although standards for unrestricted release of radioactive surface contamination (NRC 1974) have existed for about 20 years, the release of materials is not commonly practiced because of the lack of risk-based justifications. Recent guidance from international bodies (IAEA 1988) has established a basis for deriving risk-based release limits for radioactive materials. It is important, therefore, to evaluate the feasibility of recycling and associated issues necessary for the establishment of risk-based release limits for the radioactive metals

  7. An absorbent for an application to a package for a liquid radioactive isotope for medical usage

    International Nuclear Information System (INIS)

    Bang, K.S.; Lim, S.P.; Lee, J.C.; Seo, K.S.; Han, H.S.

    2004-01-01

    A radioactive isotope has to be safely transport from the producing center to the consuming center. The shipping package to safely transport the radioactive isotope should be able to withstand the prescribed conditions by law. In the field of nuclear medicine, the radioactive isotope is used in a liquid or capsule form. A Type A package, which is to transport liquid radioactive materials, shall be provided with a containment system composed of primary inner and secondary outer containment components or shall be provided with sufficient absorbent material to absorb twice the volume of the liquid contents. Hospitals prefer to use not only convenient but also re-usable packages. To apply an absorbent material to the Type A package, that is to transport liquid radioactive isotope, the free absorbency of the absorbents was estimated. In the case of a liquid with NaOH 0.4%, the free absorbency of the melanine form was the most superior at 91 g/g. In the case of a liquid with Na 0.9%, the free absorbency of the melanine form was the most excellent at 88 g/g also

  8. Decision Assessment of Clearance Level on Radioactive Waste Management

    International Nuclear Information System (INIS)

    Zainus Salimin; Gunandjar

    2007-01-01

    Radioactive waste on the safe level activity containing very small radioactive material gives small radiology influence to the human, it is not necessary to control by regulatory body. The radioactive waste on the safe level activity is safe to release as the common waste. For exemption of the control, it is required the safe activity level limits in which the value of clearance level is fulfilled by regulatory body, however until now it is not decided yet. The exemption decision is obtained if its activity is lower than or same with clearance level based on the annual effective dose receiving by public on the value is lower than or same with 0,01 mSv. The exposure pathways of radioactive waste to the human have important role for determination of clearance level. The decision assessment of clearance level on the radioactive waste management has been done by analysis of radioactive exposure pathways to the human for activities of the disposal and the recycle of solid wastes, also the release of liquid and gas effluent. For solid waste disposal, the exposure pathway was evaluated since the transportation of packed waste from the treatment facility to the disposal facility and during its operation. Exposure pathways for solid waste recycle consist of the pathways for handling and transportation of cleared material to the recycling facility, the fabrication and the utilization of its product. Exposure pathways for liquid and gas releases occur since its releases to the environment up to the human (public) by specific traffic lane. (author)

  9. Treatment of Medical Radioactive Liquid Waste Using Forward Osmosis (FO) Membrane Process

    KAUST Repository

    Lee, Songbok

    2018-04-07

    The use of forward osmosis (FO) for concentrating radioactive liquid waste from radiation therapy rooms in hospitals was systematically investigated in this study. The removal of natural and radioactive iodine using FO was first investigated with varying pHs and draw solutions (DSs) to identify the optimal conditions for FO concentration. Results showed that FO had a successful rejection rate for both natural and radioactive iodine (125I) of up to 99.3%. This high rejection rate was achieved at a high pH, mainly due to electric repulsion between iodine and membrane. Higher iodine removal by FO was also attained with a DS that exhibits a reverse salt flux (RSF) adequate to hinder iodine transport. Following this, actual radioactive medical liquid waste was collected and concentrated using FO under these optimal conditions. The radionuclides in the medical waste (131I) were removed effectively, but the water recovery rate was limited due to severe membrane fouling. To enhance the recovery rate, hydraulic washing was applied, but this had only limited success due to combined organic-inorganic fouling of the FO membrane. Finally, the effect of FO concentration on the reduction of septic tank volume was simulated as a function of recovery rate. To our knowledge, this study is the first attempt to explore the potential of FO technology for treating radioactive waste, and thus could be expanded to the dewatering of the radioactive liquid wastes from a variety of sources, such as nuclear power plants.

  10. Development of prototype liquid scintillator system for monitoring liquid radioactive waste

    International Nuclear Information System (INIS)

    Nam, Uk Won; Seon, Kwang Il; Kong, Kyoung Nam; Kim, Chang Kyu; Lee, Dong Myung; Lee, Sang Kook

    2003-01-01

    A prototype liquid scillatillator system for measurement of multiple beta-labeled mixtures was developed and its characteristic was investigated. The signal processing system consists of two photomultiplier tubes and the coincident count circuit. The characteristic of the system was analyzed using 4 beta-labeled samples ( 3 H, 14 C, 36 Cl and 90 Sr). Beta spectra from the samples were obtained without radiation shielding, and the detection limits for each nuclides were estimated based on the spectra. The estimated detection limits were compared to the legal regulation values. It is found that the liquid radioactive nuclides are detectable well below the legal regulation values

  11. Modeling of radionuclide migration and a temperature dynamics in underground disposal of liquid radioactive waste

    International Nuclear Information System (INIS)

    Larin, V.K.; Zubkov, A.A.; Balakhonov, V.G.; Sukhorukov, V.A.; Zhiganov, A.N.; Noskov, M.D.; Istomin, A.D.; Kesler, A.G.

    2002-01-01

    Mathematical model of radionuclide migration and temperature field dynamics during underground disposal of liquid radioactive wastes is presented. The model involves the description of filtration, convective-dispersion mass transfer, sorption and desorption of radionuclides, radioactive decay, convective heat transport and hear transfer. Software making possible to conduct prognosis calculations of changing state of stratum-collector of radioactive wastes was made. Results of the simulation of temperature field dynamics and behaviour of radionuclides on underground disposal of liquid radioactive wastes of the Siberian chemical plant are performed [ru

  12. Incident involving radioactive material at IAEA Safeguards Laboratory - No radioactivity released to environment

    International Nuclear Information System (INIS)

    2008-01-01

    Full text: Pressure build-up in a small sealed sample bottle in a storage safe resulted in plutonium contamination of a storage room at about 02:30 today at the IAEA's Safeguards Analytical Laboratory in Seibersdorf. All indications are that there was no release of radioactivity to the environment. Further monitoring around the laboratory will be undertaken. No one was working in the laboratory at the time. The Laboratory's safety system detected plutonium contamination in the storage room where the safe was located and in two other rooms - subsequently confirmed by a team of IAEA radiation protection experts. The Laboratory is equipped with multiple safety systems, including an air-filtering system to prevent the release of radioactivity to the environment. There will be restricted access to the affected rooms until they are decontaminated. A full investigation of the incident will be conducted. The IAEA has informed the Austrian regulatory authority. The IAEA's Laboratory in Seibersdorf is located within the complex of the Austrian Research Centers Seibersdorf (ARC), about 35 km southeast of Vienna. The laboratory routinely analyses small samples of nuclear material (uranium or plutonium) as part of the IAEA's safeguards verification work. (IAEA)

  13. Radioactive liquid wastes discharged to ground in the 200 areas during 1985

    International Nuclear Information System (INIS)

    Aldrich, R.C.

    1986-03-01

    This document summarizes radioactive liquids discharged to the ground in the 200 areas of the Hanford site and is provided pursuant to Department of Energy (DOE) Order 5484.1A, ''Environmental Protection, Safety, and Health Protection Information Reporting Requirements.'' There are twenty-eight liquid discharge streams in the 200 areas excluding sanitary sewers. Twenty-five streams were normally or potentially contaminated with radioactive material in 1985. Two streams had no potential for radioactive contamination but were included as adjustments in this report to maintain an accurate record of the total volume of the discharges to each disposal site. One stream, the 242-S Evaporator cooling water discharge, was not used during 1985

  14. Decontamination of radioactive liquid systems by modified clay minerals

    OpenAIRE

    Petrushka, Ihor; Moroz, Olexandr

    2016-01-01

    The process mechanism for sorption of strontium and cesium from liquid radioactive waste using modified bentonites from Yaziv sulfur deposit was investigated. The technique for predicting the intensity of the sorption process based on the comparison of experimental and calculated values of mass transfer coefficients was proposed. It was detected that the process of sorption extraction of strontium and cesium from liquid medium using modified clay minerals may be bes...

  15. Environmental Release Prevention and Control Plan

    International Nuclear Information System (INIS)

    Mamatey, A.; Arnett, M.

    1997-01-01

    During the history of SRS, continual improvements in facilities, process, and operations, and changes in the site''s mission have reduced the amount of radioactive liquid releases. In the early years of SRS (1958 to 1965), the amount of tritium discharged to the Savannah River averaged approximately 61,000 curies a year. During the mid-1980''s (1983 to 1988), liquid releases of tritium averaged 27,000 curies a year. By 1996, liquid releases of tritium are projected to be just 3000 curies for the year. This large projected decrease is the result of the planned shut-down of all reactors and the anticipated significant decline in the amount of tritium migrating from the site seepage basins and the Solid Waste Disposal Facility

  16. Liquid radioactive wastes from hospitals by polymeric membrane

    International Nuclear Information System (INIS)

    Arnal, J.M.; Sancho, M.; Verdu, G.; Campayo, J.M.

    1998-01-01

    Streams containing I''125 produced from RIA process, classified as radioactive waste of low activity, are generated by all different treatments applied in IN VITRO techniques. Consequently, an accumulation of solutions containing I''125 is produced in the order of 50-100 L/month approximately. The storage at sanitary centres and the accumulation caused by it creates a serious problem in the hospital. According to the specific activity and the installation spill authorization, one can choose between three ways of handling: direct discharge, temporal storage until the radioactive waste come to decay and then discharged, waste management by the authorised company (ENRESA). If the third way of discharge is applied the treatment of waste using membranes should be considered. Using membranes, important reduction coefficients in volume in the order of 10:1 are obtained. The aim of this work is the declassification of the I''125 solutions as a liquid radioactive waste using membrane techniques. Both, a radioactive concentrated waste and non-contaminated waste are obtained. (Author)

  17. INEEL Radioactive Liquid Waste Reduction Program

    International Nuclear Information System (INIS)

    Millet, C.B.; Tripp, J.L.; Archibald, K.E.; Lauerhauss, L.; Argyle, M.D.; Demmer, R.L.

    1999-01-01

    Reduction of radioactive liquid waste, much of which is Resource Conservation and Recovery Act (RCRA) listed, is a high priority at the Idaho National Technology and Engineering Center (INTEC). Major strides in the past five years have lead to significant decreases in generation and subsequent reduction in the overall cost of treatment of these wastes. In 1992, the INTEC, which is part of the Idaho National Environmental and Engineering Laboratory (INEEL), began a program to reduce the generation of radioactive liquid waste (both hazardous and non-hazardous). As part of this program, a Waste Minimization Plan was developed that detailed the various contributing waste streams, and identified methods to eliminate or reduce these waste streams. Reduction goals, which will reduce expected waste generation by 43%, were set for five years as part of this plan. The approval of the plan led to a Waste Minimization Incentive being put in place between the Department of Energy Idaho Office (DOE-ID) and the INEEL operating contractor, Lockheed Martin Idaho Technologies Company (LMITCO). This incentive is worth $5 million dollars from FY-98 through FY-02 if the waste reduction goals are met. In addition, a second plan was prepared to show a path forward to either totally eliminate all radioactive liquid waste generation at INTEC by 2005 or find alternative waste treatment paths. Historically, this waste has been sent to an evaporator system with the bottoms sent to the INTEC Tank Farm. However, this Tank Farm is not RCRA permitted for mixed wastes and a Notice of Non-compliance Consent Order gives dates of 2003 and 2012 for removal of this waste from these tanks. Therefore, alternative treatments are needed for the waste streams. This plan investigated waste elimination opportunities as well as treatment alternatives. The alternatives, and the criteria for ranking these alternatives, were identified through Value Engineering meetings with all of the waste generators. The most

  18. Radioactive liquid wastes discharged to ground in the 200 areas during 1974

    International Nuclear Information System (INIS)

    Anderson, J.D.

    1975-01-01

    Radioactive liquid wastes discharged to ground during 1974 and since startup within the Production and Waste Management control zone are summarized in tabular form. Estimates of the radioactivity discharged to individual ponds, cribs, and retention sites are also summarized. (LK)

  19. Treatment of low-level radioactive waste liquid by reverse osmosis

    International Nuclear Information System (INIS)

    Buckley, L.P.; Sen Gupta, S.K.; Slade, J.A.

    1995-01-01

    The processing of low-level radioactive waste (LLRW) liquids that result from operation of nuclear power plants with reverse osmosis systems is not common practice. A demonstration facility is operating at Chalk River Laboratories (of Atomic Energy of Canada Limited), processing much of the LLRW liquids generated at the site from a multitude of radioactive facilities, ranging from isotope production through decontamination operations and including chemical laboratory drains. The reverse osmosis system comprises two treatment steps--spiral wound reverse osmosis followed by tubular reverse osmosis--to achieve an average volume reduction factor of 30:1 and a removal efficiency in excess of 99% for most radioactive and chemical species. The separation allows the clean effluent to be discharged without further treatment. The concentrated waste stream of 3 wt% total solids is further processed to generate a solid product. The typical lifetimes of the membranes have been nearly 4000 hours, and replacement was required based on increased pressure drops and irreversible loss of permeate flux. Four years of operating experience with the reverse osmosis system, to demonstrate its practicality and to observe and record its efficiency, maintenance requirements and effectiveness, have proven it to be viable for volume reduction and concentration of LLRW liquids generated from nuclear-power-plant operations

  20. Biodegradation of ethyl acetate in radioactive liquid organic waste by bacterial communities

    International Nuclear Information System (INIS)

    Ferreira, Rafael V.P.; Sakata, Solange K.; Borba, Tania R.; Bellini, Maria H.; Marumo, Julio T.; Dutra, Fernando

    2009-01-01

    The research and development program in reprocessing of low burn-up spent fuel elements began in Brazil in 70's, originating the lab -scale hot cell, known as CELESTE located at IPEN-CNEN/SP. The program was ended at the beginning of 90's and part of the radioactive waste generated mainly from the analytical laboratories is stored at the Waste Management Laboratory. Among various types of radioactive waste generated, the organic liquid represents a major problem for its management, because it can not be directly solidified with cement. The objective of this work is to develop a pretreatment methodology to degrade the ethyl acetate present in organic liquid waste so that it can subsequently be immobilized in cement. This work was divided into two parts: selection and adaptation of three bacterial communities for growth in medium containing ethyl acetate and degradation experiments of ethyl acetate present in radioactive organic liquid waste. The results showed that from bacterial communities the highest biodegradation level observed was 77%. (author)

  1. Liquid waste processing at Comanche Peak

    International Nuclear Information System (INIS)

    Hughes-Edwards, L.M.; Edwards, J.M.

    1996-01-01

    This article describes the radioactive waste processing at Comanche Peak Steam Electric Station. Topics covered are the following: Reduction of liquid radioactive discharges (system leakage, outage planning); reduction of waste resin generation (waste stream segregation, processing methodology); reduction of activity released and off-site dose. 8 figs., 2 tabs

  2. Monitoring of low-level radioactive liquid effluent in Tokai reprocessing plant

    International Nuclear Information System (INIS)

    Mizutani, Tomoko; Koarashi, Jun; Takeishi, Minoru

    2009-01-01

    The Tokai reprocessing plant (TRP), the first reprocessing plant in Japan, has discharged low-level liquid wastes to the Pacific Ocean since the start of its operation in 1977. We have performed liquid effluent monitoring to realize an appropriate radioactive discharge control. Comparing simple and rapid analytical methods with labor-intensive radiochemical analyses demonstrated that the gross-alpha and gross-beta activities agreed well with the total activities of plutonium isotopes ( 238 Pu and 239+240 Pu) and major beta emitters (e.g., 90 Sr and 137 Cs), respectively. The records of the radioactive liquid discharge from the TRP showed that the normalized discharges of all nuclides, except for 3 H, were three or four orders of magnitude lower than those from the Sellafield and La Hague reprocessing plants. This was probably due to the installation of multistage evaporators in the liquid waste treatment process in 1980. The annual public doses for a hypothetical person were estimated to be less than 0.2 μSv y -1 from the aquatic pathway. Plutonium radioactivity ratios ( 238 Pu/ 239+240 Pu) of liquid effluents were determined to be 1.3-3.7, while those of the seabed sediment samples collected around the discharge point were 0.003-0.059, indicating no remarkable accumulation of plutonium in the regional aquatic environment. Thus, we concluded that there were no significant radiological effects on the public and the aquatic environment during the past 30-year operation of the TRP. (author)

  3. Radiological testing of products containing radioactivity

    International Nuclear Information System (INIS)

    Dixon, D.W.; Knight, A.

    1980-01-01

    Consumer products containing radioactive substances are tested by NRPB to determine how much radioactive material is likely to be released from a product if it is misused or accidentally damaged. Such testing is briefly described with particular reference to ionisation chamber smoke detectors, liquid crystal display watches illuminated with gaseous tritium light sources and anti-static brushes containing polonium-210 in the form of ceramic microspheres. (U.K.)

  4. Studies for improvement of regulatory control on the radioactive effluent released from nuclear facilities

    International Nuclear Information System (INIS)

    Cheong, Jae Hak; Park, H. M.; Song, M. C.; Lee, K. H.; Jang, J. K.; Chun, J. K.; Jeong, K. H.

    2005-05-01

    This report contains the second-year results of the research project titled 'Studies for Improvement of Regulatory Control on the Radioactive Effluent Released from Nuclear Facilities' and mainly provides technical and strategic approaches to improve performance of regulatory control on the gaseous effluent released from domestic nuclear facilities. The main result contained here includes overview and technical bases of radioactive gaseous effluent control (Chapter 1), reconsideration of the sensitivity requirements for measurement of radioactivity in gaseous effluent sample (Chapter 2), uncertainty analysis of the calculated radioactivity in gaseous effluent (Chapter 3), and improvement of quantification method of noble gas releases (Chapter 4). In addition, analysis of the impact due to combined sampling of particulate from multiple release points (Chapter 5), comparison of domestic nuclear reactors gaseous effluent data to foreign PWRs (Chapter 6), standardized sampling technique for collection of gaseous tritium (Chapter 7), and application of Xe-133 equivalent concept to gaseous effluent control (Chapter 8) are also provided. As a whole, this report provides a generic approach to improve the performance of regulatory control on the gaseous effluent. Therefore, actual enforcement of the recommendations should be preceded by establishment of a series of action plans reflecting on the site- and facility-specific design and operational features

  5. Management of Discharge of Low Level Liquid Radioactive Waste Generated in Medical, Educational, Research and Industrial Facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-07-15

    Although published information on management technologies suitable for radioactive effluents is readily available, smaller facilities such as hospitals, universities and research laboratories in some countries can benefit from more detailed guidance on identifying optimal arrangements for effectively managing their radioactive liquid effluents. A wide range of circumstances exist globally, given that the generation of radioactive liquid effluents may be regular or irregular, and the liquid effluents may be suitable for direct discharge to the environment, or may require a period of decay storage prior to discharge. Countries typically fit into one of the four following categories with respect to the status of their arrangements for the management of radioactive liquid effluents: (1) The country does not have sufficient technical, regulatory and organizational infrastructure to effectively manage its radioactive liquid effluents; (2) The country's technical infrastructure for effectively managing its radioactive liquid effluents is almost sufficient, but it is not supported by an acceptable level of regulatory and organizational capacity (e.g. legal infrastructure, administrative infrastructure); (3) The country has sufficient technical, regulatory and organizational capacity, but it is known that the application of the requirements for proper management of radioactive liquid effluents is, in many cases, not being carried out to the standard indicated by official reports; (4) The country has well developed and established regulatory and organizational capacity, which is complemented by an acceptable level of relevant technical infrastructure such that the radioactive liquid effluents can be properly managed. Facilities, as well as countries, in the first three categories will find information in this publication to assist their further development. Even countries that already have the necessary infrastructure to properly manage their liquid radioactive effluents may

  6. Management of Discharge of Low Level Liquid Radioactive Waste Generated in Medical, Educational, Research and Industrial Facilities

    International Nuclear Information System (INIS)

    2013-07-01

    Although published information on management technologies suitable for radioactive effluents is readily available, smaller facilities such as hospitals, universities and research laboratories in some countries can benefit from more detailed guidance on identifying optimal arrangements for effectively managing their radioactive liquid effluents. A wide range of circumstances exist globally, given that the generation of radioactive liquid effluents may be regular or irregular, and the liquid effluents may be suitable for direct discharge to the environment, or may require a period of decay storage prior to discharge. Countries typically fit into one of the four following categories with respect to the status of their arrangements for the management of radioactive liquid effluents: (1) The country does not have sufficient technical, regulatory and organizational infrastructure to effectively manage its radioactive liquid effluents; (2) The country's technical infrastructure for effectively managing its radioactive liquid effluents is almost sufficient, but it is not supported by an acceptable level of regulatory and organizational capacity (e.g. legal infrastructure, administrative infrastructure); (3) The country has sufficient technical, regulatory and organizational capacity, but it is known that the application of the requirements for proper management of radioactive liquid effluents is, in many cases, not being carried out to the standard indicated by official reports; (4) The country has well developed and established regulatory and organizational capacity, which is complemented by an acceptable level of relevant technical infrastructure such that the radioactive liquid effluents can be properly managed. Facilities, as well as countries, in the first three categories will find information in this publication to assist their further development. Even countries that already have the necessary infrastructure to properly manage their liquid radioactive effluents may

  7. Advances in technologies for the treatment of low and intermediate level radioactive liquid wastes

    International Nuclear Information System (INIS)

    1994-01-01

    In recent years the authorized maximum limits for radioactive discharges into the environment have been reduced considerably, and this, together with the requirement to minimize the volume of waste for storage or disposal and to declassify some wastes from intermediate to low level or to non-radioactive wastes, has initiated studies of ways in which improvements can be made to existing decontamination processes and also to the development of new processes. This work has led to the use of more specific precipitants and to the establishment of ion exchange treatment and evaporation techniques. Additionally, the use of combinations of some existing processes or of an existing process with a new technique such as membrane filtration is becoming current practice. New biotechnological, solvent extraction and electrochemical methods are being examined and have been proven at laboratory scale to be useful for radioactive liquid waste treatment. In this report an attempt has been made to review the current research and development of mature and advanced technologies for the treatment of low and intermediate level radioactive liquid wastes, both aqueous and non-aqueous. Non-aqueous radioactive liquid wastes or organic liquid wastes typically consist of oils, reprocessing solvents, scintillation liquids and organic cleaning products. A brief state of the art of existing processes and their application is followed by the review of advances in technologies, covering chemical, physical and biological processes. 213 refs, 33 figs, 3 tabs

  8. Ecotoxicological screen of Potential Release Site 50-006(d) of Operable Unit 1147 of Mortandad Canyon and relationship to the Radioactive Liquid Waste Treatment Facilities project

    International Nuclear Information System (INIS)

    Gonzales, G.J.; Newell, P.G.

    1996-04-01

    Potential ecological risk associated with soil contaminants in Potential Release Site (PRS) 50-006(d) of Mortandad Canyon at the Los Alamos National Laboratory was assessed by performing an ecotoxicological risk screen. The PRS surrounds Outfall 051, which discharges treated effluent from the Radioactive Liquid Waste Treatment Facility. Discharge at the outfall is permitted under the Clean Water Act National Pollution Discharge Elimination System. Radionuclide discharge is regulated by US Department of Energy (DOE) Order 5400.5. Ecotoxicological Screening Action Levels (ESALSs) were computed for nonradionuclide constituents in the soil, and human risk SALs for radionuclides were used as ESALs. Within the PRS and beginning at Outfall 051, soil was sampled at three points along each of nine linear transects at 100-ft intervals. Soil samples from 3 depths for each sampling point were analyzed for the concentration of a total of 121 constituents. Only the results of the surface sampling are reported in this report

  9. Wow Technology’s innovative radioactive liquid waste treatment

    Energy Technology Data Exchange (ETDEWEB)

    Marin, A.

    2015-07-01

    WOW presents its revolutionary technology and equipment for liquid radioactive waste treatment: outperforming ultimate water decontamination and purification process, enhanced sludge concentration, no secondary waste nor consumables, fully automated, remote controlled and self-decontaminating device. The WOW’s technology is based upon a never before observed discovery of fluid dynamics science: the possibility of performing a molecular separation between solute and suspended elements and the solvent. The combination of such a molecular separation process with a standard vacuum evaporation improves the abatement performances by thousands of times, with respect to those of the state of the art vacuum evaporators. In addition to this, no secondary waste is produced during the process, as no filters, membranes, resins or additives are used. WOW equipment, automated and remote controlled, self decontaminates after use and can be designed and constructed either tailored to the application needs or with a modular approach for enhanced transportability and application flexibility. After the preliminary verification by CNR, the Italian National Research Center, Wow Technology decontamination device was tested c/o LENA, the Laboratory of Applied Nuclear Energy of the University of Pavia, Italy with a simulated solution 6000 times more contaminated than the nuclear reactor’s cooling water of Fukushima-Daiichi NPP. In addition to that, WOW Technology was also used in a real case at the Radiochemistry laboratory of the Pavia’s University Chemistry department. Both the above mentioned contaminated fluids have been successfully decontaminated without production of additional or secondary waste WOW Technology has already performed on industrial scale c/o the Nuclear Repository of S.S.M. in Saluggia, Italy: 45000 liters of acid radioactive solution have been successfully decontaminated to a Decontamination Factor (DF) of 335000 for Cs-137 by one single evaporation step and

  10. ICPP radioactive liquid and calcine waste technologies evaluation final report and recommendation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    Using a formalized Systems Engineering approach, the Latched Idaho Technologies Company developed and evaluated numerous alternatives for treating, immobilizing, and disposing of radioactive liquid and calcine wastes at the Idaho Chemical Processing Plant. Based on technical analysis data as of March, 1995, it is recommended that the Department of Energy consider a phased processing approach -- utilizing Radionuclide Partitioning for radioactive liquid and calcine waste treatment, FUETAP Grout for low-activity waste immobilization, and Glass (Vitrification) for high-activity waste immobilization -- as the preferred treatment and immobilization alternative.

  11. Application of membrane technologies for liquid radioactive waste processing

    International Nuclear Information System (INIS)

    2004-01-01

    Membrane separation processes have made impressive progress since the first synthesis of membranes almost 40 years ago. This progress was driven by strong technological needs and commercial expectations. As a result the range of successful applications of membranes and membrane processes is continuously broadening. In addition, increasing application of membrane processes and technologies lies in the increasing variations of the nature and characteristics of commercial membranes and membrane apparatus. The objective of the report is to review the information on application of membrane technologies in the processing of liquid radioactive waste. The report covers the various types of membranes, equipment design, range of applications, operational experience and the performance characteristics of different membrane processes. The report aims to provide Member States with basic information on the applicability and limitations of membrane separation technologies for processing liquid radioactive waste streams

  12. Radioactivity release vs probability for a steam generator tube rupture accident

    International Nuclear Information System (INIS)

    Buslik, A.J.; Hall, R.E.

    1978-01-01

    A calculation of the probability of obtaining various radioactivity releases from a steam generator tube rupture (SGTR) is presented. The only radioactive isotopes considered are Iodine-131 and Xe-133. The particular accident path considered consists of a double-ended guillotine SGTR followed by loss of offsite power (LOSP). If there is no loss of offsite power, and no system fault other than the SGTR, it is judged that the consequences will be minimal, since the amount of iodine released through the condenser air ejector is expected to be quite small; this is a consequence of the fact that the concentration of iodine in the vapor released from the condenser air ejector is very small compared to that dissolved in the condensate water. In addition, in some plants the condenser air ejector flow is automatically diverted to containment or a high-activity alarm. The analysis presented here is for a typical Westinghouse PWR such as described in RESAR-3S

  13. Detection of free liquid in cement-solidified radioactive waste drums using computed tomography

    International Nuclear Information System (INIS)

    Steude, J.S.; Tonner, P.D.

    1991-01-01

    Acceptance criteria for disposal of radioactive waste drums require that the cement-solidified material in the drum contain minimal free liquid after the cement has hardened. Free liquid is to be avoided because it may corrode the drum, escape and cause environmental contamination. The DOE has requested that a nondestructive evaluation method be developed to detect free liquid in quantities in excess of 0.5% by volume. This corresponds to about 1 liter in a standard 208 liter (55 gallon) drum. In this study, the detection of volumes of free liquid in a 57 cm (2 ft.) diameter cement-solidified drum is demonstrated using high-energy X-ray computed tomography (CT0. In this paper it is shown that liquid concentrations of simulated radioactive waste inside glass tubes imbedded in cement can easily be detected, even for tubes with inner diameters less than 2 mm (0.08 in.). Furthermore, it is demonstrated that tubes containing water and liquid concentrations of simulated radioactive waste can be distinguished from tubes of the same size containing air. The CT images were obtained at a rate of about 6 minutes per slice on a commercially available CT system using a 9 MeV linear accelerator source

  14. Radioactive release data from Canadian nuclear generating stations 1872-1987

    International Nuclear Information System (INIS)

    1989-03-01

    All nuclear generating stations emit small quantities of radioactive effluent both into the atmosphere and in the form of liquid effluent, into the adjoining water body, be it river, lake or sea. The purpose of this document is to report on the magnitude of these emissions for each nuclear generating station in Canada and to indicate how these emissions compare with the relevant limitations imposed by the Atomic Energy Control Board as part of its regulatory and licensing program. This report incorporates histograms indicating the annual releases of tritium in air, noble gases, iodine-131, airborne particulates, tritium in water and waterborne gross beta activity for each nuclear generating station. In addition, for Pickering NGS 'A', annual released of carbon-14 are depicted for the years 1986 and 1987. In each case the emission data are compared to the Derived Emission Limit (DEL) in order that the data may be placed in perspective. At present, only Pickering NGS 'A' is required to monitor and report carbon-14 emissions. Environmental monitoring for C-14 is conducted around the Bruce site to determine the environmental impact of its emission and whether effluent monitoring will be necessary in future years. Three nuclear generating stations have been permanently taken out of service during the last few years (Gentilly NGS-1, Douglas Point NGS and NPD NGS). Some small emissions from these sites do still occur, however, due to decontamination and decommissioning operations. (11 tabs., 26 figs.)

  15. Offsite doses from SRP radioactive releases - 1985

    International Nuclear Information System (INIS)

    Marter, W.L.

    1986-01-01

    This memorandum summarizes the offsite doses from releases of radioactive materials to the environment from SRP operations in 1985. These doses were calculated for inclusion in the environmental report for 1985 to be issued by the Health Protection Department (DPSPU-86-30-1). The environmental report is prepared annually for distribution to state environmental agencies, the news media, and interested members of the public. More detailed data on offsite exposures by radionuclide and exposure pathway will be included in the environmental report

  16. Sensitivity Analysis of Dousing Spray Trip on Radioactive Release in Pressure Tube Rupture Accident with Both End Fitting Failures

    Energy Technology Data Exchange (ETDEWEB)

    Jang, M. S.; Kang, H. S; Kim, S. R. [NESS, Daejeon (Korea, Republic of)

    2015-10-15

    We analyzed the sensitivity analysis of dousing spray trip conditions on radioactive release. In terms of conservativeness, the set 1 trip would be more appropriate in RR analysis than set 2 trip, which is the general condition of RR analysis. Radioactive releases from the containment building is related to containment air pressure, which increases by the coolant discharge from loss of coolant accident and the actuation conditions of dousing spray and so on. In LOCA analysis, the dousing spray trip conditions are set for the analysis objectives; for peak pressure (PP), for pressure signal (PS), for radioactive release (RR) and etc. In RR analysis, we would determine the dousing spray trip condition to increase radioactive release to the public for conservatism. Therefore, we carried out the sensitivity analysis of dousing spray trip condition on radioactive release from containment building using GOTHIC and SMART program for CANDU.

  17. Disposal containers for radioactive waste materials and separation systems for radioactive waste materials

    International Nuclear Information System (INIS)

    Rubin, L.S.

    1986-01-01

    A separation system for dewatering radioactive waste materials includes a disposal container, drive structure for receiving the container, and means for releasably attaching the container to the drive structure. The separation structure disposed in the container adjacent the inner surface of the side wall structure retains solids while allowing passage of liquids. The inlet port structure in the container top wall is normally closed by first valve structure that is centrifugally actuated to open the inlet port and the discharge port structure at the container periphery receives liquid that passes through the separation structure and is normally closed by a second valve structure that is centrifugally actuated to open the discharge ports. The container also includes a coupling structure for releasable engagement with the centrifugal drive structure. The centrifugal force produced when the container is driven in rotation by the drive structure opens the valve structures, and radioactive waste material introduced into the container through the open inlet port is dewatered, and the waste is compacted. The ports are automatically closed by the valves when the container drum is not subjected to centrifugal force such that containment effectiveness is enhanced and exposure of personnel to radioactive materials is minimized. (author)

  18. Heat transfer enhanced microwave process for stabilization of liquid radioactive waste slurry. Final report

    International Nuclear Information System (INIS)

    White, T.L.

    1995-01-01

    The objectve of this CRADA is to combine a polymer process for encapsulation of liquid radioactive waste slurry developed by Monolith Technology, Inc. (MTI), with an in-drum microwave process for drying radioactive wastes developed by Oak Ridge National Laboratory (ORNL), for the purpose of achieving a fast, cost-effectve commercial process for solidification of liquid radioactive waste slurry. Tests performed so far show a four-fold increase in process throughput due to the direct microwave heating of the polymer/slurry mixture, compared to conventional edge-heating of the mixer. We measured a steady-state throughput of 33 ml/min for 1.4 kW of absorbed microwave power. The final waste form is a solid monolith with no free liquids and no free particulates

  19. Environmental radioactivity at the National Nuclear Research Centre, Pelindaba

    International Nuclear Information System (INIS)

    Brits, R.J.N; Prinsloo, L.; De Jesus, A.S.M.

    1981-07-01

    The revised environmental survey program, introduced during 1970 with the emphasis on monitoring of the critical paths of exposure of the general public, was continued in 1980. Results of determinations of both gross radioactivity and individual nuclides in samples of fish and water (which are critical materials for liquid-effluent releases) from the Hartbeespoort Dam and from the Crocodile River, are given and discussed. Results of 131 I, 0 Sr and gamma-spectrometric analyses of milk, the critical material for releases to the atmosphere, are presented. Results are given of regular investigations of the composition of airborne releases to the atmosphere and liquid-effluent releases to the Crocodile River, performed in order to detect other possible critical nuclides. Levels of deposited and airborn activity from nuclear-bomb tests are reported. Due to the absence of fresh fallout, the levels for most fission products have fallen below the limit of detection. No environmental radioactivity resulting from NNRC releases could be detected above the natural background or accumulated fallout levels [af

  20. Research and Development of Solar Evaporation on Low Level Radioactive Liquid Waste

    Directory of Open Access Journals (Sweden)

    ZHANG Hua

    2016-02-01

    Full Text Available Solar evaporation, which can save energy and obtain the higher decontamination factor, the larger treatment capability with the simpler designed and easy operation, was one of the general methods to treat low level radioactive liquid waste. However, the use of solar evaporation was limited because the facilities had to occupy the larger area and require sunshine for the longer duration, etc. Several cases form USA, Australian, India and South Korea were presented on R&D of solar evaporation to treat low level radioactive liquid waste.

  1. Treatment of radioactive liquid waste by tubular type reverse osmosis module

    International Nuclear Information System (INIS)

    Nishimaki, Kenzo; Koyama, Akio; Tsutsui, Tenson; Mori, Koji.

    1988-01-01

    The applicability of reverse osmosis to radioactive liquid waste treatment was studied using a tubular type module. When four modules were used in a series, circulating volume of concentrate was much greater than permeate volume, therefore solute concentration and circulating rate of concentrate can be assumed uniform in the axial direction of the modules. DFs of stable elements contained in the tap water were 36-40 for Na, 50-55 for K, 170-250 for Mg and 90-160 for Ca. When Na concentration increased about ten times, DFs for all elements slightly decreased. For actual liquid waste tagged with radionuclides, DFs were in the range of 35-40 for 134 Cs, 150-200 for 85 Sr, and 180-280 for 58 Co. These DF values indicate the possibility of the treatment of low radioactive liquid waste by reverse osmosis. (author)

  2. Methods for determining the release of radioactive material into the environment

    International Nuclear Information System (INIS)

    Farges, L.; Daw, H.T.

    1976-01-01

    The current policy on the discharge of radioactive effluents calls for the containment of radioactive wastes in most instances. The resulting doses to individuals and populations have been shown by many surveys to be very small (UNSCEAR Report, 1972). Nevertheless, small amounts of radioactive releases are made to the environment during normal operation of nuclear facilities. Whenever discharge of radioactive effluents to the environment is permitted, careful consideration is made of all the relevant factors which might lead to exposure of man. However, with the expansion of nuclear power programmes, more sophistication is required in setting discharge limits to the environment, taking into consideration future sources as well as present sources. The IAEA's recommendations conform to the dose limitation system set out by the International Commission on Radiological Protection (ICRP). The system implies that the individual dose limits should never be exceeded. Furthermore, the ICRP guidelines require that doses be kept as low as reasonably possible, taking social and economic considerations into account. This second objective, usually called optimization, implies the use of differential cost-benefit analysis. At present, decisions are still made by other procedures, for example, by applying safety factors to release limits derived only from the dose limits. However, the ICRP system of dose limitation, including optimization, appears to be a more rational approach to the establishment of release limits. Thus, it is necessary to provide basic material on concepts which are intended for use and decision making by national authorities, and the Agency plans to publish a series of complementary documents on the application of these concepts to various specific cases. (author)

  3. Disposal of liquid radioactive waste - discharge of radioactive waste waters from hospitals

    International Nuclear Information System (INIS)

    Ludwieg, F.

    1976-01-01

    A survey is given about legal prescriptions in the FRG concerning composition and amount of the liquid waste substances and waste water disposal by emitting into the sewerage, waste water decay systems and collecting and storage of patients excretions. The radiation exposure of the population due to drainage of radioactive waste water from hospitals lower by more than two orders than the mean exposure due to nuclear-medical use. (HP) [de

  4. Order of 21 October 1988 withdrawing the licence for the release of liquid radioactive effluents by the Cattenom nuclear production centre (units 1 and 2)

    International Nuclear Information System (INIS)

    1988-01-01

    The Court of Justice of the European Communities decided on 22 September 1988 that the Commission of the European Communities had to be notified and give its opinion before the competent authorities of Member States authorised the release of radioactive effluents from a nuclear installation. In compliance with that judgment, this Order repeals an Order of 21 February 1986 licensing such release (NEA) [fr

  5. Site-sensitive hazards of potential airborne radioactive release from sources on the Kola peninsula

    International Nuclear Information System (INIS)

    Bergman, R.; Thaning, L.; Baklanov, A.

    1998-02-01

    In this work we focus on cases of airborne releases from some of the sources on the Kola Peninsula - primarily nuclear reactors on submarines and the Kola Nuclear Power Plant (KNPP). The purpose of our study is to illustrate, and discuss some features - dependent on site and release characteristics - of the deposition patterns resulting from assumed unit radioactive releases to the atmosphere from a location at a fjord and from the KNPP in Polyarnye Zori. Using meteorological data for one real weather situation, the analysis is based on simulating the transport in air of assumed radioactive releases and estimating the deposition pattern on local, meso- and regional scales. By allowing unit releases to occur simultaneously from the site at the fjord and from the power plant (and with the same release profile in time) comparisons are made of differences in deposition patterns in and outside the Kola region. In this case study a set of assumed release heights, durations of the release, and particle size distributions are applied to indicate the dependence for the resulting deposition pattern on these parameters

  6. Experiment of decontamination of radioactive liquid by a biological method; Experience de decontamination de liquides radioactifs far voie biologique

    Energy Technology Data Exchange (ETDEWEB)

    Wormser, G.

    1962-07-01

    The author reports experiments of treatment of radioactive liquid effluents by percolation on a bacterial bed like the one used for the treatment of sewer wastewaters. He also reports results obtained in other countries in terms of reduction of effluent radioactivity for various radioactive ions. The installation is described and results are presented in terms of variation of contamination of an effluent with respect to its recycling on a bacterial bed [French] Dans le monde entier, on se preoccupe des moyens de decontamination pour des liquides radioactifs. Les experiences de l'auteur ont confirme qu'un lit bacterien neuf peut donner de bons resultats: il est a noter que ce procede biologique se montre selectif a l'egard des divers ions radioactifs. (auteur)

  7. Generation projection of solid and liquid radioactive wastes and spent radioactive sources in Mexico; Proyeccion de generacion de desechos radiactivos solidos, liquidos y fuentes radiactivas gastadas en Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Garcia A, E.; Hernandez F, I. Y.; Fernandez R, E. [Universidad Politecnica del Valle de Toluca, Km 5.7 Carretera Almoloya de Juarez, Estado de Mexico (Mexico); Monroy G, F.; Lizcano C, D., E-mail: fabiola.monroy@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    This work is focused to project the volumes of radioactive aqueous liquid wastes and spent radioactive sources that will be generated in our country in next 15 years, solids compaction and radioactive organic liquids in 10 years starting from the 2014; with the purpose of knowing the technological needs that will be required for their administration. The methodology involves six aspects to develop: the definition of general objectives, to specify the temporary horizon of projection, data collection, selection of the prospecting model and the model application. This approach was applied to the inventory of aqueous liquid wastes, as well as radioactive compaction organic and solids generated in Mexico by non energy applications from the 2001 to 2014, and of the year 1997 at 2014 for spent sources. The applied projection models were: Double exponential smoothing associating the tendency, Simple Smoothing and Lineal Regression. For this study was elected the first forecast model and its application suggests that: the volume of the compaction solid wastes, aqueous liquids and spent radioactive sources will increase respectively in 152%, 49.8% and 55.7%, while the radioactive organic liquid wastes will diminish in 13.15%. (Author)

  8. An analytical model for radioactive pollutant release simulation in the atmospheric boundary layer

    International Nuclear Information System (INIS)

    Weymar, Guilherme J.; Vilhena, Marco T.; Bodmann, Bardo E.J.; Buske, Daniela; Quadros, Regis

    2013-01-01

    Simulations of emission of radioactive substances in the atmosphere from the Brazilian nuclear power plant Angra 1 are a necessary tool for control and elaboration of emergency plans as a preventive action for possible accidents. In the present work we present an analytical solution for radioactive pollutant dispersion in the atmosphere, solving the time-dependent three-dimensional advection-diffusion equation. The experiment here used as a reference in the simulations consisted of the controlled releases of radioactive tritiated water vapor from the meteorological tower close to the power plant at Itaorna Beach. The wind profile was determined using experimental meteorological data and the micrometeorological parameters were calculated from empirical equations obtained in the literature. We report on a novel analytical formulation for the concentration of products of a radioactive chain released in the atmospheric boundary layer and solve the set of coupled equations for each chain radionuclide by the GILTT solution, assuming the decay of all progenitors radionuclide for each equation as source term. Further we report on numerical simulations, as an explicit but fictitious example and consider three radionuclides in the radioactive chain of Uranium 235. (author)

  9. An analytical model for radioactive pollutant release simulation in the atmospheric boundary layer

    Energy Technology Data Exchange (ETDEWEB)

    Weymar, Guilherme J.; Vilhena, Marco T.; Bodmann, Bardo E.J., E-mail: guicefetrs@gmail.com, E-mail: mtmbvilhena@gmail.com, E-mail: bejbodmann@gmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Buske, Daniela; Quadros, Regis, E-mail: danielabuske@gmail.com, E-mail: quadros99@gmail.com [Universidade Federal de Pelotas (UFPel), Capao do Leao, RS (Brazil). Programa de Pos-Graduacao em Modelagem Matematica

    2013-07-01

    Simulations of emission of radioactive substances in the atmosphere from the Brazilian nuclear power plant Angra 1 are a necessary tool for control and elaboration of emergency plans as a preventive action for possible accidents. In the present work we present an analytical solution for radioactive pollutant dispersion in the atmosphere, solving the time-dependent three-dimensional advection-diffusion equation. The experiment here used as a reference in the simulations consisted of the controlled releases of radioactive tritiated water vapor from the meteorological tower close to the power plant at Itaorna Beach. The wind profile was determined using experimental meteorological data and the micrometeorological parameters were calculated from empirical equations obtained in the literature. We report on a novel analytical formulation for the concentration of products of a radioactive chain released in the atmospheric boundary layer and solve the set of coupled equations for each chain radionuclide by the GILTT solution, assuming the decay of all progenitors radionuclide for each equation as source term. Further we report on numerical simulations, as an explicit but fictitious example and consider three radionuclides in the radioactive chain of Uranium 235. (author)

  10. Improvement of the safety regulations in the management of radioactive waste accumulated in the liquid radioactive waste water basins of the PO 'Majak' (Ozersk), the Siberian Chemical Plant (Seversk) and the Mining-Chemical Plant (Zheleznogorsk)

    International Nuclear Information System (INIS)

    Vishnevski, Y.G.; Kislov, A.I.; Irushkin, V.M.

    2002-01-01

    One of the most important problems of radiation safety in Russia is the decommissioning of the liquid radioactive waste water basins of the PO 'Majak' (Ozersk), Siberian Chemical Plant (Seversk) and Mining-Chemical Plant (Zheleznogorsk). The liquid radioactive waste water basins were constructed in 1950-1960 for the collection and storage of liquid waste from the radiochemical plants. The potential hazards of the liquid in the radioactive waste water basins are: migration of radionuclides into the soil of the liquid radioactive waste water basin floors; wind-induced carry-over of radionuclides from the liquid radioactive waste water basins; hazards (radiation included) to the environment and population arising in case physical barriers and hydraulic structures are damaged; and criticality hazards. The classification of the liquid radioactive waste water basins were developed based on the collection and analyzes of the information on liquid radioactive waste water basin characteristics and the method of multicriterion expert assessment of potential hazards. Three main directions for the improvement of safety regulation in the management of radioactive waste accumulated in the liquid radioactive waste water basins were defined: 1. Common directions for the improvement of safety regulation in the area of rehabilitation of the territories contaminated with radioactive substances. 2. Common directions for the improvement of safety regulation in the area of rehabilitation of the territories, such as the liquid radioactive waste water basins. 3. Special directions for the regulatory activities in the area of operation and decommissioning of the liquid radioactive waste water basins of the PO 'Majak' (Ozersk), Siberian Chemical Plant (Seversk) and Mining-Chemical Plant (Zheleznogorsk). As a result, concrete recommendations on safety regulation for the management of radioactive waste accumulated in the water basins were developed. (author)

  11. Development of new waste form for treatment and disposal of concentrated liquid radioactive waste

    International Nuclear Information System (INIS)

    Kwak, Kyung Kil; Ji, Young Yong

    2010-12-01

    The radioactive waste form should be meet the waste acceptance criteria of national regulation and disposal site specification. We carried out a characterization of rad waste form, especially the characteristics of radioactivity, mechanical and physical-chemical properties in various rad waste forms. But asphalt products is not acceptable waste form at disposal site. Thus we are change the product materials. We select the development of the new process or new materials. The asphalt process is treatment of concentrated liquid and spent-resin and that we decide the Development of new waste form for treatment and disposal of concentrated liquid radioactive waste

  12. Risk comparison of different treatment and disposal strategies of high level liquid radioactive waste

    International Nuclear Information System (INIS)

    Fang Dong

    1997-01-01

    The risk of different treatment and disposal strategies of high level liquid radioactive waste from spent fuel reprocessing is estimated and compared. The conclusions obtained are that risk difference from these strategies is very small and high level liquid waste can be reduced to middle and low level waste, if the decontamination factor for 99 Tc is large enough, which is the largest risk contributor in the high level radioactive waste from spent fuel reprocessing. It is also shown that the risk of high level radioactive waste could be reduced by the technical strategy of combining partitioning and transmutation

  13. Use of OND-86 recommendations for calculation of the Shelter radioactive release

    International Nuclear Information System (INIS)

    Bogatov, S.A.

    2000-01-01

    A model of radioactive release from the Shelter has been considered under current operation conditions. Integral assessment of current dust release has been done on the base of natural ventilation rate. Model predictions are consistent (20% accuracy) with experimental results of air contamination measurements at the earth surface. 12 refs., 1 tab., 6 figs

  14. Use of Eichornia crassipes for treatment of low level liquid radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Hafez, N.; Ramadan, Y.S.; Hassanin, R.A.; Gafez, M.B. (Atomic Energy Authority, Hot Lab. Center, Cairo (Egypt))

    1993-01-01

    Radioactive and non-radioactive isotopes of cobalt, cerium and cesium were found to be accumulated inside Eichornia crassipes (the water hyacinth). The rate and extent of accumulation were dependent upon environmental parameters such as pH, temperature and interference by certain anions and cations. The accumulation rate of radioactive isotopes inside Eichornia crassipes, were more rapid than non-active ions. The results showed that accumulation of such metals inside the plant could be used successfully in the treatment of low-level liquid radioactive wastes. (author) 4 figs., 2 tabs., 15 refs.

  15. Use of Eichornia crassipes for treatment of low level liquid radioactive waste

    International Nuclear Information System (INIS)

    Hafez, N.; Ramadan, Y.S.; Hassanin, R.A.; Gafez, M.B.

    1993-01-01

    Radioactive and non-radioactive isotopes of cobalt, cerium and cesium were found to be accumulated inside Eichornia crassipes (the water hyacinth). The rate and extent of accumulation were dependent upon environmental parameters such as pH, temperature and interference by certain anions and cations. The accumulation rate of radioactive isotopes inside Eichornia crassipes, were more rapid than non-active ions. The results showed that accumulation of such metals inside the plant could be used successfully in the treatment of low-level liquid radioactive wastes. (author) 4 figs., 2 tabs., 15 refs

  16. Characterization of the solid radioactive waste from Cernavoda NPP

    International Nuclear Information System (INIS)

    Iordache, M.; Lautaru, V.; Bujoreanu, D.

    2005-01-01

    During the operation of a nuclear plant significant quantities of radioactive waste result that have a very large diversity. At Cernavoda NPP large amounts of wastes are either non-radioactive wastes or radioactive wastes, each of these being managed completely different from each other. For a CANDU type reactor, the occurrence of radioactive wastes is due to contamination with the following types of radioactive substances: - fission products resulting from nuclear fuel burning; - activated products from materials composing the technological systems; - activated products in process fluids. Radioactive wastes can be in solid, liquid or gas form. At Cernavoda NPP the solid wastes represent about 70% of the waste volume which is produced during plant operation and as a consequence of maintenance and decontamination operations. The most important types of solid wastes that are obtained and then handled, processed (if necessary) and temporarily stored are: solid low-level radioactive wastes (classified as compactible and non-compactible), solid medium radioactive wastes, spent resins, used filters and filter cartridges. The liquid radioactive waste class includes organic liquids (used oil, scintillator liquids and used solvents) and aqueous wastes resulting from process system operating, from decontamination and maintenance operations. Radioactive gas wastes occur subsequently to the fission process inside the fuel elements as well as due to the neutron activation of process fluids in the reactor systems. As result of plant operation, iodine, noble gases, tritium and radioactive particles occur and are passed toward the ventilation stack in a controlled manner so that environmental release of radioactive materials with concentrations exceeding the maximum permissible level could not occur. (authors)

  17. Application of dynamic modeling for assessing the impact of radioactive releases to water sewer systems

    International Nuclear Information System (INIS)

    Sundell-Bergman, Synnove; Avila, Rodolfo; Cruz, Idalmis de la

    2008-01-01

    Full text: Potential radiological emergencies due to accidental or malicious acts involving nuclear materials requires tools for emergency dose assessment to help in the planning of countermeasures. In urban areas, the municipal sewage systems will receive the wastewater from households, industries and hospitals as well as the run-off water. Investigations have shown that sewage sludge is a sensitive indicator for radionuclides released from hospitals or spread via the atmosphere and thus simulation modelling of the fate of radionuclides entering sewage treatment plants via urban drainage systems could prove useful in emergency situations. A dynamic model (LUCIA) has been developed to assess the radiological consequences of non-homogenous releases of radionuclides to the sewage plants. In the first step the focussing has been on radioactive liquid releases from hospitals and the predictions show that there is a significant probability (> 0,2) that the sewage worker doses exceed 10 μSv/y while doses to farmers or fishermen (downstream plants) are marginal. Comparison of measured and estimated concentration values for iodine-131 reveal that the predictions made by LUCIA are reasonably good and fall within a factor of 2. For the purpose of emergency preparedness, scenarios have been defined and the fate of released radionuclides has been assessed. The main features of the LUCIA model will be presented and the application of the tool for emergency preparedness will be discussed. (author)

  18. Solidification of liquid radioactive concentrates by fixation with cement

    International Nuclear Information System (INIS)

    Pekar, A.; Breza, M.; Timulak, J.; Krajc, T.

    1985-01-01

    In testing the technology of liquid radioactive wastes cementation, the effect was mainly studied of the content of boric acid and its salts on cement solidification, the effect of additives on radionuclide leachability and the effect of the salt content on the cementation product. On the basis of experimental work carried out on laboratory scale with model samples and samples of radioactive concentrate from the V-1 nuclear power plant, the following suitable composition of the cementation mixture was determined: 40% Portland cement, 40% zeolite containing material and 20% power plant ash. The most suitable ratio of liquid radioactive wastes and the cementation mixture is 0.5. As long as in such case the salt content of the concentrate ranges between 20 and 25%, the cementation product will have a maximum salt content of 10% and a leachability of the order of 10 -3 to 10 -4 g/cm 2 per day with a mechanical strength allowing safe handling. It was also found that the quality processing of the cement paste with degassing, e.g., by vibration, is more effective for the production of a pore-free cementation product than the application of various additives which are supposed to eliminate pore formation. (Z.M.)

  19. Radioactive Liquid Waste Treatment Facility Discharges in 2011

    Energy Technology Data Exchange (ETDEWEB)

    Del Signore, John C. [Los Alamos National Laboratory

    2012-05-16

    This report documents radioactive discharges from the TA50 Radioactive Liquid Waste Treatment Facilities (RLWTF) during calendar 2011. During 2011, three pathways were available for the discharge of treated water to the environment: discharge as water through NPDES Outfall 051 into Mortandad Canyon, evaporation via the TA50 cooling towers, and evaporation using the newly-installed natural-gas effluent evaporator at TA50. Only one of these pathways was used; all treated water (3,352,890 liters) was fed to the effluent evaporator. The quality of treated water was established by collecting a weekly grab sample of water being fed to the effluent evaporator. Forty weekly samples were collected; each was analyzed for gross alpha, gross beta, and tritium. Weekly samples were also composited at the end of each month. These flow-weighted composite samples were then analyzed for 37 radioisotopes: nine alpha-emitting isotopes, 27 beta emitters, and tritium. These monthly analyses were used to estimate the radioactive content of treated water fed to the effluent evaporator. Table 1 summarizes this information. The concentrations and quantities of radioactivity in Table 1 are for treated water fed to the evaporator. Amounts of radioactivity discharged to the environment through the evaporator stack were likely smaller since only entrained materials would exit via the evaporator stack.

  20. Management of radioactive liquid waste at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Bendixsen, C.L.

    1992-01-01

    Highly radioactive liquid wastes (HLLW) are routinely produced during spent nuclear fuel processing at the Idaho Chemical Processing Plant (ICPP), located at the Idaho National Engineering Laboratory (INEL). This paper discusses the processes and safe practices for management of the radioactive process waste streams, which processes include collection, concentration, interim storage, calcination to granular solids, and long-term intermediate storage. Over four million gallons of HLLW have been converted to a recoverable granular solid form through waste liquid injection into a high-temperature, fluidized bed wherein the wastes are converted to their respective solid oxides. The development of a glass ceramic solid for the long-term permanent disposal of the high level waste (HLW) solids is also described

  1. Environmental sampling program for a solar evaporation pond for liquid radioactive wastes

    International Nuclear Information System (INIS)

    Romero, R.; Gunderson, T.C.; Talley, A.D.

    1980-04-01

    Los Alamos Scientific Laboratory (LASL) is evaluating solar evaporation as a method for disposal of liquid radioactive wastes. This report describes a sampling program designed to monitor possible escape of radioactivity to the environment from a solar evaporation pond prototype constructed at LASL. Background radioactivity levels at the pond site were determined from soil and vegetation analyses before construction. When the pond is operative, the sampling program will qualitatively and quantitatively detect the transport of radioactivity to the soil, air, and vegetation in the vicinity. Possible correlation of meteorological data with sampling results is being investigated and measures to control export of radioactivity by biological vectors are being assessed

  2. Bottom loaded filter for radioactive liquids

    International Nuclear Information System (INIS)

    Wendland, W.G.

    1983-01-01

    This invention relates to equipment for filtering liquids and more particularly to filter assemblies for use with radioactive by-products of nuclear power plants. The invention provides a compact, bottom-loaded filter assembly that can be quickly and safely loaded and unloaded without the use of complex remote equipment. The assembly is integrally shielded and does not require external shielding. The closure hatch may be automatically aligned to facilitate quick sealing attachment after replacement of the filter cartridge, and the filter cartridge may be automatically positioned within the filter housing during the replacement operation

  3. Biosorption of Am-241 and Cs-137 by radioactive liquid waste by coffee husk

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Rafael Vicente de Padua; Sakata, Solange Kazumi; Bellini, Maria Helena; Marumo, Julio Takehiro, E-mail: jtmarumo@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    Radioactive Waste Management Laboratory of Nuclear and Energy Research Institute, IPEN-CNEN/SP, has stored many types of radioactive liquid wastes, including liquid scintillators, mixed wastes from chemical analysis and spent decontamination solutions. These wastes need special attention, because the available treatment processes are often expensive and difficult to manage. Biosorption using biomass of vegetable using agricultural waste has become a very attractive technique because it involves the removal of heavy metals ions by low cost biossorbents. The aim of this study is to evaluate the potential of the coffee husk to remove Am-241 and Cs-137 from radioactive liquid waste. The coffee husk was tested in two forms, treated and untreated. The chemical treatment of the coffee husk was performed with HNO{sub 3} and NaOH diluted solutions. The results showed that the coffee husk did not showed significant differences in behavior and capacity for biosorption for Am-241 and Cs-137 over time. Coffee husk showed low biosorption capacity for Cs-137, removing only 7.2 {+-} 1.0% in 4 hours of contact time. For Am-241, the maximum biosorption was 57,5 {+-} 0.6% in 1 hours. These results suggest that coffee husk in untreated form can be used in the treatment of radioactive waste liquid containing Am-241. (author)

  4. Retention and subsequent release of radioactivity from the incineration of wastes containing microspheres

    International Nuclear Information System (INIS)

    Emery, R.J.; Watson, J.E. Jr.

    1990-01-01

    Incineration is the preferred method for disposing of animal carcasses containing radioactive microspheres at the authors University. Routine surveys of ash from successive nonradioactive burns revealed significant contamination from previously incinerated microspheres. Past studies on microsphere incineration quantified the amount of activity retained in ash, but did not address any subsequent releases. This topic was not considered in earlier studies because, in most cases, the carcasses were placed in some type of container to facilitate recovery of ash, preventing contamination of the incinerator refractory. In this study, five sets of controlled burns were performed to quantify the subsequent releases of the microsphere radioisotopes 141 Ce, 113 Sn, 102 Ru, 95 Nb, and 46 Sc. Each set consisted of three successive burns. The first burn of each set incinerated a non-radioactive carcass, the second burn, a radioactive carcass, and the third, a non-radioactive carcass. In all of the burns, the carcasses were placed directly on the incinerator refractory floor, which is the standard procedure during normal operations

  5. HABIT, Toxic and Radioactive Release Hazards in Reactor Control Room

    International Nuclear Information System (INIS)

    Stage, S.A.

    2005-01-01

    1 - Description of program or function: HABIT is a package of computer codes designed to be used for the evaluation of control room habitability in the event of an accidental release of toxic chemicals or radioactive materials. 2 - Methods: Given information about the design of a nuclear power plant, a scenario for the release of toxic or radionuclides, and information about the air flows and protection systems of the control room, HABIT can be used to estimate the chemical exposure or radiological dose to control room personnel

  6. Method of processing radioactive cesium liquid wastes

    International Nuclear Information System (INIS)

    Nishijima, Hiroaki; Asaoka, Sachio; Kondo, Tadami; Suzuki, Isao.

    1985-01-01

    Purpose: To convert and settle cesium, mainly, Cs-137 in liquid wastes in the form of pollucites, that is, cesium-containing ores. Constitution: Water, silica, alumina and alkali metal source are mixed with radioactive liquid wastes containing cesium as the main metal element ingredient, to which an onium compound is further added and they are brought into reaction till pollucite ores (Cs 16 (Al 16 Si 32 O 96 )) are formed. Since most portion of cesium is thus settled in the form of pollucites, storage safety can be attained. Further, the addition of the onium compound can moderate the condition and shorten the time till the pollucite ores are formed. The onium compound usable herein includes tetramethyl ammonium. (Kamimura, M.)

  7. Study of the Treatment of the Liquid Radioactive Waste Nong Son Uranium Ore Processing

    International Nuclear Information System (INIS)

    Nguyen Ba Tien; Trinh Giang Huong; Luu Cao Nguyen; Harvey, L.K.; Tran Van Quy

    2011-01-01

    Liquid waste from Nong Son uranium ore processing is treated with concentrated acid, agglomerated, leached, run through ion exchange and then treated with H 2 O 2 to precipitate yellowcake. The liquid radioactive waste has a pH of 1.86 and a high content of radioactive elements, such as: [U] 143.898 ppm and [Th] = 7.967 ppm. In addition, this waste contains many polluted chemical elements with high content, such as arsenic, mercury, aluminum, iron, zinc, magnesium, manganese and nickel. The application of the general method as one stage precipitation or precipitation in coordination with BaCl 2 is not effective. These methods generated a large amount of sludge with poor settling characteristics. The volume of final treated waste was large. This paper introduces the investigation of the treatment of this liquid radioactive waste by the method of two stage of precipitation in association with polyaluminicloride (PAC) and polymer. The impact of factors: pH, neutralizing agents, quantity of PAC and polymer to effect precipitation and improve the settling characteristics during processing was studied. The results showed that the processing of liquid radioactive waste treatment through two stages: first stage at pH = 3 and the second stage at pH = 8.0 with limited PAC and polymer (A 101) resulted in significant reduced volume of the treated waste. The discharged liquid satisfied the requirement of the National Technical Regulation on Industrial Waste Water (QCVN 24:2009). (author)

  8. Development of monitor for multiple beta-ray nuclides in liquid radioactive waste (IV)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Nam, Uk Won; Seon, K. I.; Kong, K. N.; Son, J. D.; Chin, H.; Park, J. H.; Yuk, I. S.; Han, W. Y. [Korea Astronomy and Space Science Institute, Daejeon (Korea, Republic of)

    2004-02-15

    In the 4{sup th} fiscal year, the performance of the system developed in previous year have been stablilized. An automatic mixer system to mix the liquid radioactive waste with the liquid scintillator was designed and manufactured to make continuous monitoring possible. Mechanical systems for the automatic transfer, injection of radioactive samples, and for sealing sample bottle and its attachment on detector system have been designed and developed for completely integrated system, The experimental has been performed for the radioactive analysis by using the system. The measured data are transferred to central computer, which controls the monitoring system, through the RS-232 communication protocol. A software for host personal computer was developed for the control and analysis of the integrated system.

  9. Nuclear refugees after large radioactive releases

    International Nuclear Information System (INIS)

    Pascucci-Cahen, Ludivine; Groell, Jérôme

    2016-01-01

    However improbable, large radioactive releases from a nuclear power plant would entail major consequences for the surrounding population. In Fukushima, 80,000 people had to evacuate the most contaminated areas around the NPP for a prolonged period of time. These people have been called “nuclear refugees”. The paper first argues that the number of nuclear refugees is a better measure of the severity of radiological consequences than the number of fatalities, although the latter is widely used to assess other catastrophic events such as earthquakes or tsunami. It is a valuable partial indicator in the context of comprehensive studies of overall consequences. Section 2 makes a clear distinction between long-term relocation and emergency evacuation and proposes a method to estimate the number of refugees. Section 3 examines the distribution of nuclear refugees with respect to weather and release site. The distribution is asymmetric and fat-tailed: unfavorable weather can lead to the contamination of large areas of land; large cities have in turn a higher probability of being contaminated. - Highlights: • Number of refugees is a good indicator of the severity of radiological consequences. • It is a better measure of the long-term consequences than the number of fatalities. • A representative meteorological sample should be sufficiently large. • The number of refugees highly depends on the release site in a country like France.

  10. Portable Filtered Air Suction System for Released Radioactive Gases Prevention under a Severe Accident of NPPs

    International Nuclear Information System (INIS)

    Gu, Beom W.; Choi, Su Y.; Rim, Chun T.

    2013-01-01

    In this paper, the portable filtered air suction system (PoFASS) for released radioactive gases prevention under a severe accident of NPP is proposed. This technology can prevent the release of the radioactive gases to the atmosphere and it can be more economical than FVCS because PoFASS can cover many NPPs with its high mobility. The conceptual design of PoFASS, which has the highest cost effectiveness and robustness to the environment condition such as wind velocity and precipitation, is suggested and the related previous research is introduced in this paper. The portable filtered air suction system (PoFASS) for released radioactive gases prevention can play a key role to mitigate the severe accident of NPP with its high cost effectiveness and robustness to the environment conditions. As further works, the detail design of PoFASS to fabricate a prototype for a demonstration will be proceeded. When released radioactive gases from the broken containment building in the severe accident of nuclear power plants (NPPs) such as the Chernobyl and Fukushima accidents occur, there are no ways to prevent the released radioactive gases spreading in the air. In order to solve this problem, several European NPPs have adopted the filtered vented containment system (FVCS), which can avoid the containment failure through a pressure relief capability to protect the containment building against overpressure. However, the installation cost of FVCS for a NPP is more than $10 million and this system has not been widely welcomed by NPP operating companies due to its high cost

  11. Freedom: a transient fission-product release model for radioactive and stable species

    International Nuclear Information System (INIS)

    Macdonald, L.D.; Lewis, B.J.; Iglesias, F.C.

    1989-05-01

    A microstructure-dependent fission-gas release and swelling model (FREEDOM) has been developed for UO 2 fuel. The model describes the transient release behaviour for both the radioactive and stable fission-product species. The model can be applied over the full range of operating conditions, as well as for accident conditions that result in high fuel temperatures. The model accounts for lattice diffusion and grain-boundary sweeping of fusion products to the grain boundaries, where the fission gases accumulate in grain-face bubbles as a result of vacancy diffusion. Release of fission-gas to the free void of the fuel element occurs through the interlinkage of bubbles and cracks on the grain boundaries. This treatment also accounts for radioactive chain decay and neutron-induced transmutation effects. These phenomena are described by mass balance equations which are numerically solved using a moving-boundary, finite-element method with mesh refinement. The effects of grain-face bubbles on fuel swelling and fuel thermal conductivity are included in the ELESIM fuel performance code. FREEDOM has an accuracy of better than 1% when assessed against an analytic solution for diffusional release. The code is being evaluated against a fuel performance database for stable gas release, and against sweep-gas and in-cell fission-product release experiments at Chalk River for active species

  12. Underground Nuclear Explosions and Release of Radioactive Noble Gases

    Science.gov (United States)

    Dubasov, Yuri V.

    2010-05-01

    Over a period in 1961-1990 496 underground nuclear tests and explosions of different purpose and in different rocks were conducted in the Soviet Union at Semipalatinsk and anovaya Zemlya Test Sites. A total of 340 underground nuclear tests were conducted at the Semipalatinsk Test Site. One hundred seventy-nine explosions (52.6%) among them were classified as these of complete containment, 145 explosions (42.6%) as explosions with weak release of radioactive noble gases (RNG), 12 explosions (3.5%) as explosions with nonstandard radiation situation, and four excavation explosions with ground ejection (1.1%). Thirty-nine nuclear tests had been conducted at the Novaya Zemlya Test Site; six of them - in shafts. In 14 tests (36%) there were no RNG release. Twenty-three tests have been accompanied by RNG release into the atmosphere without sedimental contamination. Nonstandard radiation situation occurred in two tests. In incomplete containment explosions both early-time RNG release (up to ~1 h) and late-time release from 1 to 28 h after the explosion were observed. Sometimes gas release took place for several days, and it occurred either through tunnel portal or epicentral zone, depending on atmospheric air temperature.

  13. Removal of radioactive cesium from soil by ammonium citrate solution and ionic liquid

    International Nuclear Information System (INIS)

    Ishiwata, Shunji; Kitakouji, Manabu; Taga, Atsushi; Ogata, Fumihiko; Ouchi, Hidekazu; Yamanishi, Hirokuni; Inagaki, Masayo

    2015-01-01

    Radioactive cesium has strongly bound soil as time proceeded, which could not be cleaved in mild condition. We have found that serial treatment of ammonium citrate solution and ionic liquid removed radioactive cesium from soil effectively. The sequence of the treatment is crucial, since inverse serial treatment or mixture of two kinds of solution did not show such an effect, which suggested that ammonium citrate unlocked trapped cesium in soil and ionic liquid solved it. We also found that repeating serial treatment and prolonged treatment time additively removed cesium from soil. (author)

  14. Cask for radioactive material and method for preventing release of neutrons from radioactive material

    International Nuclear Information System (INIS)

    Gaffney, M.F.; Shaffer, P.T.

    1981-01-01

    A cask for radioactive material, such as nuclear reactor fuel or spent nuclear reactor fuel, includes a plurality of associated walled internal compartments for containing such radioactive material, with neutron absorbing material present to absorb neutrons emitted by the radioactive material, and a plurality of thermally conductive members, such as longitudinal copper or aluminum castings, about the compartment and in thermal contact with the compartment walls and with other such thermally conductive members and having thermal contact surfaces between such members extending, preferably radially, from the compartment walls to external surfaces of the thermally conductive members, which surfaces are preferably in the form of a cylinder. The ends of the shipping cask also preferably include a neutron absorber and a conductive metal covering to dissipate heat released by decay of the radioactive material. A preferred neutron absorber utilized is boron carbide, preferably as plasma sprayed with metal powder or as particles in a matrix of phenolic polymer, and the compartment walls are preferably of stainless steel, copper or other corrosion resistant and heat conductive metal or alloy. The invention also relates to shipping casks, storage casks and other containers for radioactive materials in which a plurality of internal compartments for such material, e.g., nuclear reactor fuel rods, are joined together, preferably in modular construction with surrounding heat conductive metal members, and the modules are joined together to form a major part of a finished shipping cask, which is preferably of cylindrical shape. Also within the invention are methods of safely storing radioactive materials which emit neutrons, while dissipating the heat thereof, and of manufacturing the present shipping casks

  15. Individual dose due to radioactivity accidental release from fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nie, Baojie [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Ni, Muyi, E-mail: muyi.ni@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Wei, Shiping [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China)

    2017-04-05

    Highlights: • Conservative early dose of different unit fusion radioactivity release were assessed. • Data of accident level in INES for fusion reactor were proposed. • Method of environmental restoration time after fusion accident was proposed. • The maximum possible accident level for ITER like fusion reactor is 6. • We need 34–52 years to live after the fusion hypothetical accident. - Abstract: As an important index shaping the design of fusion safety system, evaluation of public radiation consequences have risen as a hot topic on the way to develop fusion energy. In this work, the comprehensive public early dose was evaluated due to unit gram tritium (HT/HTO), activated dust, activated corrosion products (ACPs) and activated gases accidental release from ITER like fusion reactor. Meanwhile, considering that we cannot completely eliminate the occurrence likelihood of multi-failure of vacuum vessel and tokamak building, we conservatively evaluated the public radiation consequences and environment restoration after the worst hypothetical accident preliminarily. The comparison results show early dose of different unit radioactivity release under different conditions. After further performing the radiation consequences, we find it possible that the hypothetical accident for ITER like fusion reactor would result in a level 6 accident according to INES, not appear level 7 like Chernobyl or Fukushima accidents. And from the point of environment restoration, we need at least 69 years for case 1 (1 kg HTO and 1000 kg dust release) and 34–52 years for case 2 (1 kg HTO and 10kg–100 kg dust release) to wait the contaminated zone drop below the general public safety limit (1mSv per year) before it is suitable for human habitation.

  16. Summary 1998. Releases control and environment monitoring for the CEA Centers

    International Nuclear Information System (INIS)

    1998-01-01

    In the framework of its environmental policy, the CEA aims at reduce as weak as possible, in regards to the technological and economic needs, its activities impacts on the people and the environment. This paper contributes to the public information on the radioactive gaseous and liquid releases during the year 1998. It presents data on the releases and the radioactivity levels around the CEA sites and gathers the associated regulation and monitoring methods. (A.L.B.)

  17. Characterization of the solid radioactive waste From Cernavoda NPP

    International Nuclear Information System (INIS)

    Iordache, M.; Laotaru, V.

    2005-01-01

    Full text: During the operation of a nuclear plant significant quantities of radioactive waste result that have a very large diversity. At Cernavoda NPP large amounts of wastes are either non-radioactive wastes or radioactive wastes, each of these being managed completely different from which other. For a CANDU type reactor, the appearance of radioactive wastes is due to contamination with the following types of radioactive substances: - fission products resulting from nuclear fuel burning; - activated products from materials composing the technological systems; - activated products in process fluids. Radioactive wastes can be in solid, liquid or gas form. At Cernavoda NPP the solid wastes represent about 70% of the waste volume which is produced during plant operation and as a consequence of maintenance and decontamination operations. The most important types of solid wastes that are obtained and then handled, processed (if necessary) and temporarily stored are: solid low-level radioactive wastes (classified as compactible and non-compactible), solid medium radioactive wastes, spent resins, used filters and filter cartridges. The liquid radioactive waste class includes organic liquids (used oil, scintillator liquids and used solvents) and aqueous wastes resulting from process system operating, from decontamination and maintenance operations. Radioactive gas wastes occur subsequently to the fission process inside the fuel elements as well as due to the neutron activation of process fluids in the reactor systems. As result of plant operation, iodine, noble gases, tritium and radioactive particles occur and are passed toward the ventilation stack in a controlled manner so that environmental release of radioactive materials with concentrations exceeding the maximum permissible level could not occur. (authors)

  18. Method of treating the waste liquid of a washing containing a radioactive substance

    International Nuclear Information System (INIS)

    Sawaguchi, Yusuke; Tsuyuki, Takashi; Kaneko, Masato; Sato, Yasuhiko; Yamaguchi, Takashi.

    1975-01-01

    Object: To separate waste liquid resulting from washing and which contains a radioactive substance and surface active agent into high purity water and a solid waste substance containing a small quantity of surface active agent. Structure: To waste liquid from a waste liquid tank is added a pH adjusting agent for adjusting the pH to 5.5, and the resultant liquid is sent to an agglomeration reaction tank, in which an inorganic agglomerating agent is added to the waste liquid to cause a major proportion of the radioactive substance and surface active agent to form flocks produced through agglomeration. Then, the waste liquid is sent from the agglomeration reaction tank to a froth separation tank, to which air is supplied through a perforated plate to cause frothing. The over-flowing liquid is de-frothed, and then the insoluble matter is separated as sludge, followed by hydroextraction and drying for solidification. The treated liquid extracted from a froth separation tank is sent to an agglomerating agent recovery tank for separation of the agglomeration agent, and then the residual surface active agent is removed by adsorption in an active carbon adsorption tower, followed by concentration by evaporation in an evaporating can. The concentrated liquid is extracted and then solidified with cement or asphalt. (Kamimura, M.)

  19. Liquid radioactive waste processing system in Improved OPR-1000

    International Nuclear Information System (INIS)

    Lee, Soonmin; Kim, Kiljung; Park, Jungsu

    2008-01-01

    The design goal of liquid rad waste system is to minimize the release of radioactive materials to the environment, the occupational radiation exposure to workers, and the solid rad waste volume generated from LRS operation. In 1998, KOPEC in conjunction with KHNP (Korea Hydro and Nuclear Power Co.) started a special task study which had been focused on the worldwide advanced technologies in the liquid rad waste process area by considering the design goals above. As a result of this task, KOPEC and KHNP finally decided to adopt a reverse osmosis processing method for Improved OPR-1000 in Korea. The advanced LRS design incorporating the R/O process has been introduced into Shin-Wolsong 1 and 2 (SWN 1 and 2) as well as Shin-Kori 1 and 2 (SKN 1 and 2), which are recently under construction, and also is adopted for Shin-Kori 3 and 4 (SKN 3 and 4) and Shin-Ulchin 1 and 2 (SUN 1 and 2), which are planned for the near future construction as the first APR-1400 type of Korean reactors. The LRS shop performance test for SKN 1 and 2 (Improved OPR-1000 R/O package system) was conducted by DOOSAN and DTS (Diversified Technologies Services, Inc) in January, 2008. The purpose of the test was to demonstrate the performance of actual R/O system to be installed in SKN 1 and 2 site. In this paper, overall system configuration and the shop performance test result is presented based on Improved OPR-1000 LRS R/O Package system

  20. Biosorption of Am-241 and Cs-137 by radioactive liquid waste by coffee husk

    International Nuclear Information System (INIS)

    Ferreira, Rafael Vicente de Padua; Sakata, Solange Kazumi; Bellini, Maria Helena; Marumo, Julio Takehiro

    2011-01-01

    Radioactive Waste Management Laboratory of Nuclear and Energy Research Institute, IPEN-CNEN/SP, has stored many types of radioactive liquid wastes, including liquid scintillators, mixed wastes from chemical analysis and spent decontamination solutions. These wastes need special attention, because the available treatment processes are often expensive and difficult to manage. Biosorption using biomass of vegetable using agricultural waste has become a very attractive technique because it involves the removal of heavy metals ions by low cost biossorbents. The aim of this study is to evaluate the potential of the coffee husk to remove Am-241 and Cs-137 from radioactive liquid waste. The coffee husk was tested in two forms, treated and untreated. The chemical treatment of the coffee husk was performed with HNO 3 and NaOH diluted solutions. The results showed that the coffee husk did not showed significant differences in behavior and capacity for biosorption for Am-241 and Cs-137 over time. Coffee husk showed low biosorption capacity for Cs-137, removing only 7.2 ± 1.0% in 4 hours of contact time. For Am-241, the maximum biosorption was 57,5 ± 0.6% in 1 hours. These results suggest that coffee husk in untreated form can be used in the treatment of radioactive waste liquid containing Am-241. (author)

  1. Radioactive iodine releases from nuclear power plant, (1)

    International Nuclear Information System (INIS)

    Hashimoto, Tatsuya

    1974-01-01

    Concerning the release of radioactive iodine from nuclear power plants, the guidelines and data both in Japan and abroad are briefed, including those in the United States, Tsuruga nuclear power station and working Group of the Environmental Radiation Study Committee. In case of the Tsuruga nuclear power station, the radiation dose and other data for a few years are presented. Parameters and factors proposed by the working group cover such as the dose through food intake and respiration, concentration factor, etc. (Mori, K.)

  2. Combined methods for liquid radioactive waste treatment. Final report of a co-ordinated research project 1997-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-02-01

    The report contains 13 papers presented at the final research co-ordination meeting of the CRP. The subjects covered include processes and technologies for treatment and conditioning of liquid radioactive wastes. It quite often includes the application of several steps, such as filtration, precipitation, sorption, ion exchange, evaporation and/or membrane separation to meet the requirements both for the release of decontaminated effluents into the environment and the conditioning of waste concentrates for disposal. Combination of the processes and their consecutive or simultaneous application is also described. It results in an improved decontamination, waste volume reduction, safety and overall cost effectiveness in the treatment, conditioning and disposal of these wastes.

  3. Combined methods for liquid radioactive waste treatment. Final report of a co-ordinated research project 1997-2001

    International Nuclear Information System (INIS)

    2003-02-01

    The report contains 13 papers presented at the final research co-ordination meeting of the CRP. The subjects covered include processes and technologies for treatment and conditioning of liquid radioactive wastes. It quite often includes the application of several steps, such as filtration, precipitation, sorption, ion exchange, evaporation and/or membrane separation to meet the requirements both for the release of decontaminated effluents into the environment and the conditioning of waste concentrates for disposal. Combination of the processes and their consecutive or simultaneous application is also described. It results in an improved decontamination, waste volume reduction, safety and overall cost effectiveness in the treatment, conditioning and disposal of these wastes

  4. Method of solidifying liquid radioactive wastes

    International Nuclear Information System (INIS)

    Pekar, A.; Petrovic, J.; Timulak, J.

    1987-01-01

    Liquid radioactive waste containing boric acid salts is mixed with zeolite tuff and neutralized by lime. Power plant fly ash containing single-component or mixed Portland cement is then added to the mixture. Prior to packaging, anion-active bitumen emulsion or an aqueous emulsion of fatty acid salts and of free fatty acids insoluble in water can be added. Examples are given listing accurate proportions of the individual components. The advantage of the said solidification method is the use of easily available raw materials and improved values of extractability of the resulting product radionuclides. (E.S.)

  5. Analysis of a Radioactive Release in a Nuclear Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Poppiti, James [Dept. of Energy, Washington, DC (United States); Nelson, Roger [Dept. of Energy, Carlsbad, NM (United States); MacMillan, Walter J. [Nuclear Waste Partners, Carlsbad, NM (United States); Cunningham, Scott

    2017-07-01

    The Waste Isolation Pilot Plant (WIPP) is a 655-meter deep mine near Carlsbad, New Mexico, used to dispose the nation’s defense transuranic waste. Limited airborne radioactivity was released from a container of radioactive waste in WIPP on 14 February, 2014. As designed, a mine ventilation filtration system prevented the large scale release of contamination from the underground. However, isolation dampers leaked, which allowed the release of low levels of contaminants after the event until they were sealed. None of the exposed individuals received any recordable dose. While surface contamination was limited, contamination in the ventilation system and portions of the underground was substantial. High efficiency particulate air (HEPA) filters in the operating ventilation system ensure continued containment during recovery and resumption of disposal operations. However, ventilation flow is restricted since the incident, with all exhaust air directed through the filters. Decontamination and natural fixation by the hygroscopic nature of the salt host rock has reduced the likelihood of further contamination spread. Contamination control and ventilation system operability are crucial for resumption of operations. This article provides an operational assessment and evaluation of these two key areas.

  6. Analysis of a Radioactive Release in a Nuclear Waste Disposal Facility

    International Nuclear Information System (INIS)

    Poppiti, James; Nelson, Roger; MacMillan, Walter J.; Cunningham, Scott

    2017-01-01

    The Waste Isolation Pilot Plant (WIPP) is a 655-meter deep mine near Carlsbad, New Mexico, used to dispose the nation's defense transuranic waste. Limited airborne radioactivity was released from a container of radioactive waste in WIPP on 14 February, 2014. As designed, a mine ventilation filtration system prevented the large scale release of contamination from the underground. However, isolation dampers leaked, which allowed the release of low levels of contaminants after the event until they were sealed. None of the exposed individuals received any recordable dose. While surface contamination was limited, contamination in the ventilation system and portions of the underground was substantial. High efficiency particulate air (HEPA) filters in the operating ventilation system ensure continued containment during recovery and resumption of disposal operations. However, ventilation flow is restricted since the incident, with all exhaust air directed through the filters. Decontamination and natural fixation by the hygroscopic nature of the salt host rock has reduced the likelihood of further contamination spread. Contamination control and ventilation system operability are crucial for resumption of operations. This article provides an operational assessment and evaluation of these two key areas.

  7. Device for concentrating radioactive liquid wastes

    International Nuclear Information System (INIS)

    Adachi, Takuji; Uchiyama, Yoshio; Ukaji, Hideo.

    1981-01-01

    Purpose: To prevent the heat-transfer surface of a heat-transfer tube from adhering scale. Constitution: A differential-pressure generator is provided between a heater and an evaporator in order to make the vapor pressure at the heater side higher than that at the evaporator side. Pressure detectors are installed at the heating can outlet and at the evaporating can inlet. The detected pressure is converted to a signal, which is applied to a flow rate regulator, and so differential pressure production valve is operated. Thus, it can prevent the formation of a liquid lost region due to the evaporation under the pressure-decrease at the heating can side during the concentrating operation of the radioactive liquid waste, and also prevents the corrosion or explosion of the heat transfer tube due to the deposition of scale even if temperature of the heat transfer surface of the heat transfer tube is abnormally increased. (J.P.N.)

  8. Safety disposal studies of radioactive and hazardous wastes using cement

    International Nuclear Information System (INIS)

    Aly, M.M.E.

    2000-01-01

    radioactive waste is generated from the production of nuclear energy and from the use of radioactive materials applications, agriculture and medicine. the important of safe management of radioactive waste for the protection of human health and the environment has long been recognized. conditioning of radioactive waste is the transform of radioactive waste into a suitable form for storage and disposal. common immobilization methods include solidification of low radioactive waste in cement or bitumen.in order to improve cement properties to decrease the release of liquid radioactive waste into the environment and its dispersion to a level where the risks to individuals, population and the environment

  9. Calculation of releases of radioactive materials in gaseous effluents from nuclear-powered merchant ships (NMS-GEFF code)

    International Nuclear Information System (INIS)

    Cardile, F.P.; Bangart, R.L.; Collins, J.T.

    1978-06-01

    The Intergovernmental Maritime Consultative Organization IMCO) is currently preparing guidelines concerning the safety of nuclear-powered merchant ships. An important aspect of these guidelines is the determination of the releases of radioactive material in effluents from these ships and the control exercised by the ships over these releases. To provide a method for the determination of these releases, the NRC staff has developed a computerized model, the NMS-GEFF Code, which is described in the following chapters. The NMS-GEFF Code calculates releases of radioactive material in gaseous effluents for nuclear-powered merchant ships using pressurized water reactors

  10. Implementation of a management applied program for liquid radioactive waste treatment

    International Nuclear Information System (INIS)

    Lee, Y. H.; Ann, S. J.; Jo, H. S.; Son, J. S.

    2003-01-01

    A data collection of a liquid radioactive waste treatment process of a research organization became necessary while developing the RAWMIS(Radioactive Waste Management Integration System) which it can generate personal history management for efficient management of a waste, documents, all kinds of statistics. This paper introduces an input and output application program design to do to database with data in the results and a stream process of a treatment that analyzed the waste occurrence present situation and data by treatment process. Data on the actual treatment process that is not limited experiment improve by a document, human traces, saving of material resources and improve with efficiency of tracking about a radioactive waste and a process and give help to radioactive waste material valance and inventory study

  11. RESUS: A code for low volatile radio-nuclide release from liquids due to vapor bubble burst induced liquid jet formation and disintegration

    International Nuclear Information System (INIS)

    Koch, M.K.; Starflinger, J.; Linnemann, Th.; Brockmeier, U.; Unger, H.; Schuetz, W.

    1995-01-01

    In the field of nuclear safety, the release of volatile and low volatile radio-nuclides from liquid surfaces into a gas atmosphere is important for aerosol source term considerations particularly in late severe accident sequences. In case of a hypothetical nuclear reactor accident involving a failure of the primary system, primary coolant and radio-nuclides may be released into the containment to frequently form a liquid pool which may be contaminated by suspended or solved fuel particles and fission products. Under this scope, the release code package REVOLS/RENONS was developed for radio-nuclide release from liquid surfaces. Assuming the absence of gas or vapor bubbles in the liquid, the evaporative release of volatile components, calculated by the REVOLS code, is governed by diffusive and convective transport processes, whereas the release of low volatiles, calculated by the RENONS code, may be governed by mechanical processes which leads to droplet entrainment in case of wavy liquid pool surface conditions into the containment atmosphere by means of convection. For many accident sequences, in which gas is injected into a pool or liquid area elsewhere, predominantly when saturation temperatures can be reached, the release of low volatile species from liquid surfaces due to bubble burst is identified as a decisive release mechanism also. Together with the liquid, the particles which are located at the pool surface or suspended in the pool, are released into the atmosphere. Consequently, the code RESUS.MOD1 (RESUSpension) is presently extended to include the calculation of the release of droplets and suspended radio-nuclide particles due to bubble burst induced liquid jet formation and disintegration above liquid surfaces. Experimental investigations indicate the influence of bubble volume and shape at the pool surface as well as bubble stabilization or destabilization, and furthermore the system pressure and temperatures as well as fluid properties, on droplet

  12. Conditioning the radioactive oils released from the Cernavoda NPP operation

    International Nuclear Information System (INIS)

    Popescu, I.; Deneanu, N.; Dulama, M.; Baboiescu, E.

    2001-01-01

    The radioactive oils released during the Cernavoda NPP operation are lubricating oils used in the primary circuit pumps, hydraulic oils used in the reactor charging machine, as well as, waste turbine oils. The primary cooling circuit is the main way of contamination. This waste contain variable amounts of tritium of activities up to 10 8 Bq/l and relatively small amounts of β/γ emitting radionuclides. The radioactive oils resulting of the heat transport system may contain as much as 1.8 TBq/l of tritium. At INR Pitesti, Department of Radioactive Waste Processing, studies were done concerning the radioactive oils conditioning in concrete. It was found that the minimal compression level required for disposal, of 50 MPa, was exceeded. As consequence, the conditioned waste fulfils the conditions of acceptance provided by the Radioactive Waste National Repository at Baita-Bihor. The main stages of the process are: - dosing the radioactive oil, water and NOFOX9 and NOFOX4 emulsion additives for the 220 l or 80 l barrel; - mixing these ingredients up to reaching the emulsion state; - dosing the cement and lime; - adding the enhancing silicate and shaking the mixture to get a homogeneous matrix; - closing the barrel; - conditioning the 80 l barrel within the 220 l container. The production capacity is of 5 to 8 barrels/shift

  13. Environmental radioactivity at the National Nuclear Research Centre, Pelindaba

    International Nuclear Information System (INIS)

    Brits, R.J.N.; Van der Westhuizen, G.S.H.; Annandale, J.

    1983-06-01

    The revised environmental survey program, introduced during 1970 with the emphasis on monitoring of the critical paths of exposure of the general public, was continued in 1982. Results of determinations of both gross radioactivity and individual nuclides in samples of fish and water (which are critical materials for liquid-effluent releases) from the Hartebeespoort Dam and from the Crocodile River, are given and discussed. Results of 131 I, 90 Sr and gamma-spectrometric analyses of milk, the critical material for releases to the atmosphere, are presented. Results are given of regular investigations of the composition of airborne releases to the atmosphere and liquid-effluent releases to the Crocodile River, performed in order to detect other possible critical nuclides. Levels of deposited and airborne activity from nuclear-bomb tests are reported. Due to absence of fresh fallout material the levels for most fission products have fallen below the limit of detection. No environmental radioactivity due to releases from the Pelindaba site could be detected above the natural background or accumulated fallout levels. Unplanned releases of UF 6 occur sometimes. Accordingly, some of the environmental samples were also analysed for uranium. The results obtained so far do not indicate an increase in uranium levels in the environment

  14. Processing method and processing device for liquid waste containing surface active agent and radioactive material

    International Nuclear Information System (INIS)

    Nishi, Takashi; Matsuda, Masami; Baba, Tsutomu; Yoshikawa, Ryozo; Yukita, Atsushi.

    1998-01-01

    Washing liquid wastes containing surface active agents and radioactive materials are sent to a deaerating vessel. Ozone is blown into the deaerating vessel. The washing liquid wastes dissolved with ozone are introduced to a UV ray irradiation vessel. UV rays are irradiated to the washing liquid wastes, and hydroxy radicals generated by photodecomposition of dissolved ozone oxidatively decompose surface active agents contained in the washing liquid wastes. The washing liquid wastes discharged from the UV ray irradiation vessel are sent to an activated carbon mixing vessel and mixed with powdery activated carbon. The surface active agents not decomposed in the UV ray irradiation vessel are adsorbed to the activated carbon. Then, the activated carbon and washing liquid wastes are separated by an activated carbon separating/drying device. Radioactive materials (iron oxide and the like) contained in the washing liquid wastes are mostly granular, and they are separated and removed from the washing liquid wastes in the activated carbon separating/drying device. (I.N.)

  15. Radioactive waste generation in the nuclear reactors in Romania

    International Nuclear Information System (INIS)

    Popescu, I.V.

    2002-01-01

    The successful use of nuclear fission as major source of energy for this century is based upon the technological capabilities acquired to face the issue of radioactive waste and spent fuel. The management of radioactive waste is complex and implies solving the following major problems: - isolation of the radioisotopes from the complex of effluents released in the environment; - processing the separated radioisotopes for subsequent storing and final disposal; - transport of processed and conditioned wastes towards disposal repository; - selecting the sites for storage and final disposal. During reactor operation liquid and gaseous effluents are released to the environment as well as radioactive materials. All these may have an dangerous impact upon the environment when the international regulations, i.e. the ALARA principle are not strictly observed. The maximal values for the radioactive release are established by national regulations which are concordant with the IAEA principles. The amount of radioactive materials released depends of the reactor type and the measures adopted to reduce these releases. The average values of these releases during the normal operation of the reactor constitute the 'source term'. Its calculation implies several factors such as: the reactor type; the radionuclide concentration in the primary cooling systems; the transport mechanisms and leaks resulting in liquid and gaseous radionuclide emissions; the efficiency of the barriers and engineered safety systems built to reduce the amounts of radionuclide in the effluents. The concentration of radionuclides in the primary cooling circuit depends on the reactor power level, fuel burnup, fuel sheath type, tightness of the fuel cans, impurity concentration, chemical additives in the fluid of the primary cooling system, the total volume of this fluid, as well as its purification system. The methods applied to facilitate the calculation of the source term are described. In 1998 the spent fuel

  16. Air modelling as an alternative to sampling for low-level radioactive airborne releases

    International Nuclear Information System (INIS)

    Morgenstern, M.Y.; Hueske, K.

    1995-01-01

    This paper describes our efforts to assess the effect of airborne releases at one DOE laboratory using air modelling based on historical data. Among the facilities affected by these developments is Los Alamos National Laboratory (LANL) in New Mexico. RCRA, as amended by the Hazardous and Solid Waste Amendments (HSWA) in 1984, requires all facilities which involve the treatment, storage, and disposal of hazardous waste obtain a RCRA/HSWA waste facility permit. LANL complied with CEARP by initiating a process of identifying potential release sites associated with LANL operations prior to filing a RCRA/HSWA permit application. In the process of preparing the RCRA/HSWA waste facility permit application to the U.S. Environmental Protection Agency (EPA), a total of 603 Solid Waste Management Units (SWMUs) were identified as part of the requirements of the HSWA Module VIH permit requirements. The HSWA Module VIII permit requires LANL to determine whether there have been any releases of hazardous waste or hazardous constituents from SWMUs at the facility dating from the 1940's by performing a RCRA Facility Investigation to address known or suspected releases from specified SWMUs to affected media (i.e. soil, groundwater, surface water, and air). Among the most troublesome of the potential releases sites are those associated with airborne radioactive releases. In order to assess health risks associated with radioactive contaminants in a manner consistent with exposure standards currently in place, the DOE and LANL have established Screening Action Levels (SALs) for radioactive soil contamination. The SALs for each radionuclide in soil are derived from calculations based on a residential scenario in which individuals are exposed to contaminated soil via inhalation and ingestion as well as external exposure to gamma emitters in the soil. The applicable SALs are shown

  17. The release code package REVOLS/RENONS for fission product release from a liquid sodium pool into an inert gas atmosphere

    International Nuclear Information System (INIS)

    Starflinger, J.; Scholtyssek, W.; Unger, H.

    1994-12-01

    For aerosol source term considerations in the field of nuclear safety, the investigation of the release of volatile and non-volatile species from liquid surfaces into a gas atmosphere is important. In case of a hypothetical liquid metal fast breeder reactor accident with tank failure, primary coolant sodium with suspended or solved fuel particles and fission products may be released into the containment. The computer code package REVOLS/RENONS, based on a theoretical mechanistic model with a modular structure, has been developed for the prediction of sodium release as well as volatile and non-volatile radionuclide release from a liquid pool surface into the inert gas atmosphere of the inner containment. Hereby the release of sodium and volatile fission products, like cesium and sodium iodide, is calculated using a theoretical model in a mass transfer coefficient formulation. This model has been transposed into the code version REVOLS.MOD1.1, which is discussed here. It enables parameter analysis under highly variable user-defined boundary conditions. Whereas the evaporative release of the volatile components is governed by diffusive and convective transport processes, the release of the non-volatile ones may be governed by mechanical processes which lead to droplet entrainment from the wavy pool surface under conditions of natural or forced convection into the atmosphere. The mechanistic model calculates the liquid entrainment rate of the non-volatile species, like the fission product strontium oxide and the fuel (uranium dioxide) from a liquid pool surface into a parallel gas flow. The mechanistic model has been transposed into the computer code package REVOLS/RENONS, which is discussed here. Hereby the module REVOLS (RElease of VOLatile Species) calculates the evaporative release of the volatile species, while the module RENONS (RElease of NON-Volatile Species) computes the entrainment release of the non-volatile radionuclides. (orig./HP) [de

  18. Siting considerations for radioactivity in reactor effluents during normal operation

    International Nuclear Information System (INIS)

    Graf, J.M.; Strom, P.O.

    1975-01-01

    In selecting a proper site for a nuclear power station, the consideration of radioactivity released in effluents can be handled in a straightforward manner using the U. S. Atomic Energy Commission's proposed Appendix I to 10 CFR 50, which gives numerical guidelines for design objectives for meeting the criterion ''as low as practicable'' for radioactive material in light-water-cooled nuclear power reactor effluents. By relating the release of radioactive material, the site meteorological conditions, and site boundary distance through appropriate dose models, the suitability of a given site can be determined. ''Rules of thumb'' for comparing anticipated releases to design objectives can be constructed for rapid assessment using the maximum permissible concentration values of 10 CFR 20 as dose factors. These rules of thumb tend to underpredict the allowed releases except in the case of radiocesium in liquids. For gaseous releases, these rules of thumb can be made up in convenient nomogram form for a quick assessment of allowed releases based on local site meteorological conditions. (U.S.)

  19. Biochemical process of low level radioactive liquid simulation waste containing detergent

    International Nuclear Information System (INIS)

    Kundari, Noor Anis; Putra, Sugili; Mukaromah, Umi

    2015-01-01

    Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive element in the waste was thorium with activity of 5.10 −5 Ci/m 3 . The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod’s model and the decreasing of COD and BOD were first order with the rate constant of 0.01 hour −1

  20. Biochemical process of low level radioactive liquid simulation waste containing detergent

    Energy Technology Data Exchange (ETDEWEB)

    Kundari, Noor Anis, E-mail: nooranis@batan.go.id; Putra, Sugili; Mukaromah, Umi [Sekolah Tinggi Teknologi Nuklir – Badan Tenaga Nuklir Nasional Jl. Babarsari P.O. BOX 6101 YKBB Yogyakarta 55281 Telp : (0274) 48085, 489716, Fax : (0274) 489715 (Indonesia)

    2015-12-29

    Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive element in the waste was thorium with activity of 5.10{sup −5} Ci/m{sup 3}. The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod’s model and the decreasing of COD and BOD were first order with the rate constant of 0

  1. Biochemical process of low level radioactive liquid simulation waste containing detergent

    Science.gov (United States)

    Kundari, Noor Anis; Putra, Sugili; Mukaromah, Umi

    2015-12-01

    Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive element in the waste was thorium with activity of 5.10-5 Ci/m3. The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod's model and the decreasing of COD and BOD were first order with the rate constant of 0.01 hour-1.

  2. Method to prepare essentially organic waste liquids containing radioactive or toxic materials

    International Nuclear Information System (INIS)

    Baehr, W.; Drobnik, S.H.; Hild, W.; Kroebel, R.; Meyer, A.; Naumann, G.

    1976-01-01

    Waste solutions occuring in nuclear technology containing radioactive or toxic materials can be solidified by mixing with a polymerisable mixture with subsequent polymerization. An improvement of this method, especially for liquids in which the radioactive components are present as organic compounds is achieved by adding a mixture of at least one monomeric vinyl compound, at least one polyvinyl compound and appropriate catalysts and by polymerizing at temperatures between 15 and 150 0 C. Should the waste liquid contain mineral acid, this is first neutralized by the addition of CaO or MgO. In processing oils or soaps, the addition of swelling agent for polystyrol resins is advantageous. 16 examples illustrate the invention. (UWI) [de

  3. Assay for dihydroorotase using high-performance liquid chromatography with radioactivity detection

    International Nuclear Information System (INIS)

    Mehdi, S.; Wiseman, J.S.

    1989-01-01

    An assay for measuring dihydroorotase activity was devised. Radiolabeled substrate and product were separated by high-performance liquid chromatography using a reverse-phase column with ion-pairing, and the radioactivity was quantitated by flow detection

  4. Nuclide separation modeling through reverse osmosis membranes in radioactive liquid waste

    OpenAIRE

    Lee, Byung-Sik

    2015-01-01

    The aim of this work is to investigate the transport mechanism of radioactive nuclides through the reverse osmosis (RO) membrane and to estimate its effectiveness for nuclide separation from radioactive liquid waste. An analytical model is developed to simulate the RO separation, and a series of experiments are set up to confirm its estimated separation behavior. The model is based on the extended Nernst–Plank equation, which handles the convective flux, diffusive flux, and electromigration f...

  5. Solidification Technologies for Radioactive and Chemical Liquid Waste Treatment - Final CRADA Report

    International Nuclear Information System (INIS)

    Castiglioni, Andrew J.; Gelis, Artem V.

    2016-01-01

    This project, organized under DOE/NNSA's Global Initiatives for Proliferation Prevention program, joined Russian and DOE scientists in developing more effective solidification and storage technologies for liquid radioactive waste. Several patent applications were filed by the Russian scientists (Russia only) and in 2012, the technology developed was approved by Russia's Federal State Unitary Enterprise RADON for application throughout Russia in cleaning up and disposing of radioactive waste.

  6. Control of synthesis and release of radioactive acetylcholine in brain slices from the rat. Effects of neurotropic drugs

    Science.gov (United States)

    Grewaal, D. S.; Quastel, J. H.

    1973-01-01

    1. Studies of the synthesis and release of radioactive acetylcholine in rat brain-cortex slices incubated in Locke–bicarbonate–[U-14C]glucose media, containing paraoxon as cholinesterase inhibitor, revealed the following phenomena: (a) dependence of K+-or protoveratrine-stimulated acetylcholine synthesis and release on the presence of Na+ and Ca2+ in the incubation medium, (b) enhanced release of radioactive acetylcholine by substances that promote depolarization at the nerve cell membrane (e.g. high K+, ouabain, protoveratrine, sodium l-glutamate, high concentration of acetylcholine), (c) failure of acetylcholine synthesis to keep pace with acetylcholine release under certain conditions (e.g. the presence of ouabain or lack of Na+). 2. Stimulation by K+ of radioactive acetylcholine synthesis was directly proportional to the external concentration of Na+, but some synthesis and release of radioactive acetylcholine occurred in the absence of Na+ as well as in the absence of Ca2+. 3. The Na+ dependence of K+-stimulated acetylcholine synthesis was partly due to suppression of choline transport, as addition of small concentrations of choline partly neutralized the effect of Na+ lack, and partly due to the suppression of the activity of the Na+ pump. 4. Protoveratrine caused a greatly increased release of radioactive acetylcholine without stimulating total radioactive acetylcholine synthesis. Protoveratrine was ineffective in the absence of Ca2+ from the incubation medium. It completely blocked K+ stimulation of acetylcholine synthesis and release. 5. Tetrodotoxin abolished the effects of protoveratrine on acetylcholine release. It had blocking effects (partial or complete) on the action of high K+, sodium l-glutamate and lack of Ca2+ on acetylcholine synthesis and release. 6. Unlabelled exogenous acetylcholine did not diminish the content of labelled tissue acetylcholine, derived from labelled glucose, suggesting that no exchange with vesicular acetylcholine took

  7. Release of radioactive fission products from BN-600 reactor untight fuel elements

    International Nuclear Information System (INIS)

    Osipov, S.L.; Tsikunov, A.G.; Lisitsin, E.C.

    1996-01-01

    The experimental data on the release of radioactive fission products from BN-600 reactor untight fuel elements are given in the report. Various groups of radionuclides: inert gases Xe, Kr, volatile Cs, J, non-volatile Nb, and La are considered. The results of calculation-experimental study of transfer and distribution of radionuclides in the reactor primary circuit, gas system and sodium coolant are considered. It is shown that some complex radioactivity transfer processes can be described by simple mathematical models. (author)

  8. Kinetics of hydrogen release from dissolutions of ammonia borane in different ionic liquids

    International Nuclear Information System (INIS)

    Valero-Pedraza, María José; Martín-Cortés, Alexandra; Navarrete, Alexander; Bermejo, María Dolores; Martín, Ángel

    2015-01-01

    Ammonia borane is a promising hydrogen storage material that liberates hydrogen by thermolysis at moderate temperatures, but it also presents major limitations for practical applications including a long induction time before the initiation of hydrogen release and a difficult regeneration. Previous works have demonstrated that by dissolution of ammonia borane into several ionic liquids, and particularly in 1-butyl-3-methylimidazolium chloride bmimCl, the induction period at the beginning of the thermolysis is eliminated, but some problems persist, including foaming and the formation of a residue after thermolysis that is insoluble in the ionic liquid. In this work, the release of hydrogen from ammonia borane dissolved in different ionic liquids has been analyzed, measuring the kinetics of hydrogen release, visually following the evolution of the sample during the process using pressure glass tube reactors, and analyzing the residue by spectroscopic techniques. While dissolutions of ammonia borane in most ionic liquids analyzed show similar properties as dissolutions in bmimCl, using ionic liquids with bis(trifluoromethylsulfanyl)imide Tf_2N anion the foaming problem is reduced, and in some cases the residue remains dissolved in the ionic liquid, while with ionic liquids with choline anion higher hydrogen yields are achieved that indicate that the decomposition of ammonia borane proceeds through a different path. - Highlights: • Hydrogen release from ammonia borane dissolved in 13 ionic liquids has been studied. • Induction time is shortened and hydrogen release rate is accelerated in all cases. • The best results are obtained using ionic liquids with Tf_2N anion. • Ch cation ionic liquids enable higher H_2 yield, but cyclotriborazane is produced.

  9. Nuclide separation modeling through reverse osmosis membranes in radioactive liquid waste

    Directory of Open Access Journals (Sweden)

    Byung-Sik Lee

    2015-12-01

    Full Text Available The aim of this work is to investigate the transport mechanism of radioactive nuclides through the reverse osmosis (RO membrane and to estimate its effectiveness for nuclide separation from radioactive liquid waste. An analytical model is developed to simulate the RO separation, and a series of experiments are set up to confirm its estimated separation behavior. The model is based on the extended Nernst–Plank equation, which handles the convective flux, diffusive flux, and electromigration flux under electroneutrality and zero electric current conditions. The distribution coefficient which arises due to ion interactions with the membrane material and the electric potential jump at the membrane interface are included as boundary conditions in solving the equation. A high Peclet approximation is adopted to simplify the calculation, but the effect of concentration polarization is included for a more accurate prediction of separation. Cobalt and cesium are specifically selected for the experiments in order to check the separation mechanism from liquid waste composed of various radioactive nuclides and nonradioactive substances, and the results are compared with the estimated cobalt and cesium rejections of the RO membrane using the model. Experimental and calculated results are shown to be in excellent agreement. The proposed model will be very useful for the prediction of separation behavior of various radioactive nuclides by the RO membrane.

  10. Nuclide separation modeling through reverse osmosis membranes in radioactive liquid waste

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Sik [KEPCO Engineering and Construction, Gimcheon (Korea, Republic of)

    2015-12-15

    The aim of this work is to investigate the transport mechanism of radioactive nuclides through the reverse osmosis (RO) membrane and to estimate its effectiveness for nuclide separation from radioactive liquid waste. An analytical model is developed to simulate the RO separation, and a series of experiments are set up to confirm its estimated separation behavior. The model is based on the extended Nernst-Plank equation, which handles the convective flux, diffusive flux, and electromigration flux under electroneutrality and zero electric current conditions. The distribution coefficient which arises due to ion interactions with the membrane material and the electric potential jump at the membrane interface are included as boundary conditions in solving the equation. A high Peclet approximation is adopted to simplify the calculation, but the effect of concentration polarization is included for a more accurate prediction of separation. Cobalt and cesium are specifically selected for the experiments in order to check the separation mechanism from liquid waste composed of various radioactive nuclides and nonradioactive substances, and the results are compared with the estimated cobalt and cesium rejections of the RO membrane using the model. Experimental and calculated results are shown to be in excellent agreement. The proposed model will be very useful for the prediction of separation behavior of various radioactive nuclides by the RO membrane.

  11. Radioactivity measurement of barium carbonate [14C] by liquid scintillation counting

    International Nuclear Information System (INIS)

    Kobayashi, Katsutoshi; Hoizumi, Kiyoshi

    1985-03-01

    Two methods of sample preparation for the measurement of specific activity of BaCO 3 [ 14 C] by external standard method in liquid scintillation counting were studied. BaCO 3 [ 14 C] was decomposed by perchloric acid solution and generated CO 2 [ 14 C] was absorbed by ethylene glycol monomethyl ether solution of monoethanolamine as the method 1 or aqueous sodium hydroxide as the method 2. In order to prepare the sample solution of adequate radioactivity concentration, these carbonate solutions by the methods 1 and 2 were diluted with the suitable organic solvent and distilled water respectively. One tenth millilitre of these sample solutions was added into 10 ml of PPO-toluene scintillator containing 0.1 ml of monoethanolamine in a counting vial and homogeneously dissolved with ethyl alcohol. The results of the radioactivity measurement of BaCO 3 [ 14 C] based on the different method agreed within 5 % and the counting rate was found to be stable for as long as 7 deays or more. Both methods of preparation are suitable for the routine measurement because of their simplicity and feasibility. In the case of method 2, the liquid radioactive waste is almost inorganic solution and recovery in the form of BaCO 3 [ 14 C] is easily performed, so that this method is very advantageous from the view point of the radioactive waste treatement. (author)

  12. Solid and liquid radioactive waste management of the Nuclear Technology Development Center (CDTN) - NUCLEBRAS

    International Nuclear Information System (INIS)

    Guzella, M.F.R.; Miaw, S.T.W.; Mourao, R.P.; Prado, M.A.S. do; Reis, L.C.A.; Santos, P.O.; Silva, E.M.P.

    1986-01-01

    Low level liquid and solid wastes are produced in several laboratories of the NUCLEAR TECHNOLOGY DEVELOPMENT CENTER (CDTN)-NUCLEBRAS. In the last years, the intensification of technical activities at the Center has increased the radioactive waste volumes. Therefore, the implementation of a Radioactive Waste Management Program has begun. This Program includes the systematic of activities from the waste collection to the transportation for the final disposal. The liquid and solid waste are collected separately in proper containers and stored for later treatment according to the processes available or under development at the Center. (Author) [pt

  13. Solid and liquid radioactive waste management of the Nuclear Technology Development Center (CDTN)- Nuclebras

    International Nuclear Information System (INIS)

    Guzella, M.F.R.; Mourao, R.P.; Reis, L.C.A.; Silva, E.M.P.; Miaw, S.T.W.; Prado, M.A.S.; Santos, P.O.

    1986-01-01

    Low level liquid and solid wastes are produced in several laboratories of the NUCLEAR TECHNOLOGY DEVELOPMENT CENTER (CDTN) - NUCLEBRAS. In the last years, the intensification of technical activities at the Center has increased the radioactive waste volumes. Therefore, the implementation of a Radioactive Waste Management Program has begun. This Program includes the systematic of activities from the waste collection to the transportation for the final disposal. The liquid and solid waste are collected separately in proper containers and stored for later treatment according to the processes available or under development at the Center. (Author) [pt

  14. Method for the simultaneous recovery of radionuclides from liquid radioactive wastes using a solvent

    Science.gov (United States)

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

    2001-01-01

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

  15. Environmental releases for calendar year 1997

    International Nuclear Information System (INIS)

    Gleckler, B.P.

    1998-01-01

    This report fulfills the annual environmental release reporting requirements of US Department of Energy (DOE) Order 5400.1. This report provides supplemental information to the Hanford Site Environmental Report (PNNL-11795). The Hanford Site Environmental Report provides an update on the environmental status of the Hanford Site. The sitewide annual report summarizes the degree of compliance with applicable environmental regulations and informs the public concerning the impact of Hanford Site operations on the surrounding environment. Like the Hanford Site Environmental Report, this annual report presents a summary of the environmental releases from facilities and activities managed by the Fluor Daniel Hanford, Incorporated (FDH), and Bechtel Hanford, Incorporated (BHI). In addition to the summary data, this report also includes detailed data on air emissions, liquid effluents, and hazardous substances released to the environment during calendar year 1997. Comprehensive data summaries of air emissions and liquid effluents in 1997 are displayed in Tables ES-1 through ES-5. These tables represent the following: Table ES-1--Radionuclide air emissions data (detailed data on emissions are presented in Section 2.0); Table ES-2--Data on radioactive liquid effluents discharged to the soil (detailed data are presented in Section 3.0); Table ES-3--Radionuclides discharged to the Columbia River (detailed data are presented in Section 3.0); Table ES-4--Nonradioactive air emissions data (detailed data are presented in Section 2.0); Table ES-5--Total Volumes and Flow Rates of 200/600 Area Radioactive Liquid Effluents (detailed data are presented in Section 3.0)

  16. Situation 2002: release monitoring and surveillance of environment of Cea centers

    International Nuclear Information System (INIS)

    2003-07-01

    This publication renders an account of the situation of the releases of liquid and gaseous radioactive effluents, for the year 2002, as well as the radioactivity levels measured in the vicinity of Cea centers through the systematic surveillance of atmosphere, waters, vegetation and milk. An analysis on five years allows to follow their evolution. (N.C.)

  17. Application of two-barrier model of radioactive agent transport in sea water for analyzing artificial radionuclide release from containers with radioactive waste dumped in Kara Sea

    Energy Technology Data Exchange (ETDEWEB)

    Grishin, Denis S.; Laykin, Andrey I.; Kuchin, Nickolay L.; Platovskikh, Yuri A. [Krylov State Research Center, Saint Petersburg, 44 Moskovskoe shosse, 196158 (Russian Federation)

    2014-07-01

    Modeling of artificial radionuclide transport in sea water is crucial for prognosis of radioecological situation in regions where dumping of radioactive waste had been made and/or accidents with nuclear submarines had taken place. Distribution of artificial radionuclides in bottom sediments can be a detector of radionuclide release from dumped or sunk objects to marine environment. Proper model can determine the dependence between radionuclide distribution in sediments and radionuclide release. Following report describes two-barrier model of radioactive agent transport in sea water. It was tested on data from 1994 - 2013 expeditions to Novaya Zemlya bays, where regular dumping of solid radioactive waste was practiced by the former USSR from the early 1960's until 1990. Two-barrier model agrees with experimental data and allows more accurate determination of time and intensity of artificial radionuclide release from dumped containers. (authors)

  18. Condition assessment of the Los Alamos National Laboratory radioactive liquid waste collection system

    International Nuclear Information System (INIS)

    Edgemon, G.L.; Moss, W.D.; Worland, V.P.

    2004-01-01

    The radioactive liquid waste collection system (RLWCS) at Los Alamos National Laboratory (LANE) is a site-wide double-encased piping system installed in 1982 that allows radioactive liquid waste (RLW) producing facilities to gravity drain their waste to the radioactive liquid waste treatment facility (RLWTF) through a system of underground high-density polyethylene (HDPE) pipes and vaults. The RLWCS stretches approximately four miles and typically receives approximately 10,000 gallons of RLW per day for treatment at the RLWTF. Uncertainty of the current condition of the RLWCS was recently identified as a potential risk to the future continued availability of the RLW treatment function. A condition assessment was performed in April 2004 to evaluate the risks and estimate the remaining useful life of the existing RLWCS. Several representative and 'worst-case' RLWCS primary piping sections and their associated inspection vaults were selected for direct visual assessment, remote borescopic examination, and in-situ durometer testing. This field investigation combined with an RLWCS materials compatibility review showed that the primary piping of the RLWCS is in relatively good condition, with only a few noteworthy areas of degradation.

  19. Separation of aromatic precipitates from simulated high level radioactive waste by hydrolysis, evaporation and liquid-liquid extraction

    International Nuclear Information System (INIS)

    Young, S.R.; Shah, H.B.; Carter, J.T.

    1991-01-01

    The Defense Waste Processing Facility (DWPF) at the SRS will be the United States' first facility to process High Level radioactive Waste (HLW) into a borosilicate glass matrix. The removal of aromatic precipitates by hydrolysis, evaporation and liquid-liquid extraction will be a key step in the processing of the HLW. This step, titled the Precipitate Hydrolysis Process, has been demonstrated by the Savannah River Laboratory with the Precipitate Hydrolysis Experimental Facility (PHEF). The mission of the PHEF is to demonstrate processing of simulated high level radioactive waste which contains tetraphenylborate precipitates and nitrite. Reduction of nitrite by hydroxylamine nitrate and hydrolysis of the tetraphenylborate by formic acid is discussed. Gaseous production, which is primarily benzene, nitrous oxide and carbon dioxide, has been quantified. Production of high-boiling organic compounds and the accumulation of these organic compounds within the process are addressed

  20. Design of Biochemical Oxidation Process Engineering Unit for Treatment of Organic Radioactive Liquid Waste

    International Nuclear Information System (INIS)

    Zainus Salimin; Endang Nuraeni; Mirawaty; Tarigan, Cerdas

    2010-01-01

    Organic radioactive liquid waste from nuclear industry consist of detergent waste from nuclear laundry, 30% TBP-kerosene solvent waste from purification or recovery of uranium from process failure of nuclear fuel fabrication, and solvent waste containing D 2 EHPA, TOPO, and kerosene from purification of phosphoric acid. The waste is dangerous and toxic matter having low pH, high COD and BOD, and also low radioactivity. Biochemical oxidation process is the effective method for detoxification of organic waste and decontamination of radionuclide by bio sorption. The result process are sludges and non radioactive supernatant. The existing treatment facilities radioactive waste in Serpong can not use for treatment of that’s organics waste. Dio chemical oxidation process engineering unit for continuous treatment of organic radioactive liquid waste on the capacity of 1.6 L/h has been designed and constructed the equipment of process unit consist of storage tank of 100 L capacity for nutrition solution, 2 storage tanks of 100 L capacity per each for liquid waste, reactor oxidation of 120 L, settling tank of 50 L capacity storage tank of 55 L capacity for sludge, storage tank of 50 capacity for supernatant. Solution on the reactor R-01 are added by bacteria, nutrition and aeration using two difference aerators until biochemical oxidation occurs. The sludge from reactor of R-01 are recirculated to the settling tank of R-02 and on the its reverse operation biological sludge will be settled, and supernatant will be overflow. (author)

  1. Application of reverse osmosis to the treatment of liquid effluents produced by nuclear power plants

    International Nuclear Information System (INIS)

    Huet, Y.; Poulat, B.; Menjeaud, C.

    1989-01-01

    Radioactive liquid effluents generated during the operation of PWR nuclear power units are currently treated by two independent systems. The effluents from the reactor coolant system are recycled, unlike the others, which, after treatment, are released into the river or ocean that provides cooling water for the unit. The objective of the treatment of nonrecycled effluents is to separate from them as much of the radioactive particles that they contain as possible, so as to release into the environment a maximum volume of nonradioactive waste, and to be left with only a minimum volume of concentrated waste, containing most of the initial radioactivity, which must be loaded into casks for storage. Membrane-based filtration techniques, because they have excellent separation performances, can logically be used for this decontamination of the liquid effluents. Having developed its own reverse osmosis membrane, a possible application in a nuclear power plant, i.e., integration of a reverse osmosis unit into a radioactive liquid effluent treatment system is presented. (author)

  2. A Bayesian Analysis of the Radioactive Releases of Fukushima

    DEFF Research Database (Denmark)

    Tomioka, Ryota; Mørup, Morten

    2012-01-01

    the types of nuclides and their levels of concentration from the recorded mixture of radiations to take necessary measures. We presently formulate a Bayesian generative model for the data available on radioactive releases from the Fukushima Daiichi disaster across Japan. From the sparsely sampled...... the Fukushima Daiichi plant we establish that the model is able to account for the data. We further demonstrate how the model extends to include all the available measurements recorded throughout Japan. The model can be considered a first attempt to apply Bayesian learning unsupervised in order to give a more......The Fukushima Daiichi disaster 11 March, 2011 is considered the largest nuclear accident since the 1986 Chernobyl disaster and has been rated at level 7 on the International Nuclear Event Scale. As different radioactive materials have different effects to human body, it is important to know...

  3. Membrane technologies for liquid radioactive waste treatment

    International Nuclear Information System (INIS)

    Chmielewski, A.G.; Harasimowicz, M.; Zakrzewska-Trznadel, G.

    1998-01-01

    At Institute of Nuclear Chemistry and Technology (INCT) the membrane method for purification of radioactive wastes applied such processes as ultrafiltration (UF), 'seeded' ultrafiltration and reverse osmosis (RO) was developed. On the basis of the results obtained in laboratory experiments the pilot plant for radioactive effluents treatment was built. The plant was composed of UF unit (AMICON H 26P30 capillary module) and two RO units (NITTO NTR 739 HF S-4 spiral wound LPRO modules). The capacity of the pilot plant was up to 200 L/h and the specific activity of wastes purified in the system - below 10 4 Bq/L. Decontamination factor for entire system is higher than 5 x10 3 . Another possibility for radioactive wastes treatment is membrane distillation (MD), non-isothermal process employing hydrophobic polymer membrane, which is developed at INCT now. Preliminary tests with liquid radwaste were carried out on laboratory unit with permeation test-cell holding flat sheet membrane. As a hydrophobic barrier membranes made of two polymers were used: polytetrafluoroethylene (PTFE) and polypropylene (PP). The process was arranged in direct contact membrane distillation configuration. The permeate condensed directly in the cold stream (distilled water) and retentate was enriched in radionuclides. The further experiments carried out with capillary module BFMF 06-30-33 (Euro-Sep Ltd.) with polypropylene capillaries, diameter 0.33 mm and cut off 0.6 μm proved previous results. A pilot plant employing GORE-TEX membrane distillation was constructed. The plant can clean the low-level radioactive wastes from nuclear centre, at a throughput about 0.05 m 3 /h

  4. Study of biosorbents application on the treatment of radioactive liquid wastes with americium-241

    International Nuclear Information System (INIS)

    Borba, Tania Regina de

    2010-01-01

    The use of nuclear energy for many different purposes has been intensified and highlighted by the benefits that it provides. Medical diagnosis and therapy, agriculture, industry and electricity generation are examples of its application. However, nuclear energy generates radioactive wastes that require suitable treatment ensuring life and environmental safety. Biosorption and bioaccumulation represent an emergent alternative for the treatment of radioactive liquid wastes, providing volume reduction and physical state change. This work aimed to study biosorbents for the treatment of radioactive liquid wastes contaminated with americium-241 in order to reduce the volume and change the physical state from liquid to solid. The biosorbents evaluated were Saccharomyces cerevisiae immobilized in calcium alginate beads, inactivated and free cells of Saccharomyces cerevisiae, calcium alginate beads, Bacillus subtilis, Cupriavidus metallidurans and Ochrobactrum anthropi. The results were quite satisfactory, achieving 100% in some cases. The technique presented in this work may be useful and viable for implementing at the Waste Management Laboratory of IPEN - CNEN/SP in short term, since it is an easy and low cost method. (author)

  5. Monitoring water for radioactive releases in the United States

    International Nuclear Information System (INIS)

    Porter, C.R.; Broadway, J.A.; Kahn, B.

    1990-01-01

    The major radiological environmental monitoring programs for water in the United States are described. The applications of these programs for monitoring radioactive fallout, routine discharges from nuclear facilities, and releases due to accidents at such facilities are discussed, and some examples of measurements are presented. The programs monitor rainfall, surface water, and water supplies. Samples are usually collected and analyzed on a monthly or quarterly schedule, but the frequency is increased in response to emergencies. (author)

  6. Improved technique for in situ measurement of radioactivity in liquid effluents

    International Nuclear Information System (INIS)

    Balasundar, S.; Jose, M.T.; Ravi, T.; Sundaram, V.M.; Raghunath, V.M.

    1993-01-01

    As a pre-requisite for handling and disposal of radioactive liquid effluents, they should be categorized according to their chemical nature and the type and level of radioactivity. A continuous monitoring of these effluents is necessary to assess the discharged activity and to detect any unusually large increase in the activity level. It may also be required to assess the beta, gamma specific activities in the effluents independently instead of just combined beta-gamma activity as is generally done. (author). 1 ref., 2 tabs

  7. New approaches to deriving limits of the release of radioactive material into the environment

    International Nuclear Information System (INIS)

    Lindell, B.

    1977-01-01

    During the last few years, new principles have been developed for the limitation of the release of radioactive material into the environment. It is no longer considered appropriate to base the limitation on limits for the concentrations of the various radionuclides in air and water effluents. Such limits would not prevent large amounts of radioactive material from reaching the environment should effluent rates be high. A common practice has been to identify critical radionuclides and critical pathways and to base the limitation on authorized dose limits for local ''critical groups''. If this were the only limitation, however, larger releases could be permitted after installing either higher stacks or equipment to retain the more short-lived radionuclides for decay before release. Continued release at such limits would then lead to considerably higher exposure at a distance than if no such installation had been made. Accordingly there would be no immediate control of overlapping exposures from several sources, nor would the system guarantee control of the future situation. The new principles described in this paper take the future into account by limiting the annual dose commitments rather than the annual doses. They also offer means of controlling the global situation by limiting not only doses in critical groups but also global collective doses. Their objective is not only to ensure that individual dose limits will always be respected but also to meet the requirement that ''all doses be kept as low as reasonably achievable''. The new approach is based on the most recent recommendations by the ICRP and has been described in a report by an IAEA panel (Procedures for establishing limits for the release of radioactive material into the environment). It has been applied in the development of new Swedish release regulations, which illustrate some of the problems which arise in the practical application

  8. New approaches to deriving limits of the release of radioactive material into the environment

    International Nuclear Information System (INIS)

    Lindell, B.

    1977-01-01

    During the last few years, new principles have been developed for the limitation of the release of radioactive material into the environment. It is no longer considered appropriate to base the limitation on limits for the concentrations of the various radionuclides in air and water effluents. Such limits would not prevent large amounts of radioactive material from reaching the environment should effluent rates be high. A common practice has been to identify critical radionuclides and critical pathways and to base the limitation on authorized dose limits for local ''critical groups''. If this were the only limitation, however, larger releases could be permitted after installing either higher stacks or equipment to retain the more shortlived radionuclides for decay before release. Continued release at such limits would then lead to considerably higher exposure at a distance than if no such installation had been made. Accordingly there would be no immediate control of overlapping exposures from several sources, nor would the system guarantee control of the future situation. The new principles described in this paper take the future into account by limiting the annual dose commitments rather than the annual doses. They also offer means of controlling the global situation by limiting not only doses in critical groups but also global collective doses. Their objective is not only to ensure that individual dose limits will always be respected but also to meet the requirement that ''all doses be kept as low as reasonably achievable''. The new approach is based on the most recent recommendations by the ICRP and has been described in a report by an IAEA panel (Procedures for Establishing Limits for the Release of Radioactive Material into the Environment). It has been applied in the development of new Swedish release regulations, which illustrate some of the problems which arise in the practical application. (author)

  9. Operations of the LR56 radioactive liquid cask transport system at U.S. Department of Energy sites

    International Nuclear Information System (INIS)

    Davidson, J.S.; Hornstra, D.J.; Sazawal, V.K.; Clement, G.

    1996-01-01

    The LR56 cask system is licensed for use in France under Certificate of Compliance F/309/B(U)F for transport of 4,000-liter volumes of radioactive liquids. Three LR56 cask systems (with modifications for use at Department of Energy (DOE) sites) have been purchased for delivery at the Hanford Site, Oak Ridge National Laboratory (ORNL), and Savannah River Site (SRS). The LR56 cask systems will be used for on-site transfers of Type B quantities of radioactive liquid waste. The ORNL unit will also be used as a Type A packaging for transfers of radioactive liquids between DOE sites. This paper discusses LR56 operating features and the use of the cask system at the three DOE sites

  10. PAVAN, Atmospheric Dispersion of Radioactive Releases from Nuclear Power Plants

    International Nuclear Information System (INIS)

    2001-01-01

    1 - Description of program or function: PAVAN estimates down-wind ground-level air concentrations for potential accidental releases of radioactive material from nuclear facilities. Options can account for variation in the location of release points, additional plume dispersion due to building wakes, plume meander under low wind speed conditions, and adjustments to consider non-straight trajectories. It computes an effective plume height using the physical release height which can be reduced by inputted terrain features. 2 - Method of solution: Using joint frequency distributions of wind direction and wind speed by atmospheric stability, the program provides relative air concentration (X/Q) values as functions of direction for various time periods at the exclusion area boundary (EAB) and the outer boundary of the low population zone (LPZ). Calculations of X/Q values can be made for assumed ground-level releases or evaluated releases from free-standing stacks. The X/Q calculations are based on the theory that material released to the atmosphere will be normally distributed (Gaussian) about the plume centerline. A straight-line trajectory is assumed between the point of release and all distances for which X/Q values are calculated. 3 - Restrictions on the complexity of the problem: - The code cannot handle multiple emission sources

  11. Radioactive liquid waste discharged from Nuclear Electric licensed sites during 1991

    International Nuclear Information System (INIS)

    Austin, L.S.; Odell, K.J.

    1993-03-01

    This report presents the detailed isotopic composition of radioactive liquid waste discharged from Nuclear Electric licensed sites in 1991. Liquid discharges from those Magnox stations using pond storage of irradiated fuel contained low levels of activation and fission products, while those from Wylfa and the AGR stations contained lower levels of activation products with only traces of fission products. Discharges were similar to those observed in previous years, with any changes concordant with changes in stations' generation performance. (author)

  12. A history of effluent releases from the Texas A and M University reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bates, E F; Neff, R D; Sandel, P S; Schoenbucher, B [Texas A and M University (United States)

    1974-07-01

    Since 1966 records of radioactive effluents releases from the Texas A and M University Research Reactor have been compiled. These data include particulate activity, noble gases, and liquid effluent releases. Particulate activity releases with half-lives greater than eight days were negligible and are not included in this presentation. Conversion from an MTR plate reactor to a TRIGA fueled reactor was completed in August 1968. Records of effluent releases of Argon-4l and liquids for the past, five years are summarized, in this presentation. These release data are compared to the current limits specified: in 10 CPR 20 and the limits appearing in proposed Appendix.

  13. Low and medium level liquid waste processing at the new La Hague reprocessing plant

    International Nuclear Information System (INIS)

    Alexandre, D.

    1986-05-01

    Reprocessing of spent nuclear fuels produces low and medium activity liquid wastes. These radioactive wastes are decontamined before release in environment. The new effluent processing plant, which is being built at La Hague, is briefly described. Radionuclides are removed from liquid wastes by coprecipitation. The effluent is released after decantation and filtration. Insoluble sludges are conditioned in bitumen [fr

  14. Regulatory analysis on criteria for the release of patients administered radioactive material

    International Nuclear Information System (INIS)

    Schneider, S.; McGuire, S.A.

    1994-05-01

    The Nuclear Regulatory Commission (NRC) has received two petitions to amend its regulations in 10 CFR Parts 20 and 35 as they apply to doses received by members of the public exposed to patients released from a hospital after they have been administered radioactive material. While the two petitions are not identical they both request that the NRC establish a dose limit of 5 millisieverts (0.5 rem) per year for individuals exposed to patients who have been administered radioactive materials. This Regulatory Analysis evaluates three alternatives. Alternative 1 is for the NRC to amend its patient release criteria in 10 CFR 35.75 to use the more stringent dose limit of 1 millisievert per year in 10 CFR 20.1301(a) for its patient release criteria. Alternative 2 is for the NRC to continue using the existing patient release criteria in 10 CFR 35.75 of 1,110 megabecquerels of activity or a dose rate at one meter from the patient of 0.05 millisievert per hour. Alternative 3 is for the NRC to amend the patient release criteria in 10 CFR 35.75 to specify a dose limit of 5 millisieverts for patient release. The evaluation indicates that Alternative 1 would cause a prohibitively large increase in the national health care cost from retaining patients in a hospital longer and would cause significant personal and psychological costs to patients and their families. The choice of Alternatives 2 or 3 would affect only thyroid cancer patients treated with iodine-131. For those patients, Alternative 3 would result in less hospitalization than Alternative 2. Alternative 3 has a potential decrease in national health care cost of $30,000,000 per year but would increase the potential collective dose from released therapy patients by about 2,700 person-rem per year, mainly to family members

  15. Regulatory analysis on criteria for the release of patients administered radioactive material

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, S.; McGuire, S.A. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications; Behling, U.H.; Behling, K.; Goldin, D. [Cohen (S.) and Associates, Inc., McLean, VA (United States)

    1994-05-01

    The Nuclear Regulatory Commission (NRC) has received two petitions to amend its regulations in 10 CFR Parts 20 and 35 as they apply to doses received by members of the public exposed to patients released from a hospital after they have been administered radioactive material. While the two petitions are not identical they both request that the NRC establish a dose limit of 5 millisieverts (0.5 rem) per year for individuals exposed to patients who have been administered radioactive materials. This Regulatory Analysis evaluates three alternatives. Alternative 1 is for the NRC to amend its patient release criteria in 10 CFR 35.75 to use the more stringent dose limit of 1 millisievert per year in 10 CFR 20.1301(a) for its patient release criteria. Alternative 2 is for the NRC to continue using the existing patient release criteria in 10 CFR 35.75 of 1,110 megabecquerels of activity or a dose rate at one meter from the patient of 0.05 millisievert per hour. Alternative 3 is for the NRC to amend the patient release criteria in 10 CFR 35.75 to specify a dose limit of 5 millisieverts for patient release. The evaluation indicates that Alternative 1 would cause a prohibitively large increase in the national health care cost from retaining patients in a hospital longer and would cause significant personal and psychological costs to patients and their families. The choice of Alternatives 2 or 3 would affect only thyroid cancer patients treated with iodine-131. For those patients, Alternative 3 would result in less hospitalization than Alternative 2. Alternative 3 has a potential decrease in national health care cost of $30,000,000 per year but would increase the potential collective dose from released therapy patients by about 2,700 person-rem per year, mainly to family members.

  16. Parametric study of radioactive release from a breached containment

    International Nuclear Information System (INIS)

    Kenigsberg, A.; Hasan, D.; Elias, E.

    1980-01-01

    A hypothetical accident is analyzed, in which an external (out-of-plant) natural or man-made event causes a loss-of-coolant accident after penetrating the containment wall. The computer codes CONTEMPT and RELAP4 have been used to study the containment thermal-hydraulic behavior during the accident. Results are given in the form of graphs showing the thermal-hydraulic response of the containment and the profile of radioactive release to the atmosphere. The physical model and input data are discussed. 13 refs

  17. Engineering study radioactive liquid waste treatment plant refurbishment

    International Nuclear Information System (INIS)

    Suazo, I.L.

    1994-01-01

    This feasibility study will investigate the opportunities, restrictions and cost impact to refurbish the existing Radioactive Liquid Waste Treatment Plant (RLWTP) while utilizing the same basic criteria that was used in the development of the new Radioactive Liquid Waste Treatment Facility (RLWTF). The objective of this study is to perform a more in-depth analysis of refurbishing the existing than has been done in the past so as to provide a basis for comparison between refurbishing the existing or constructing a new. The existing plant is located at Technical Area 50 (TA-50) within the Los Alamos National Laboratory (LANL). The initial structure was built in 1963. Over the ensuing years, the building has been modified and several additions have been constructed. In 1966, laboratories, ion exchange and pretreatment functions were added. The decontamination and decommissioning activities and ventilation equipment were added in 1984. The following assumptions are the basic parameters considered in the development of a design concept to refurbish the RLWTP: (1) Allow continued operation of the during retrofit construction. (2) Design the necessary expansion within the site constraints. (3) Satisfy National Pollutant Discharge Elimination System (NPDES) and National Emission Standards for Hazardous Air Pollutants (NESHAPS) permit conditions and other environmental regulations. (4) Comply with present DOE Orders and building code requirements. The refurbishment concept is a phased demolition and construction process

  18. Elimination of slightly radioactive liquid effluent by dilution. Its consequences (1960)

    International Nuclear Information System (INIS)

    Bovard, P.; Candillon, C.

    1960-01-01

    Nuclear centres often have to solve problems raised by the elimination of large volumes of slightly radioactive liquid effluent. As things stand at present, the method usually adopted consists in expelling this effluent into the main water system in order to dilute it to a maximum, and thus to lower its radioactive isotope concentration to below the norms imposed by the Public Health Service. This technique requires systematic checking of the water system and its dependences, and demands a thorough knowledge of adsorption and fixation processes. (author) [fr

  19. Influence of radiation on the system liquid radioactive wastes: geologic formation

    International Nuclear Information System (INIS)

    Spitsyn, V.I.; Balukova, V.D.; Kabakchi, S.A.; Medvedeva, M.L.

    1979-01-01

    Introduction of liquid radioactive wastes into deep strata-collectors results in a number of physical-chemical processes: precipitation, dissolution, complex formation, sorption, etc. The area occupied by the injected waste and changes in the nature of the liquid phase depend primarily on radiolysis processes in the heterogeneous system of liquid waste-stratal material occurring at elevated temperatures and pressures. Experiments that simulate actual conditions of temperature, pressure and high radiation levels on this system have been performed. Results are presented for radiolytic gas formation and for changes in the liquid phase and sorption capacity of stratal minerals. It is shown that the temperature increase in the stratum-collector significantly enhances waste decomposition processes, promotes sorption of radionuclides and decreases the mobility of the waste in the formation

  20. Conceptual Design of Portable Filtered Air Suction Systems For Prevention of Released Radioactive Gas under Severe Accidents of NPP

    Energy Technology Data Exchange (ETDEWEB)

    Gu, Beom W.; Choi, Su Y.; Yim, Man S.; Rim, Chun T. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    It becomes evident that severe accidents may occur by unexpected disasters such as tsunami, heavy flood, or terror. Once radioactive material is released from NPP through severe accidents, there are no ways to prevent the released radioactive gas spreading in the air. As a remedy for this problem, the idea on the portable filtered air suction system (PoFASS) for the prevention of released radioactive gas under severe accidents was proposed. In this paper, the conceptual design of a PoFASS focusing on the number of robot fingers and robot arm rods are proposed. In order to design a flexible robot suction nozzle, mathematical models for the gaps which represent the lifted heights of extensible covers for given convex shapes of pipes and for the covered areas are developed. In addition, the system requirements for the design of the robot arms of PoFASS are proposed, which determine the accessible range of leakage points of released radioactive gas. In this paper, the conceptual designs of the flexible robot suction nozzle and robot arm have been conducted. As a result, the minimum number of robot fingers and robot arm rods are defined to be four and three, respectively. For further works, extensible cover designs on the flexible robot suction nozzle and the application of the PoFASS to the inside of NPP should be studied because the radioactive gas may be released from connection pipes between the containment building and auxiliary buildings.

  1. Radioactive waste disposal areas and associated environmental surveillance data at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Oakes, T.W.; Shank, K.E.

    1979-12-01

    Environmental surveillance data have been collected around radioactive waste disposal areas for the past thirty years at Oak Ridge National Laboratory (ORNL). The wealth of data collected around the ORNL radioactive waste burial grounds is presented in this review. The purpose of this paper is to describe the solid waste burial grounds in detail along with the environmental monitoring data. The various monitoring systems are reviewed, and the liquid discharge trends are discussed. Monitoring at White Oak Dam, the last liquid control point for the Laboratory, was started in the late 1940's and is continuing. Presently, a network of five environmental monitoring stations is in operation to monitor the radionuclide content of surface waters in the White Oak Creek watershed. Facts observed during the lifetime of the disposal sites include: (1) a large amount of 106 Ru released during 1959 to 1964 due to the fact that Conasauga shale did not retain this element as well as it retained other radionuclides. (2) Large quantities of tritiated water have been released to the Clinch River in recent years, but, from a practical standpoint, little can be done to inhibit or control these releases. (3) A general downward trend in the number of curies released has been observed for all other radionuclides. A number of corrective measures that have been initiated at ORNL to reduce the radioactive liquid discharges are outlined in the paper

  2. Experimental Study on Ice Forming Process of Cryogenic Liquid Releasing underwater

    Science.gov (United States)

    Zhang, Bin; Wu, Wanqing; Zhang, Xingdong; Zhang, Yi; Zhang, Chuanlin; Zhang, Haoran; Wang, Peng

    2017-11-01

    Cryogenic liquid releasing into water would be a process combines hyperactive boiling with ice forming. There are still few researches on the experimental study on the environmental conditions for deciding ice forming speed and liquid surviving state. In this paper, to advance our understanding of ice forming deciding factors in the process of LN2 releasing underwater, a visualization experimental system is built. The results show that the pressure difference significantly influences the ice forming speed and liquid surviving distance, which is observed by the experiment and theoretically analysed by Kelvin-Helmholtz instability. Adding nucleating agent is helpful to provide ice nucleus which can accelerate the ice forming speed. Water flowing has some effect on changing pressure difference, which can affect the ice forming speed and liquid surviving distance.

  3. Quality evaluation of extemporaneous delayed-release liquid formulations of lansoprazole.

    Science.gov (United States)

    Melkoumov, Alexandre; Soukrati, Amina; Elkin, Igor; Forest, Jean-Marc; Hildgen, Patrice; Leclair, Grégoire

    2011-11-01

    The quality attributes of extemporaneous delayed-release liquid formulations of lansoprazole for oral administration were evaluated. A novel liquid formulation (3 mg/mL) of Prevacid FasTab in an Ora-Blend vehicle was prepared and compared with the Prevacid FasTab 30 mg and Prevacid-sodium bicarbonate 1 M formulation (3 mg/mL). The latter formulation was combined with hydrochloric acid 0.1 N, and the remaining lansoprazole content was assayed by high-performance liquid chromatography (HPLC). A batch of delayed-release liquid formulation was prepared to evaluate content uniformity. For content assay, three samples were prepared for each evaluated condition and each sample was analyzed in triplicate by HPLC. The lansoprazole in the sodium bicarbonate formulation was extensively degraded by quantities of hydrochloric acid 0.1 N in excess of 100 mL. Storage time and temperature had a significant effect on lansoprazole stability in the Ora-Blend formulation. The drug remained stable for seven days when the formulation was stored at 4.5-5.5 °C, but storage at 21-22 °C or the reduction of pH with citric acid accelerated lansoprazole degradation. The amount of lansoprazole released from the Ora-Blend formulation during the buffer stage of the dissolution test decreased with increases in formulation storage time, in formulation storage temperature, and in the amount of lansoprazole released and degraded during the acid stage of the test. An extemporaneous formulation consisting of lansoprazole microgranules in Ora-Blend maintained acceptable quality attributes when stored for three days at 4.5-5.5 °C.

  4. Apparatus of vaporizing and condensing liquid radioactive wastes and its operation method

    International Nuclear Information System (INIS)

    Irie, Hiromitsu; Tajima, Fumio.

    1975-01-01

    Object: To prevent corrosion of material for a vapor-condenser and a vapor heater and to prevent radioactive contamination of heated vapor. Structure: Liquid waste is fed from a liquid feeding tank to a vapor-condenser to vaporize and condense the waste. Uncondensed liquid waste, which is not in a level of a given density, is temporally stored in a batch tank through a switching valve and a pipe. Prior to successive feeding from the liquid feeding tank, the uncondensed liquid waste within the batch tank is returned by a return pump to the condenser, after which a new liquid is fed from the liquid feeding tank for re-vaporization and condensation in the vapor-condenser. Then, similar operation is repeated until the uncondensed liquid waste assumes a given density, and when the uncondensed liquid waste reaches a given density, the condensed liquid waste is discharged into the storage tank through the switching valve. (Ohara, T.)

  5. Radioactive waste management

    International Nuclear Information System (INIS)

    Alfredson, P.G.; Levins, D.M.

    1975-08-01

    Present and future methods of managing radioactive wastes in the nuclear industry are reviewed. In the stages from uranium mining to fuel fabrication, the main purpose of waste management is to limit and control dispersal into the environment of uranium and its decay products, particularly radium and radon. Nuclear reactors produce large amounts of radioactivity but release rates from commercial power reactors have been low and well within legal limits. The principal waste from reprocessing is a high activity liquid containing essentially all the fission products along with the transuranium elements. Most high activity wastes are currently stored as liquids in tanks but there is agreement that future wastes must be converted into solids. Processes to solidify wastes have been demonstrated in pilot plant facilities in the United States and Europe. After solidification, wastes may be stored for some time in man-made structures at or near the Earth's surface. The best method for ultimate disposal appears to be placing solid wastes in a suitable geological formation on land. (author)

  6. Optimal operation planning of radioactive waste processing system by fuzzy theory

    International Nuclear Information System (INIS)

    Yang, Jin Yeong; Lee, Kun Jai

    2000-01-01

    This study is concerned with the applications of linear goal programming and fuzzy theory to the analysis of management and operational problems in the radioactive processing system (RWPS). The developed model is validated and verified using actual data obtained from the RWPS at Kyoto University in Japan. The solution by goal programming and fuzzy theory would show the optimal operation point which is to maximize the total treatable radioactive waste volume and minimize the released radioactivity of liquid waste even under the restricted resources. (orig.)

  7. A Genetic Algorithm Approach to the Optimization of a Radioactive Waste Treatment System

    International Nuclear Information System (INIS)

    Yang, Yeongjin; Lee, Kunjai; Koh, Y.; Mun, J.H.; Kim, H.S.

    1998-01-01

    This study is concerned with the applications of goal programming and genetic algorithm techniques to the analysis of management and operational problems in the radioactive waste treatment system (RWTS). A typical RWTS is modeled and solved by goal program and genetic algorithm to study and resolve the effects of conflicting objectives such as cost, limitation of released radioactivity to the environment, equipment utilization and total treatable radioactive waste volume before discharge and disposal. The developed model is validated and verified using actual data obtained from the RWTS at Kyoto University in Japan. The solution by goal programming and genetic algorithm would show the optimal operation point which is to maximize the total treatable radioactive waste volume and minimize the released radioactivity of liquid waste even under the restricted resources. The comparison of two methods shows very similar results. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  9. Processing method and device for radioactive liquid waste

    International Nuclear Information System (INIS)

    Matsuo, Toshiaki; Nishi, Takashi; Matsuda, Masami; Yukita, Atsushi.

    1997-01-01

    When only suspended particulate ingredients are contained as COD components in radioactive washing liquid wastes, the liquid wastes are heated by a first process, for example, an adsorption step to adsorb the suspended particulate ingredients to an activated carbon, and then separating and removing the suspended particulate ingredients by filtration. When both of the floating particle ingredients and soluble organic ingredients are contained, the suspended particulate ingredients are separated and removed by the first process, and then soluble organic ingredients are removed by other process, or both of the suspended particulate ingredients and the soluble organic ingredients are removed by the first process. In an existent method of adding an activated carbon and then filtering them at a normal temperature, the floating particle ingredients cover the layer of activated carbon formed on a filter paper or fabric to sometimes cause clogging. However, according to the method of the present invention, since disturbance by the floating particle ingredients does not occur, the COD components can be separated and removed sufficiently without lowering liquid waste processing speed. (T.M.)

  10. Modified microspheres for cleaning liquid wastes from radioactive nuclides

    International Nuclear Information System (INIS)

    Danilin, Lev; Drozhzhin, Valery

    2007-01-01

    An effective solution of nuclear industry problems related to deactivation of technological and natural waters polluted with toxic and radioactive elements is the development of inorganic sorbents capable of not only withdrawing radioactive nuclides, but also of providing their subsequent conservation under conditions of long-term storage. A successful technical approach to creation of sorbents can be the use of hollow aluminosilicate microspheres. Such microspheres are formed from mineral additives during coal burning in furnaces of boiler units of electric power stations. Despite some reduction in exchange capacity per a mass unit of sorbents the latter have high kinetic characteristics that makes it possible to carry out the sorption process both in static and dynamic modes. Taking into account large industrial resources of microspheres as by-products of electric power stations, a comparative simplicity of the modification process, as well as good kinetic and capacitor characteristics, this class of sorbents can be considered promising enough for solving the problems of cleaning liquid radioactive wastes of various pollution levels. (authors)

  11. Methodology for calculating radiation doses from radioactivity released to the environment

    International Nuclear Information System (INIS)

    Killough, G.G.; McKay, L.R.

    1976-03-01

    This document represents a compilation of the principal environmental transport and dosimetry models developed, adapted, and implemented by the Radiological Analyses and Applications Group of the Environmental Sciences Division of the Oak Ridge National Laboratory. The transport of released radioactivity through the natural environment is discussed in four sections: atmospheric dispersion, resuspension of material by wind action, terrestrial transport, and movement of material in underground water seepage. The discussion of dose to man and biota is divided into internal and external exposure sections. And finally, a developmental model (CONDOS) which estimates the dose to a population resulting from the manufacture, storage, distribution, use, and disposal of consumer products which contain radioactivity is described. Numerous tables are included

  12. Method and apparatus for treating liquid contaminated with radioactive particulate solids

    International Nuclear Information System (INIS)

    Hirs, G.

    1976-01-01

    A method and apparatus reduces the amount of radioactive solids resulting from the filtration of particulate contaminants from liquid in a nuclear reactor plant. A filtration system includes a pre-filter comprising a sheet filter medium through which the reactor liquid passes to remove relatively large particulate contaminants for storage or disposal. The reactor liquid is then passed through a bed of granular filter medium to accumulate substantially all the previously non-filtered contaminants and thereby provide a clarified liquid suitable for reuse in the reactor. Backwash liquid is flowed through the granular filter bed to remove and entrain the accumulated contaminants into a slurry which is received by a reservoir where the slurry is maintained quiescently to settle the contaminants. Removal of liquid from the reservoir concentrates the contaminants for storage or further processing, without the necessity of large quantities of filter aids that would increase the quantity of storage-requiring contaminated solids

  13. Release of Radioactive Scrap Metal/Scrap Metal (RSM/SM) at Nevada Test Site (NTS)

    International Nuclear Information System (INIS)

    1993-01-01

    Reynolds Electrical and Engineering Company, Inc. (REECo) is the prime contractor to the US Department of Energy (DOE) in providing service and support for NTS operations. Mercury Base Camp is the main control point for the many forward areas at NTS, which covers 1,350 square miles. The forward areas are where above-ground and underground nuclear tests have been performed over the last 41 years. No metal (or other material) is returned to Mercury without first being tested for radioactivity. No radioactive metals are allowed to reenter Mercury from the forward areas, other than testing equipment. RAMATROL is the monitor check point. They check material in various ways, including swipe tests, and have a large assortment of equipment for testing. Scrap metal is also checked to address Resource Conservation and Recovery Act concerns. After addressing these issues, the scrap metals are categorized. Federal Property Management Regulations (FPMR) are followed by REECo. The nonradioactive scrap material is sold through the GSA on a scheduled basis. Radioactive scrap metal are presently held in forward areas where they were used. REECo has gained approval of their Nevada Test Site Defense Waste Acceptance Criteria, Certification, and Transfer Requirements, NVO-325 application, which will allow disposal on site, when RSM is declared a waste. The guideline that REECo uses for release limits is DOE Order 5480.11, Radiation Protection for Occupational Works, Attachment 2, Surface Radioactivity Guides, of this order, give release limits for radioactive materials. However, the removal of radioactive materials from NTS require approval by DOE Nevada Operations Office (DOE/NV) on a case-by-case basis. Requirements to consider before removal are found in DOE Order 5820.2A, Radioactive Waste Management

  14. Status of outdoor radioactive contamination at the Hanford Site

    International Nuclear Information System (INIS)

    McKinney, S.M.; Markes, B.M.

    1994-12-01

    This document summarizes the status of outdoor radioactive contamination near Hanford Site facilities and disposal sites. It defines the nature and areal extend of the radioactively contaminated areas and describes the historical, ongoing, and planned radiological monitoring and control activities. Radioactive waste has been disposed of to the soil column since shortly after the reactors and production facilities began operating. Radioactive liquid wastes were placed directly into the ground via liquid discharges to cribs, ponds, ditches, and reverse wells. Solid wastes were placed in trenches, burial vaults, and caissons. Although the Hanford Site covers 1,450 km 2 , the radioactively contaminated area is only about 36 km 2 or 2.5% of the original site. Over time, contamination has migrated from some of the waste management sites through various vectors (e.g., burrowing animals, deep-rooted vegetation, erosion, containment system failure) or has been deposited to the surface soil via spills and unplanned releases (e.g., line leaks/breaks, tank leaks, and stack discharges) and created areas of outdoor radioactivity both on and below the surface. Currently 26 km 2 are posted as surface contamination and 10 km 2 are posted as underground contamination

  15. Renewal of release and sampling authorizations of the Cattenom nuclear power plant. Complementary note on the authorizations demands of radioactive release; Renouvellement des autorisations de rejets et de prelevements de la centrale nucleaire de Cattenom. Note complementaire sur les demandes d'autorisation de rejets radioactifs

    Energy Technology Data Exchange (ETDEWEB)

    Marignac, Y.; Coeytaux, X

    2003-10-15

    The present note follows the exchanges with the contracting authority, the Cattenom nuclear power plant, during the public hearing stood on the 10. of october 2003 at Cattenom. In the frame of the evaluation of the inquiry file by WISE -Paris, this note constitutes a deepening on the question of authorization demands for the radioactive releases. The elements supplied by E.D.F in the file submitted to the inquiry, and the presentation of these demands and the information complements brought during the meeting of the 10. of october, raise different problems of evaluation and justification relative to the evolution of radioactive release of Cattenom. The note analyses in particular, the increasing of dosimetry impact expected from the progression of real release forecasted by the operator, and the weight linked to the feedback of clad failure problem. It discusses the justification brought by E.D.F. to the standing and even the increasing of liquid tritium release instead of the expected reduction, facing the arguments of the operator to the experience feedback on the park and to the uncertainties on the future progress. (N.C.)

  16. The initial phase of sudden releases of superheated liquid

    International Nuclear Information System (INIS)

    Schmidli, J.

    1994-04-01

    The catastrophic failure of a pressure vessel containing a liquefied substance, leading to an instantaneous release of its whole contents is considered as one of the major technological hazards. Due to the rapid depressurization caused by vessel failure, the fluid becomes superheated and unstable. Part of the fluid will evaporate using its internal energy and the two-phase mixture forming will be accelerated. This flashing process can be very violent, as experiments and incidents actually happened have shown. In the past, a number of dispersion models were developed to predict the history of an instantaneous release. In most of these models the source term is considered to be a gas volume at rest and not a rapidly expanding aerosol, as could be observed. Furthermore, it is usually assumed that all of the remaining fluid is entrained into the expanding cloud and nothing is deposited on the ground to form a pool. This work concentrates on the initial phase of the sudden release of superheated liquids with the aim to gain a better understanding of the flashing process and of the physical mechanisms involved, leading to a reliable prediction of the source term. Therefore, more than 400 experiments with propane, butane, refrigerant 12 and 114 were conducted. The experiments were initiated by shattering spherical glass flasks of different sizes. The main parameters varied were the liquid superheat and the filling level of the vessel. Using high-speed video and movie recordings and very fast responding measurement devices, it was possible to study the initial phase of such releases during which gravity plays no role. For sufficiently large released internal energy, the initial evolution of the release was always spherical with a constant radial expansion velocity during he first milliseconds until instabilities appeared at the surface of the droplet/vapor cloud that was formed. For all the experimental conditions, the fraction of the initial liquid falling on the ground

  17. Study and modelling of an innovative coprecipitation reactor for radioactive liquid wastes decontamination

    International Nuclear Information System (INIS)

    Flouret, Julie

    2013-01-01

    In order to decontaminate radioactive liquid wastes of low and intermediate levels, the coprecipitation is the process industrially used. The aim of this PhD work is to optimize the continuous process of coprecipitation. To do so, an innovative reactor is designed and modelled: the continuous reactor/classifier. Two model systems are studied: the coprecipitation of strontium by barium sulphate and the sorption of cesium by PPFeNi. The simulated effluent contains sodium nitrate in order to consider the high ionic strength of radioactive liquid wastes. First, each model system is studied on its own, and then a simultaneous treatment is performed. The kinetic laws of nucleation and crystal growth of barium sulphate are determined and incorporated into the coprecipitation model. Kinetic studies and sorption isotherms of cesium by PPFeNi are also performed in order to acquire the necessary data for process modelling. The modelling realised enables accurate prediction of the residual strontium and cesium concentrations according to the process used: it is a valuable tool for the optimization of existing units, but also the design of future units. The continuous reactor/classifier presents many advantages compared to the classical continuous process: the decontamination efficiency of strontium and cesium is highly improved while the volume of sludge generated by the process is reduced. A better liquid/solid separation is observed in the reactor/classifier and the global installation is significantly more compact. Thus, the radioactive liquid wastes treatment processes can be intensified by the continuous reactor/classifier, which represents a very promising technology for future industrial application. (author) [fr

  18. Risk-based prioritization for the interim remediation of inactive low-level liquid radioactive waste underground storage tanks at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    Chidambariah, V.; Travis, C.C.; Trabalka, J.R.; Thomas, J.K.

    1992-09-01

    The paper presents a risk-based approach for rapid prioritization of low-level liquid radioactive waste underground storage tanks (LLLW USTs), for possible interim corrective measures and/or ultimate closure. The ranking of LLLW USTs is needed to ensure that tanks with the greatest potential for adverse impact on the environment and human health receive top priority for further evaluation and remediation. Wastes from the LLLW USTs at Oak Ridge National Laboratory were pumped out when the tanks were removed from service. The residual liquids and sludge contain a mixture of radionuclides and chemicals. Contaminants of concern that were identified in the liquid phase of the inactive LLLW USTs include the radionuclides 90 Sr, 137 Cs, and 233 U and the chemicals carbon tetrachloride, trichloroethane, tetrachloroethene, methyl ethyl ketone, mercury, lead, and chromium. The risk-based approach for prioritization of the LLLW USTs is based upon three major criteria: (1) leaking characteristics of the tank, (2) location of the tanks, and (3) toxic potential of the tank contents. Leaking characteristics of LLLW USTs will aid in establishing the potential for the release of contaminants to environmental media. In this study, only the liquid phase was assumed to be released to the environment. Scoring criteria for release potential of LLLW USTs was determined after consideration of the magnitude of any known leaks and the tank type for those that are not known to leak

  19. Study of the degradation of liquid-organic radioactive wastes by electrochemical methods

    International Nuclear Information System (INIS)

    Hernandez A, J. I.

    2015-01-01

    In this study degradation studies were performed on blank samples, in which two electrochemical cells with different electrodes were used, the first is constituted by mesh electrodes Ti/Ir-Ta/Ti and the second by rod electrodes Ti/Ddb, using as reference an electrolytic medium of scintillation liquid and scintillation liquid more water, applying different potentials ranging from 1 to 25 V. After obtaining the benchmarks, the treatment was applied to samples containing organic liquid radioactive waste, in this case a short half-life radioisotope as Sulfur-35, the degradation characterization of organic compounds was performed in infrared spectrometry. (Author)

  20. The determination of cesium and rubidium in highly radioactive waste liquid

    International Nuclear Information System (INIS)

    Wei Songsheng

    1991-01-01

    Cesium and rubidium in high-level waste liquid were determined by atomic absorption spectrometry with the instrument modified for analyzing radioactive samples. The results show that the method is effective and safe. The error of the method is less than +- 3%, and it has been used in the production of cesium

  1. Thermal treatment of historical radioactive solid and liquid waste into the CILVA incinerator

    International Nuclear Information System (INIS)

    Deckers, Jan; Mols, Ludo

    2007-01-01

    Since the very beginning of the nuclear activities in Belgium, the incineration of radioactive waste was chosen as a suitable technique for achieving an optimal volume reduction of the produced waste quantities. Based on the 35 years experience gained by the operation of the old incinerator, a new industrial incineration plant started nuclear operation in May 1995, as a part of the Belgian Centralized Treatment/Conditioning Facility named CILVA. Up to the end of 2006, the CILVA incinerator has burnt 1660 tonne of solid waste and 419 tonne of liquid waste. This paper describes the type and allowable radioactivity of the waste, the incineration process, heat recovery and the air pollution control devices. Special attention is given to the treatment of several hundreds of tonne historical waste from former reprocessing activities such as alpha suspected solid waste, aqueous and organic liquid waste and spent ion exchange resins. The capacity, volume reduction, chemical and radiological emissions are also evaluated. BELGOPROCESS, a company set up in 1984 at Dessel (Belgium) where a number of nuclear facilities were already installed is specialized in the processing of radioactive waste. It is a subsidiary of ONDRAF/NIRAS, the Belgian Nuclear Waste Management Agency. According to its mission statement, the activities of BELGOPROCESS focus on three areas: treatment, conditioning and interim storage of radioactive waste; decommissioning of shut-down nuclear facilities and cleaning of contaminated buildings and land; operating of storage sites for conditioned radioactive waste. (authors)

  2. Radioactive sources of main radiological concern in the Kola-Barents region

    International Nuclear Information System (INIS)

    Bergman, R.; Baklanov, A.

    1998-07-01

    This overview focuses on some major issues for risk analysis appearing in our recent study surveying radioactive sources on the Kola Peninsula, along with adjacent parts of the Arctic seas. The main issues of the parts are as follows: An introduction to the presence of radioactive sources and environmental contamination in the Barents Euro-Arctic Region and the current status as regards various significant studies. Radioactive contamination in man and the environment on the Kola Peninsula, as well as radioactive transfer during the last three decades from external sources to the Kola-Barents region. The main conclusion from the findings is that the contamination is generally relatively low and that neither the activity levels in samples of soil, vegetation, and the important food-chains, nor the content in man indicate any changes since 1986 that could not be explained by the combined effect of the cumulative deposition from the nuclear weapons testing and the accident in Chernobyl. The radioactive sources of main concern in the region belong to the following categories: nuclear power submarine and cruiser naval bases; civil nuclear power ice-breaker fleet; building and repairing shipyards; nuclear power plants; radioactive waste and spent nuclear fuel storage facilities; sunken reactors/ships; liquid radioactive waste dumping; solid radioactive waste dumping; nuclear weapon bases; nuclear weapon tests; civil nuclear explosions; nuclear accidents; mining radioactive ore deposits and plants; new projects and others. Several case studies concerning releases in the Kola-Barents region are reviewed, and followed by consequence analyses for the categories of primary interest covering: a) airborne releases from the Kola NPP, and from submarines or spent nuclear fuel; b) releases from objects in the marine environment including submarines, dumped reactors, and various other radioactive objects and waste; c) releases from liquid and solid wastes stored on land or during

  3. Radioactive sources of main radiological concern in the Kola-Barents region

    Energy Technology Data Exchange (ETDEWEB)

    Bergman, R.; Baklanov, A

    1998-07-01

    This overview focuses on some major issues for risk analysis appearing in our recent study surveying radioactive sources on the Kola Peninsula, along with adjacent parts of the Arctic seas. The main issues of the parts are as follows: An introduction to the presence of radioactive sources and environmental contamination in the Barents Euro-Arctic Region and the current status as regards various significant studies. Radioactive contamination in man and the environment on the Kola Peninsula, as well as radioactive transfer during the last three decades from external sources to the Kola-Barents region. The main conclusion from the findings is that the contamination is generally relatively low and that neither the activity levels in samples of soil, vegetation, and the important food-chains, nor the content in man indicate any changes since 1986 that could not be explained by the combined effect of the cumulative deposition from the nuclear weapons testing and the accident in Chernobyl. The radioactive sources of main concern in the region belong to the following categories: nuclear power submarine and cruiser naval bases; civil nuclear power ice-breaker fleet; building and repairing shipyards; nuclear power plants; radioactive waste and spent nuclear fuel storage facilities; sunken reactors/ships; liquid radioactive waste dumping; solid radioactive waste dumping; nuclear weapon bases; nuclear weapon tests; civil nuclear explosions; nuclear accidents; mining radioactive ore deposits and plants; new projects and others. Several case studies concerning releases in the Kola-Barents region are reviewed, and followed by consequence analyses for the categories of primary interest covering: a) airborne releases from the Kola NPP, and from submarines or spent nuclear fuel; b) releases from objects in the marine environment including submarines, dumped reactors, and various other radioactive objects and waste; c) releases from liquid and solid wastes stored on land or during

  4. Assessment of the Potential Release of Radioactivity from Installations at AERE, Harwell. Implications for Emergency Planning

    Energy Technology Data Exchange (ETDEWEB)

    Flew, Elizabeth M.; Lister, B. A.J. [Atomic Energy Research Establishment, Harwell, Didcot, Berks. (United Kingdom)

    1969-10-15

    As part of a review of the site emergency organization at AERE, a realistic reappraisal has been made of the potential hazard to the establishment and district of all the buildings containing significant amounts of radioactive materials. To assess the amount of radioactive material which might be released under specified accident conditions, four factors have been applied to the total building stock: (a) the fraction of the stock involved, (b) the percentage conversion to an aerosol form, (c) release from primary containment and (d) release from the building. These potential release figures have been compared with ''site hazardous release'' figures for relevant radionuclides i.e., amounts which could give rise to exposure above defined reference levels. Those buildings with more than one tenth of the potential for causing; a site or district hazard have been designated ''risk buildings''. Such a quantitative assessment, although necessarily very approximate, has been of considerable value in determining the type and complexity of the central emergency organization required and in producing the right emphasis on exercises, training, liaison with outside bodies, etc. (author)

  5. Pyrohydrolytic separation technique for fluoride and chloride from radioactive liquid wastes

    International Nuclear Information System (INIS)

    Sawant, R.M.; Mahajan, M.A.; Shah, D.J.; Thakur, U.K.; Ramakumar, K.L.

    2009-01-01

    A rapid method for simultaneous determination of fluorine and chlorine in liquid samples with ion chromatography after pyrohydrolysis separation was proposed for routine analysis. The elements were separated from radioactive liquid sample by pyrohydrolysis and were subsequently determined with ion chromatography. Total time taken to determine these elements is about 45 min including 30 min for the pyrohydrolysis and 15 min for ion chromatography. The results of recovery tests ranged 95% or above. The limits of detection for F and Cl are 0.5 and 0.8 mgkg -1 , respectively. (author)

  6. Solidification of radioactive liquid wastes. A comparison of treatment options for spent resins and concentrates

    International Nuclear Information System (INIS)

    Roth, A.; Willmann, F.; Ebata, M.; Wendt, S.

    2008-01-01

    Ion exchange is one of the most common and effective treatment methods for liquid radioactive waste. However, spent ion exchange resins are considered to be problematic waste that in many cases require special approaches and pre-conditioning during its immobilization to meet the acceptance criteria for disposal. Because of the function that they fulfill, spent ion exchange resins often contain high concentrations of radioactivity and pose special handling and treatment problems. Another very common method of liquid radioactive waste treatment and water cleaning is the evaporation or diaphragm filtration. Both treatment options offer a high volume reduction of the total volume of liquids treated but generate concentrates which contain high concentrations of radioactivity. Both mentioned waste streams, spent resins as well as concentrates, resulting from first step liquid radioactive waste treatment systems have to be conditioned in a suitable manner to achieve stable waste products for final disposal. The most common method of treatment of such waste streams is the solidification in a solid matrix with additional inactive material like cement, polymer etc. In the past good results have been achieved and the high concentration of radioactivity can be reduced by adding the inactive material. On the other hand, under the environment of limited space for interim storage and the absence of a final repository site, the built-up of additional volume has to be considered as very critical. Moreover, corrosive effects on cemented drums during long-term interim storage at the surface have raised doubts about the long-term stability of such waste products. In order to avoid such disadvantages solidification methods have been improved in order to get a well-defined product with a better load factor of wastes in the matrix. In a complete different approach, other technologies solidify the liquid radioactive wastes without adding of any inactive material by means of drying

  7. Processing and monitoring liquid, radioactive effluents from the Wiederaufarbeitungsanlage Karlsruhe

    International Nuclear Information System (INIS)

    Hoehlein, G.; Huppert, K.L.; Winter, M.

    1977-01-01

    The Wiederaufarbeitungsanlage Karlsruhe (WAK) serves as a demonstration plant for the processing of highly-irradiated uranous oxide. The high active waste concentrates find interim storage at the WAK until they are solidified at a later stage. In contrast to this, the slightly- and the medium-active liquid wastes are transported to the decontamination facility of the Nuclear Research Centre Karlsruhe, where they are immediately processed. These liquid wastes contain about 1 per thousand of the activity inventary of the fuel elements processed. Monitoring of the radioactive waste water of the WAK is carried out by the Nuclear Research Centre's department radiation protection and safety. (orig.) [de

  8. US and Russian innovative technologies to process low-level liquid radioactive wastes: The Murmansk initiative

    International Nuclear Information System (INIS)

    Dyer, R.S.; Duffey, R.B.; Penzin, R.; Sorlie, A.

    1996-01-01

    This paper documents the status of the technical design for the upgrade and expansion to the existing Low-level Liquid Radioactive Waste (LLLRW) treatment facility in Murmansk, the Russian Federation. This facility, owned by the Ministry of Transportation and operated by the Russian company RTP Atomflot in Murmansk, Russia, has been used by the Murmansk Shipping Company (MSCo) to process low-level liquid radioactive waste generated by the operation of its civilian icebreaker fleet. The purpose of the new design is to enable Russia to permanently cease the disposal at sea of LLLRW in the Arctic, and to treat liquid waste and high saline solutions from both the Civil and North Navy Fleet operations and decommissioning activities. Innovative treatments are to be used in the plant which are discussed in this paper

  9. Risk assessment based on current release standards for radioactive surface contamination

    International Nuclear Information System (INIS)

    Chen, S.Y.

    1993-09-01

    Standards for uncontrolled releases of radioactive surface contamination have been in existence in the United States for about two decades. Such standards have been issued by various agencies, including the US Department of Energy. This paper reviews the technical basis of published standards, identifies areas in need of revision, provides risk interpretations based on current technical knowledge and the regulatory environment, and offers suggestions for improvements

  10. Decontamination and decommissioning of TAN radioactive liquid-waste-evaporator system (PM-2A). Final report

    International Nuclear Information System (INIS)

    Smith, D.L.

    1983-03-01

    This report describes the decontamination and decommissioning of the Test Area North (TAN) liquid waste evaporator (PM-2A). The PM-2A facility included the aboveground evaporator system, two underground holding tanks and feedlines, an electrical distribution subsystem, and one above ground concrete tank. Much surface soil of the PM-2A area was also radioactively contaminated. Stabilization of the liquid and sludge in the holding tanks, a major task, was achieved by pumping most of the liquid into 55-gal drums and mixing it with cement. The drums were buried in the Radioactive Waste Management Complex (RWMC). The remaining liquid and sludge were dried in place by layers of diatomaceous earth. The most contaminated surface soil was removed, and the area backfilled with clean topsoil and graded, reducing the surface radiation field to background. A 6-ft-high chain link fence now surrounds the area. Most of the area was seeded to crested wheatgrass. 46 figures, 9 tables

  11. The Assessment of Radioactive Liquid Waste Treatment Generated From The Fuel Reprocessing Plant Using Chemical Coagulation Method

    International Nuclear Information System (INIS)

    Kuncoro Arief, H; M Birmano, Dj

    1998-01-01

    Reprocessing of nuclear spent fuel produced 8 lot of radioactive liquid waste still bearing uranium and transuranium. The assessment of the radioactive liquid waste treatment with FeCI 3 as coagulant has been done. Decontamination factor and separation efficiency can be calculated from known activities of initial and post-treatment wastes. It can be concluded that some factors i.e. pH of treatment process, quantity of coagulant, mixing rate, and mixing time have influenced the treatment product

  12. Radioactivity level in the Raul Doamnei - Arges (Prundu Dam) ecosystem

    International Nuclear Information System (INIS)

    Toma, Al.; Pavelescu, M.; Margeanu, S.

    1997-01-01

    Hydrosphere is an important pathway of dispersion of the radionuclides released in the environment following nuclear activities. In this work the determination of the main pathway of radionuclide migration in the Raul Doamnei - Arges ecosystem (Prundu Dam) was carried out. The region is significant because it is a receptor of the radioactive liquid effluents resulting from Institute for Nuclear Research/Nuclear Fuel Plants compound activities. An evaluation of the source term (liquid effluents, deposition, soil erosion) and of natural and artificial radioactive isotope inventory in different zones of the ecosystem. Finally, the multi-compartmental radio-ecologic model applicable to this case is presented and an evaluation of the inter-compartment transfer parameters is given

  13. Radioactive waste management research at CEGB Berkeley nuclear laboratories

    International Nuclear Information System (INIS)

    Bradbury, D.

    1988-01-01

    The CEGB is the major electric utility in the United Kingdom. This paper discusses how, at the research laboratories at Berkeley (BNL), several programs of work are currently taking place in the radioactive waste management area. The theme running through all this work is the safe isolation of radionuclides from the environment. Normally this means disposal of waste in solid form, but it may also be desirable to segregate and release nonradioactive material from the waste to reduce volume or improve the solid waste characteristics (e.g., the release of liquid or gaseous effluents after treatment to convert the radioactivity to solid form). The fuel cycle and radioactive waste section at BNL has a research program into these aspects for wastes arising from the operation or decommissioning of power stations. The work is done both in-house and on contract, with primarily the UKAEA

  14. 5l/h pump for dosing corrosion radioactive liquids

    International Nuclear Information System (INIS)

    Przybylovich, S.; Shraer, V.; Chermak, R.

    1977-01-01

    The technical requirements, design and main technical characteristics of the pump for dosing corrosion and radioactive liquids with capacity up to 5 l/h are described. The design is based on the popular sixvertical split casing pump. The pump has four separate pump membrane type blocks with nonstraight hydraulic membrane control. The membranes are made of the cold worked CrNi(18/10)type stainless steel with thickness up to 0.1 mm and have the lifetime up to 3000 hours. The remote pump heads are used for pumping radioactive fluids when the pumping goes behind the safe wall, separating the pump from a hot lab. The tests showed that the pump secures the satisfactory accuracy of dozing and uniformity of pumping and that it is really possible to achieve the required life time of 10000 hours by this pump

  15. Analysis of 99Tc in the radioactive liquid waste after extraction into suitable solvent

    International Nuclear Information System (INIS)

    Sonar, N.L.; Vaishali De; Pardeshi, V.; Raghvendra, Y.; Valsala, T.P.; Sonavane, M.S.; Kulkarni, Y.; Raj Kanwar

    2012-01-01

    99 Tc is one of the long lived fission product with high fission yield. >From radioactive waste management point of view it is very much essential to evaluate the concentration of technetium in the radioactive liquid waste in order to finalise the treatment process to extract/isolate it from the stream which is discharged to the environment. For the estimation of 99 Tc in the radioactive liquid waste stream, extraction of the stable complex of technetium-tetraphenyl arsonium chloride (TPAC) into chloroform followed by beta counting was studied. Various parameters like pH, time of equilibration, concentration of TPAC in chloroform, use of other solvent for extraction as well as interference of various other radionuclides present in the waste were also studied. The radioactive liquid waste being handled in plant contains high concentrations of salts in the form of sodium nitrate. Hence effect of salt concentration on the percentage extraction was also evaluated. The extraction behavior does not dependent on change in the pH of the solution. Almost 99.5% extraction was observed in the pH range of 1-13.0. High concentration of salt is affecting the extraction. However, this can be taken care by diluting the radioactive waste. It takes almost 90 min time for maximum extraction. Presence of radionuclides like 137 Cs, 90 Sr are not interfering the extraction of 99 Tc. However, 106 Ru is getting slightly extracted along with 99 Tc. The error due to 106 Ru can be eliminated by taking gamma spectrum and deducting the activity from the total beta activity to get 99 Tc activity. Nitrobenzene can be used for extraction of Tc-TPAC complex in place of chloroform. (author)

  16. Solidification of radioactive liquid wastes, Treatment options for spent resins and concentrates - 16405

    International Nuclear Information System (INIS)

    Roth, Andreas

    2009-01-01

    Ion exchange is one of the most common and effective treatment methods for liquid radioactive waste. However, spent ion exchange resins are considered to be problematic waste that in many cases require special approaches and pre-conditioning during its immobilization to meet the acceptance criteria for disposal. Because of the function that they fulfill, spent ion exchange resins often contain high concentrations of radioactivity and pose special handling and treatment problems. Another very common method of liquid radioactive waste treatment and water cleaning is the evaporation or diaphragm filtration. Both treatment options offer a high volume reduction of the total volume of liquids treated but generate concentrates which contain high concentrations of radioactivity. Both mentioned waste streams, spent resins as well as concentrates, resulting from first step liquid radioactive waste treatment systems have to be conditioned in a suitable manner to achieve stable waste products for final disposal. Spent resin and concentrate treatment often appear as a specific task in decommissioning projects, because in the past those waste streams typically had been stored in tanks for the lifetime of the plant and needs to be retrieved, conditioned and packed prior to dismantling activities. Additionally a large amount of contaminated liquids will be generated by utilizing decontamination processes and needs to be processed further on. Such treatment options need to achieve waste products acceptable for final disposal, because due to the closure of the site no interim storage can be envisaged. The most common method of treatment of such waste streams is the solidification in a solid matrix with additional inactive material like cement, polymer etc. In the past good results have been achieved and the high concentration of radioactivity can be reduced by adding the inactive material. On the other hand, under the environment of limited space for interim storage and the absence

  17. Equipment experience in a radioactive LFCM [liquid-fed ceramic melter] vitrification facility

    International Nuclear Information System (INIS)

    Holton, L.K. Jr.; Dierks, R.D.; Sevigny, G.J.; Goles, R.W.; Surma, J.E.; Thomas, N.M.

    1986-11-01

    Since October 1984, the Pacific Northwest Laboratory (PNL) has operated a pilot-scale radioactive liquid-fed ceramic melter (RLFCM) vitrification process in shielded manipulator hot cells. This vitrification facility is being operated for the Department of Energy (DOE) to remotely test vitrification equipment components in a radioactive environment and to develop design and operation data that can be applied to production-scale projects. This paper summarizes equipment and process experience obtained from the operations of equipment systems for waste feeding, waste vitrification, canister filling, canister handling, and vitrification off-gas treatment

  18. Development of a universal solvent for the decontamination of acidic liquid radioactive wastes

    Science.gov (United States)

    Todd, T. A.; Brewer, K. N.; Law, J. D.; Wood, D. J.; Herbest, R. S.; Romanovskiy, V. N.; Esimantovskiy, V. M.; Smirnov, I. V.; Babain, V. A.

    1999-01-01

    A teritiary solvent containing chlorinated cobalt dicarbollide, polyethylene glycol and diphenylcarbamoylmethylphosphine oxide was evaluated in different non-nitroaromatic diluents for the separation of cesium, strontium, actinides and rare earth elements from acidic liquid radioactive waste. Decontamination factors of >95% for Cs, 99.7% for Sr, and 99.99% for actinides were achieved in four successive batch contacts using actual radioactive waste. Pilot plant testing in centrifugal contactors using simulated wastes, has demonstrated removal of >99% of all targeted ions.

  19. Bottom loaded filter for radioactive liquids

    International Nuclear Information System (INIS)

    Wendland, W.G.

    1980-01-01

    A bottom loaded filter assembly for filtering radioactive liquids through a replaceable cartridge filter is disclosed. The filter assembly includes a lead-filled jacket enveloping a housing having a chamber therein for the filter cartridge. A track arrangement carries a hatch for sealing the chamber. A spacer plug supports the cartridge within guide means associated with the inlet conduit in the chamber. The plug and cartridge drop out of the chamber when the hatch is unbolted and move laterally of the chamber. During cartridge replacement, a new plug and cartridge are supported in the guide means by a spacer bar inserted across the track means under the chamber. The hatch is then slid under the chamber and bolted to the vessel, engaging an o-ring to seal the chamber

  20. Hydrogen gettering the overpressure gas from highly radioactive liquids

    International Nuclear Information System (INIS)

    Riley, D.L.; Schicker, J.R.

    1996-04-01

    Remediation of current inventories of high-activity radioactive liquid waste (HALW) requires transportation of Type-B quantities of radioactive material, possibly up to several hundred liters. However, the only currently certified packaging is limited to quantities of 50 ml (0.01 gal) quantities of Type-B radioactive liquid. Efforts are under way to recertify the existing packaging to allow the shipment of up to 4 L (1.1 gal) of Type-B quantities of HALW, but significantly larger packaging could be needed in the future. Scoping studies and preliminary designs have identified the feasibility of retrofitting an insert into existing casks, allowing the transport of up to 380 L (100 gal) of HALW. However, the insert design and ultimate certification strategy depend heavily on the gas-generating attributes of the HALW. A non-vented containment vessel filled with HALW, in the absence of any gas-mitigation technologies, poses a deflagration threat and, therefore, gas generation, specifically hydrogen generation, must be reliably controlled during all phases of transportation. Two techniques are available to mitigate hydrogen accumulation: recombiners and getters. Getters have an advantage over recombiners in that oxides are not required to react with the hydrogen. A test plan was developed to evaluate three forms of getter material in the presence of both simulated HALW and the gases that are produced by the HALW. These tests demonstrated that getters can react with hydrogen in the presence of simulated waste and in the presence of several other gases generated by the HALW, such as nitrogen, ammonia, nitrous oxide, and carbon monoxide. Although the use of such a gettering system has been shown to be technically feasible, only a preliminary design for its use has been completed. No further development is planned until the requirement for bulk transport of Type-B quantities of HALW is more thoroughly defined

  1. Models and methods to evaluate consequences of the release of airborne radioactivity from NNPs

    International Nuclear Information System (INIS)

    Anh, T.H.

    1989-01-01

    To examine the radioactive contamination and possible consequences of a nuclear power plant on living organisms during its operation periodes, the computer programmes were elaborated for assessing its fluences on the environment. The authors have resolved the following problems: i) Calculation of fission product inventories in the reactor core; ii) Calculation of the atmospheric dispersion of the released radionuclides under the meteorological conditions as well as the deposition of the radioactive substances on the soil; iii) Calculation of the irradiation doses

  2. Radioactive waste disposal on a non-industrial scale

    International Nuclear Information System (INIS)

    1990-01-01

    A 13 minute videotape deals with the following points: 1) Exposure pathways for solid, liquid and gaseous effleunt; 2) Critical pathways; 3) Critical groups; 4) Controlling authorities; 5) Principles of disposal, including a) concentrate and contain or b) delay and decay or c) dilute and disperse and 6) record keeping. The possible effects on Man and the Environment, of the release of radioactive wastes are discussed, and the principles underlying safe disposal of such wastes are explained. There are illustrations of procedures used in Imperial College for dealing with both high and low activity waste, and methods suitable for disposal of solid, liquid and gaseous forms are described. The programme gives a useful introduction to an important aspect of work with radioactive materials, but is only intended as a supplement to practical training. (author)

  3. Simulation Studies of Diffusion-Release and Effusive-Flow of Short-Lived Radioactive Isotopes

    CERN Document Server

    Zhang, Yan; Kawai, Yoko

    2005-01-01

    Delay times associated with diffusion release from targets and effusive-flow transport of radioactive isotopes to ion sources are principal intensity limiters at ISOL-based radioactive ion beam facilities, and simulation studies with computer models are cost effective methods for designing targets and vapor transport systems with minimum delay times to avoid excessive decay losses of short lived ion species. A finite difference code, Diffuse II, was recently developed at the Oak Ridge National Laboratory to study diffusion-release of short-lived species from three principal target geometries. Simulation results are in close agreement with analytical solutions to Fick’s second equation. Complementary to the development of Diffuse II, the Monte-Carlo code, Effusion, was developed to address issues related to the design of fast vapor transport systems. Results, derived by using Effusion, are also found to closely agree with experimental measurements. In this presentation, the codes will be used in conc...

  4. Maxine: A spreadsheet for estimating dose from chronic atmospheric radioactive releases

    Energy Technology Data Exchange (ETDEWEB)

    Jannik, Tim [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Bell, Evaleigh [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, Kenneth [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-24

    MAXINE is an EXCEL© spreadsheet, which is used to estimate dose to individuals for routine and accidental atmospheric releases of radioactive materials. MAXINE does not contain an atmospheric dispersion model, but rather doses are estimated using air and ground concentrations as input. Minimal input is required to run the program and site specific parameters are used when possible. Complete code description, verification of models, and user’s manual have been included.

  5. Decontamination factor Improvement and Waste Reduction of Full-scaled Evaporation System for Liquid Radioactive Waste Treatment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Tae; Ju, Young Jong; Seol, Jeung Gun; Cho, Nam Chan [KNF, Daejeon (Korea, Republic of); Ha, Dong Hwan; Kim, Yun Kwan [Jeontech Co., Suwon (Korea, Republic of)

    2016-05-15

    Liquid radioactive waste is produced from nuclear power plants, nuclear research centers, radiopharmaceuticals and nuclear fuel fabrication plants, etc. Ion-exchange, chemical precipitation, evaporation, filtration, liquid/solid extraction and centrifugal are applied to treat the liquid waste. Chemical precipitation requires low capital and operation cost. However, it produces large amount of secondary waste and has low DF (decontamination factor). Evaporation process removes variety of radionuclides in high DF. But, it also has problems in scaling and foaming [3, 4]. In this study, it is investigated that the effect of switching lime precipitation and centrifugal processes to evaporation system for improvement of removal efficiency and decrease of waste in full-scaled radioactive wastewater treatment plant. By swapping full-scaled wastewater treatment system from the centrifugal and the lime precipitation to the evaporator and the crystallizer in the nuclear fuel fabrication plant, it was possible to increase removal efficiency and to minimize waste productivity. Radioactivity concentration of effluent is decreased from 0.01 Bq/mL to ND level. Besides, waste production was reduced from 15 drums/yr to 2 drums/yr (87%).

  6. Collective dose estimates by the marine food pathway from liquid radioactive wastes dumped in the Sea of Japan

    International Nuclear Information System (INIS)

    Togawa, O.; Povinec, P.P.; Pettersson, H.B.L.

    1999-01-01

    IAEA-MEL has been engaged in an assessment programme related to radioactive waste dumping by the former USSR and other countries in the western North Pacific Ocean and its marginal seas. This paper focuses on the Sea of Japan and on estimation of collective doses from liquid radioactive wastes. The results from the Japanese-Korean-Russian joint expeditions are summarized, and collective doses for the Japanese population by the marine food pathway are estimated from liquid radioactive wastes dumped in the Sea of Japan and compared with those from global fallout and natural radionuclides. The collective effective dose equivalents by the annual intake of marine products caught in each year show a maximum a few years after the disposals. The total dose from all radionuclides reaches a maximum of 0.8 man Sv in 1990. Approximately 90% of the dose derives from 137Cs, most of which is due to consumption of fish. The total dose from liquid radioactive wastes is approximately 5% of that from global fallout, the contribution of which is below 0.1% of that of natural 210Po

  7. Method and apparatus for evaporating radioactive liquid and calcinating the residue

    International Nuclear Information System (INIS)

    Sridhar, T.S.

    1984-01-01

    This invention provides an apparatus and a process for evaporating liquid wastes and calcining the residue. The liquid is sprayed against a hollow, rotating heated cylinder within a casing. The dried residue is scraped from the rotating cylinder and released through a valve at the bottom of the casing, while the effluent gas is filtered

  8. Simplified analytical model to simulate radionuclide release from radioactive waste trenches

    International Nuclear Information System (INIS)

    Sa, Bernardete Lemes Vieira de

    2001-01-01

    In order to evaluate postclosure off-site doses from low-level radioactive waste disposal facilities, a computer code was developed to simulate the radionuclide released from waste form, transport through vadose zone and transport in the saturated zone. This paper describes the methodology used to model these process. The radionuclide released from the waste is calculated using a model based on first order kinetics and the transport through porous media was determined using semi-analytical solution of the mass transport equation, considering the limiting case of unidirectional convective transport with three-dimensional dispersion in an isotropic medium. The results obtained in this work were compared with other codes, showing good agreement. (author)

  9. Radioactive effluents from CANDU 6 reactors during normal operation

    International Nuclear Information System (INIS)

    Boss, C.R.; Allsop, P.J.

    1995-12-01

    During routine operation of a CANDU 6 reactor, various gaseous, liquid, and solid radioactive wastes are generated. The layout of the CANDU 6 reactor and the design of its systems ensure that these are minimized, but small quantities of gaseous and liquid wastes are continually discharged at very low concentrations. This report discusses the make-up of these chronically generated gaseous and liquid effluents. From a safety perspective, the doses to individual members of the public resulting from radioactive wastes chronically discharged from CANDU 6 reactors have been negligible. Similarly, doses to the regional and global populations have been negligible, generally less than 0.001% of background. While far below regulatory limits, releases of tritium, noble gases and gross β - -γ have been the most radiologically significant emissions, while radioiodine and particulates have had the greatest potential to deliver public dose. (author). 8 refs., 16 tabs., 3 figs

  10. Development of a technology and a pilot plant for treatment of small volumes of liquid radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, I G; Gradev, G D [Bulgarian Academy of Sciences, Sofia (Bulgaria). Inst. for Nuclear Research and Nuclear Energy

    1997-02-01

    The development of technology for treatment of liquid radioactive waste is described. Waste arisings are estimated. Liquid wastes of concern are mainly low active wastes according to the Bulgarian legislation. The activity is determined by the presence of {sup 134}Cs, {sup 137}Cs, {sup 60}Co, {sup 90}Sr, {sup 144}Ce, {sup 65}Zc, {sup 54}Mn, {sup 110m}Ag. Different precipitation processes are compared. The mixed iron hydroxide - calcium phosphate precipitation is determined as suitable for decontamination of the liquid radioactive waste. Effective decontamination is achieved when precipitation is followed by ion exchange. Additional increase of the decontamination is possible when sorbents are added during the precipitation step. The sorption and desorption of radionuclides on zeolites are studied. Cement solidification and thermal treatment of zeolites are studied for immobilization of radioactive material from precipitation and ion exchange. Both methods produce stable waste forms suitable for containment of the radionuclides. (author). 17 refs, 3 figs, 12 tabs.

  11. The state of radioactive waste management and of personnel radiation exposure in nuclear power generating facilities in fiscal 1983

    International Nuclear Information System (INIS)

    1985-01-01

    (1) The state of radioactive waste management in nuclear power generating facilities: In the nuclear power stations, the released quantities of radioactive gaseous and liquid wastes are all below the control objective levels. For the respective nuclear power stations, the released quantities of radioactive gaseous and liquid wastes in fiscal 1983 and the objective levels are given in table. And, the quantities of solid wastes taken into storage and the cumulative amounts are given. For reference, the results each year since fiscal 1974 are shown. (2) The state of personnel radiation exposure in nuclear power generating facilities: In the nuclear power stations, the personnel radiation exposures are all below the permissible levels. The dose distribution etc. in the respective nuclear power stations are given in table. For reference, the results each year since fiscal 1974 are shown. (Mori, K.)

  12. Comparison among the rice bark in the raw and active forms in the removal of 241Am and 137Cs from liquid radioactive wastes

    International Nuclear Information System (INIS)

    Ferreira, Rafael V.P.; Lima, Josenilson B. de; Bellini, Maria Helena; Sakata, Solange Kazumi; Marumo, Julio Takehiro

    2011-01-01

    New techniques involving treatment of radioactive wastes which associate simplicity and low cost have been directed the attention for the bio sorption, which is a process were solid vegetable or micro-organism for the retention, removing, or recovering of heavy metals from a liquid environment. This study evaluated the capacity of a bio sorbent to remove Am-241 and Cs-137 from liquid radioactive waste. The chosen material was the rice bark employed in the raw or activated forms. The obtained results suggest that the bio sorption, with the activated rice bark, can be a viable technique for the treatment of liquid radioactive wastes containing Am-241 and Cs-137 present in liquid radioactive wastes

  13. Effectiveness of liquid radioactive waste purification by inorganic granulated sorbents

    International Nuclear Information System (INIS)

    Komarevskij, V.M.; Stepanets, O.V.; Sharygin, L.M.; Matveev, S.A.

    1995-01-01

    Study results on purification of simulative and real liquid radioactive wastes from fission products radionuclides and by inorganic corrosion-nature sorbents 'Thermoxide' are presented. Properties by sorption of cesium, strontium and cobalt are studied; results of experiments on purification of weakly-salted water solutions (waste waters, ships drainage tanks, showers and laundries) of the Beloyarsk NPP are presented. Sorbents source characteristics are determined. 4 refs., 2 figs., 3 tabs

  14. Radioactive release from VVER-1000 reactors after a terror attack

    International Nuclear Information System (INIS)

    Sdouz, G.

    2005-01-01

    Full text: One of the terror scenarios for nuclear power plants is a severe damage of the reactor containment caused by a plane crash or a missile. Due to the loss of electric power the cooling of the core is not maintained leading to a core melt accident. Normally in the course of severe accidents an intact containment has the ability to retain a large part of the radioactive inventory. The goal of this work is the investigation of the behavior of the radioactive release from a VVER-1000-type reactor during a severe accident with a large containment leak from the beginning of the accident. The results are compared with the release in a severe accident via a very small leakage due to the untightness of the containment. This work supplements a series of studies investigating the behavior of a VVER-1000-type reactor during severe accidents under different accident management strategies. The focus in this study is on the 'station blackout'-sequence (or TMLB' in the WASH-1400 nomenclature). The calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. Up to the melt-through of the cavity bottom the thermal-hydraulics phenomena are almost identical to the TMLB'-case with an intact containment from the beginning. The phenomena occur slightly delayed due to the large containment leak. When the core-concrete-interaction begins the resulting gases leave the containment through the large leak and do not cause a pressure increase. The containment pressure remains at ambient pressure. Due to the different behavior and to the different release times of the nuclides the deviations to the scenario with an intact containment show a great variety. From this comparison it can be shown that the intact containment retains the nuclides up to a factor of 6000. (author)

  15. Removal of some Fission Products from Low Level Liquid Radioactive Waste by Chemical Precipitation liquid/Co-precipitation / Phosphate Coagulant

    International Nuclear Information System (INIS)

    Borai, E.H.; Attallah, M.F.; Hilal, M.A.; Abo-Aly, M.M.; Shehata, F.A.

    2008-01-01

    In Egypt radioactive waste has been generated from various uses of radioactive materials. Presence of cesium demonstrated a major problem from the removal point of view even by conventional and advanced technologies. Selective chemical precipitation has been oriented for removal of some fission products including 137 Cs from low level liquid radioactive waste (LLLRW). The aim of the present study was focused to investigate the effectiveness of various phosphate compounds that improved the precipitation process and hence the decontamination factor. The results showed that, maximum removal of 137 Cs reaching 46.4 % using di-sodium hydrogen phosphate as a selective coagulant. It was found that significant enhancement of co-precipitation of 137 Cs (62.5 %) was obtained due to presence of Nd 3+ in the LLLRW

  16. Order of 4 august 1989 on licensing the release of gaseous radioactive effluents by the Cattenom nuclear production centre (units 3 and 4)

    International Nuclear Information System (INIS)

    1989-08-01

    This Order fixes the conditions and limits of authorised releases of gaseous radioactive effluents from Units 3 and 4 of the Cattenom nuclear power plant. It specifies these are maximum limits, below which the radioactive releases should be as low as possible [fr

  17. OPS liquid pathway generic study. Topical report No. 22A60

    International Nuclear Information System (INIS)

    1977-06-01

    An evaluation of the consequences of radioactivity released from an offshore nuclear plant to liquid pathways as a result of postulated accidents more severe than the design basis Loss-of-Coolant Accident is presented

  18. Proposal of the concept of selection of accidents that release large amounts of radioactive substances in the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Ono, Masato; Honda, Yuki; Takada, Shoji; Sawa, Kazuhiro

    2015-01-01

    In Position, construction and equipment of testing and research reactor to be subjected to the use standards for rules Article 53 (prevention of expansion of the accident to release a large amount of radioactive material) generation the frequency is a lower accident than design basis accident, when what is likely to release a large amount of radioactive material or radiation from the facility has occurred, and take the necessary measures in order to prevent the spread of the accident. There is provided a lower accident than frequency design basis accidents, for those that may release a large amount of radioactive material or radiation. (author)

  19. Small-scale experimental study of vaporization flux of liquid nitrogen released on water.

    Science.gov (United States)

    Gopalaswami, Nirupama; Olewski, Tomasz; Véchot, Luc N; Mannan, M Sam

    2015-10-30

    A small-scale experimental study was conducted using liquid nitrogen to investigate the convective heat transfer behavior of cryogenic liquids released on water. The experiment was performed by spilling five different amounts of liquid nitrogen at different release rates and initial water temperatures. The vaporization mass fluxes of liquid nitrogen were determined directly from the mass loss measured during the experiment. A variation of initial vaporization fluxes and a subsequent shift in heat transfer mechanism were observed with changes in initial water temperature. The initial vaporization fluxes were directly dependent on the liquid nitrogen spill rate. The heat flux from water to liquid nitrogen determined from experimental data was validated with two theoretical correlations for convective boiling. It was also observed from validation with correlations that liquid nitrogen was found to be predominantly in the film boiling regime. The substantial results provide a suitable procedure for predicting the heat flux from water to cryogenic liquids that is required for source term modeling. Copyright © 2015 Elsevier B.V. All rights reserved.

  20. Bottom loaded filter for radioactive liquid

    International Nuclear Information System (INIS)

    Wendland, W.G.

    1980-01-01

    A specification is given for a bottom loaded filter assembly for filtering radioactive liquids through a replaceable cartridge filter, which includes a lead-filled jacket enveloping a housing having a chamber therein for the filter cartridge. A track arrangement carries a hatch for sealing the chamber. A spacer plug supports the cartridge within guide means associated with the inlet conduit in the chamber. The plug and cartridge drop out of the chamber when the hatch is unbolted and moved laterally of the chamber along the track. During cartridge replacement a new plug and cartridge are supported in the guide means by a spacer bar inserted across the track means under the chamber. The hatch is then slid under the chamber and bolted to a flange on the housing, engaging an O-ring to seal the chamber. (author)

  1. Source term estimation and the isotopic ratio of radioactive material released from the WIPP repository in New Mexico, USA

    International Nuclear Information System (INIS)

    Thakur, P.

    2016-01-01

    After almost 15 years of operations, the Waste Isolation Pilot Plant (WIPP) had one of its waste drums breach underground as a result of a runaway chemical reaction in the waste it contained. This incident occurred on February 14, 2014. Moderate levels of radioactivity were released into the underground air. A small portion of the contaminated underground air also escaped to the surface through the ventilation system and was detected approximately 1 km away from the facility. According to the source term estimation, the actual amount of radioactivity released from the WIPP site was less than 1.5 mCi. The highest activity detected on the surface was 115.2 μBq/m 3 for 241 Am and 10.2 μBq/m 3 for 239+240 Pu at a sampling station located 91 m away from the underground air exhaust point and 81.4 μBq/m 3 of 241 Am and 5.8 μBq/m 3 of 239+240 Pu at a monitoring station located approximately 1 km northwest of the WIPP facility. The dominant radionuclides released were americium and plutonium, in a ratio that matches the content of the breached drum. Air monitoring across the WIPP site intensified following the first reports of radiation detection underground to determine the extent of impact to WIPP personnel, the public, and the environment. In this paper, the early stage monitoring data collected by an independent monitoring program conducted by the Carlsbad Environmental Monitoring & Research Center (CEMRC) and an oversight monitoring program conducted by the WIPP's management and operating contractor, the Nuclear Waste Partnership (NWP) LLC were utilized to estimate the actual amount of radioactivity released from the WIPP underground. The Am and Pu isotope ratios were measured and used to support the hypothesis that the release came from one drum identified as having breached that represents a specific waste stream with this radionuclide ratio in its inventory. This failed drum underwent a heat and gas producing reaction that overpowered its vent and

  2. Treatment of low level radioactive liquid waste containing appreciable concentration of TBP degraded products.

    Science.gov (United States)

    Valsala, T P; Sonavane, M S; Kore, S G; Sonar, N L; De, Vaishali; Raghavendra, Y; Chattopadyaya, S; Dani, U; Kulkarni, Y; Changrani, R D

    2011-11-30

    The acidic and alkaline low level radioactive liquid waste (LLW) generated during the concentration of high level radioactive liquid waste (HLW) prior to vitrification and ion exchange treatment of intermediate level radioactive liquid waste (ILW), respectively are decontaminated by chemical co-precipitation before discharge to the environment. LLW stream generated from the ion exchange treatment of ILW contained high concentrations of carbonates, tributyl phosphate (TBP) degraded products and problematic radio nuclides like (106)Ru and (99)Tc. Presence of TBP degraded products was interfering with the co-precipitation process. In view of this a modified chemical treatment scheme was formulated for the treatment of this waste stream. By mixing the acidic LLW and alkaline LLW, the carbonates in the alkaline LLW were destroyed and the TBP degraded products got separated as a layer at the top of the vessel. By making use of the modified co-precipitation process the effluent stream (1-2 μCi/L) became dischargeable to the environment after appropriate dilution. Based on the lab scale studies about 250 m(3) of LLW was treated in the plant. The higher activity of the TBP degraded products separated was due to short lived (90)Y isotope. The cement waste product prepared using the TBP degraded product was having good chemical durability and compressive strength. Copyright © 2011 Elsevier B.V. All rights reserved.

  3. The state of radioactive waste management and personnel radiation exposure in commercial nuclear power plants in fiscal 1981

    International Nuclear Information System (INIS)

    1982-01-01

    Radioactive waste management: The owners of commercial nuclear power plants are obligated to control the release of gaseous and liquid radioactive wastes below the objective release levels, and to store solid wastes in containers on the site. As for the former, the released (radioactive) quantity in fiscal 1981 (from April, 1981, to March, 1982,) together with the objective levels are given for respective nuclear power stations; and as for the latter, the stored quantity and also the cumulative quantity up to the year are given. Radiation exposure: The owners of commercial nuclear power plants are obligated to control the personnel exposure below the permissible level. The personnel exposure dose in fiscal 1981 is given for respective nuclear power station. (Mori, K.)

  4. An update on radioactive release and exposures after the Fukushima Dai-ichi nuclear disaster.

    LENUS (Irish Health Repository)

    McLaughlin, P D

    2012-09-01

    On 11 March 2011, the Richter scale 0.9-magnitude Tokohu earthquake and tsunami struck the northeast coast of Japan, resulting in widespread injury and loss of life. Compounding this tragic loss of life, a series of equipment and structural failures at the Fukushima Dai-ichi nuclear power plant (FDNP) resulted in the release of many volatile radioisotopes into the atmosphere. In this update, we detail currently available evidence about the nature of immediate radioactive exposure to FDNP workers and the general population. We contrast the nature of the radioactive exposure at FDNP with that which occurred at the Chernobyl power plant 25 years previously. Prediction of the exact health effects related to the FDNP release is difficult at present and this disaster provides the scientific community with a challenge to help those involved and to continue research that will improve our understanding of the potential complications of radionuclide fallout.

  5. Method of conversion of caustic liquid radioactive wastes containing sodium nitrates into solid insoluble products

    International Nuclear Information System (INIS)

    Barney, G.S.; Brownell, L.E.

    1975-01-01

    A proposal is made to convert caustic, liquid, radioactive wastes containing sodium nitrate into a solid product by reaction with powdered aluminium silicate at temperatures between 30 0 and 100 0 C, which is practically insoluble (10 -7 to 10 -10 g/cm 2 -day) and is thermally stable. A cancrinite is formed which binds the radioactive salts in the cage-like structure of its crystal lattice. The method is also suitable for liquid wastes from the Purex method as well as for wastes containing fission products of Cs 137 and Sr 90 in concentrations of 0.37 M to 0.01 M. Numerous detailed examples explain the invention. (UW/LH) [de

  6. Isopleths of surface concentration and surface exposure rate due to a radioactive cloud released from a stack

    International Nuclear Information System (INIS)

    Kobayashi, Hideo; Yabuta, Hajimu; Katagiri, Hiroshi; Obata, Kazuichi; Kokubu, Morinobu

    1982-03-01

    Various calculations are made to estimate the distributions of concentration and γ-exposure rate due to a radioactive cloud released from a point source to the atmosphere. In this report, the isopleths of concentration and γ-exposure rate which were calculated are given in graphs to enable rapid prediction of the influence of released radioactive material in the emergency situation. Recently there are facilities which are equipped with a system to display the calculation results on CRT; but such practice is rather rare. By placing the calculated isopleths of reduction scale 1/25000 or 1/50000 on the usual map, any facilities without the CRT system can readily estimate the influence of an accidental release. The graphs of isopleths are given with the release height (11 values of 0 to 200 m at about 20 m intervals) and the atmospheric stability (6 classes) as parameters. Calculations of γ-exposure rates were made using the computer code GAMPUL developed by T. Hayashi and T. Shiraishi. In the calculation of radioactive concentrations and γ-exposure rates, the vertical diffusion depths, σsub(z), exceeding 1000 m are taken to be 1000 m according to the Meteorological Guide for the Safety Analysis of Power Reactor (J.AEC). The comparison between with and without this limitation in σsub(z) is made in the case of downwind axial surface distributions. (author)

  7. Sorption of radioscesium from liquid radioactive waste on clay and immobilization by baking the clay at elevated temperature

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, F.; Ghaffar, A. [Pakistan Inst. of Nuclear Science and Technology, Islamabad (Pakistan)

    2011-07-01

    The cesium-137 is the most problematic radionuclide in the radioactive wastes. It belongs to the IA group of the periodic table, highly reactive towards water and has very high mobility. Due to beta and gamma radiation hazards of radiocesium its decontamination and disposal requires some special tools and techniques. In this study globules of clay material was used for the removal of cesium from low level liquid radioactive wastes and further processed for immobilization. The aim of this study was to assess the solidification and immobilization of secondary waste. The secondary waste, after sorption of cesium from the liquid radioactive waste generated at this institute, was found compatible to the cement matrix used for the cementation process. The procedure for immobilization of low level radioactive waste with cementation using vitreous clay material as an additive was developed. (orig.)

  8. Sorption of radioscesium from liquid radioactive waste on clay and immobilization by baking the clay at elevated temperature

    International Nuclear Information System (INIS)

    Rashid, F.; Ghaffar, A.

    2011-01-01

    The cesium-137 is the most problematic radionuclide in the radioactive wastes. It belongs to the IA group of the periodic table, highly reactive towards water and has very high mobility. Due to beta and gamma radiation hazards of radiocesium its decontamination and disposal requires some special tools and techniques. In this study globules of clay material was used for the removal of cesium from low level liquid radioactive wastes and further processed for immobilization. The aim of this study was to assess the solidification and immobilization of secondary waste. The secondary waste, after sorption of cesium from the liquid radioactive waste generated at this institute, was found compatible to the cement matrix used for the cementation process. The procedure for immobilization of low level radioactive waste with cementation using vitreous clay material as an additive was developed. (orig.)

  9. Evaluation of radioactive wastes in Instituto de Energia Atomica (Sao Paulo-Brazil)

    International Nuclear Information System (INIS)

    Sawakuchi, R.S.

    1978-01-01

    An evaluation of present and future production of radioactive waste in several departments of the Instituto de Energia Atomica has been done. Taking into account this evaluation, the criteria for disposal and convenient treatment technique have been studied. The most critical form of liquid radioactive waste is that of 131 I processing because high concentration of radiotellurium always accompanies this form of waste. Ion exchange and precipitation techniques were used to study this waste processing. Two kinds of resins were used by the ion exchange method: the strong anionic and the stron cationic. Quantitative tellurium retention has not been attained by the ion exchange method using either resins. The technique of precipitation of radioactive tellurium as ammonium tellurate was also used, allowing us to obtain more than 99% of tellurium removal. The remaining radioactive wastes can be eliminated using the storage for decay criteria with further release to the sewers in the case of liquids and burial in the case of solids. (Author) [pt

  10. On the self-diffusion process in liquid metals and alloys by the radioactive tracer method

    International Nuclear Information System (INIS)

    Ganovici, L.

    1978-01-01

    A theoretical and experimental study of self-diffusion process in liquid metals and alloys is presented. There are only a few pure metals for which diffusion coefficients in a liquid state are known. The thesis aims at increasing the number of liquid metals for which diffusion coefficients are available, by determining these values for liquids: Cd, Tl, Sb and Te. The self-diffusion coefficients of Te in some tellurium based liquid alloys such as Tl 2 Te, PbTe and Bi 90 Te 10 were also determined. Self-diffusion coefficients have been measured using two radioactive tracer methods: a) the capillary-reservoir method; b) the semi-infinite capillary method. The self-diffusion coefficients were derived from the measured radioactive concentration profile, using the solutions of Fick's second law for appropriate initial and limit conditions. The temperature dependence study of self-diffusion coefficients in liquids Cd, Tl, Sb and Te, was used to check some theoretical models on the diffusion mechanism in metallic melts. The experimental diffusion data interpreted in terms of the Arrhenius type temperature dependence, was used to propose two simple empiric relations for determining self diffusion coefficients of group I liquid metals and for liquid semi-metals. It was established a marked decrease of self-diffusion coefficients of liquid Te close to the solidification temperature. The diffusivity of Te in liquid Tl 2 Te points to an important decrease close to the solidification temperature. A simplified model was proposed for the diffusion structural unit in this alloy and the hard sphere model for liquid metals was checked by comparing the theoretical and experimental self-diffusion coefficients. (author)

  11. Evaluation of transport properties of nanofiltration membranes exposed to radioactive liquid waste

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Elizabeth E.M.; Barbosa, Celina C.R.; Bastos, Edna T.R., E-mail: eemo@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeira, RJ (Brazil); Afonso, Julio C., E-mail: Julio@iq.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Inst. de Quimica. Dept. de Quimica Analitica

    2011-07-01

    The application of membrane separation processes (PSM) for treatment of radioactive waste requires the selection of a suitable membrane for the treatment of waste, as the membrane will be directly exposed to the radioactive liquid waste, and also exposed to ionizing radiation. The nanofiltration membrane is most suitable for treatment of radioactive waste, since it has high rejection of multivalent ions. Usually the membranes are made of polymers and depending on the composition of the waste, type and dose of radiation absorbed may be changes in the structure of the membrane, resulting in loss of its transport properties. We tested two commercial nanofiltration membranes: NF and SW Dow/Filmtec. The waste liquid used was obtained in the process of conversion of uranium hexafluoride gas to solid uranium dioxide, known as 'carbonated water'. The membranes were characterized as their transport properties (hydraulic permeability, permeate flux and salt rejection) before and after their immersion in the waste for 24 hours. The surface of the membranes was also evaluated by SEM and FTIR. It was observed that in both the porosity of the membrane selective layer was altered, but not the membrane surface charge, which is responsible for the selectivity of the membrane. The NF membranes and SW showed uranium ion rejection of 64% and 55% respectively. (author)

  12. 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

  13. Reduction of INTEC Analytical Radioactive Liquid Wastes

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  14. Radioactive liquid wastes discharged to ground in the 200 Areas during 1978

    International Nuclear Information System (INIS)

    Anderson, J.D.; Poremba, B.E.

    1979-01-01

    This document is issued quarterly for the purpose of summarizing the radioactive liquid wastes that have been discharged to the ground in the 200 Areas. In addition to data for 1978, cumulative data since plant startup are presented. Also, in this document is a listing of decayed activity to the various plant sites

  15. Radioactive liquid wastes discharged to ground in the 200 areas during the first three quarters of 1974

    International Nuclear Information System (INIS)

    Anderson, J.D.

    1974-01-01

    An overall summary of radioactive liquid wastes discharged to ground during the first three quarters of 1974 and since startup within the Production and Waste Management control zone is presented in tabular form. Estimates of the radioactivity discharged to individual ponds, cribs, and specific retention sites are given. (LK)

  16. Principal prerequisites and practice for using deep aquifers for disposal of liquid radioactive wastes

    International Nuclear Information System (INIS)

    Spitsyn, V.I.; Pimenov, M.K.; Balukova, V.D.; Leontichuk, A.S.; Kokorin, I.N.; Yudin, F.P.; Rakov, N.A.

    1977-01-01

    One of the most promising methods of safe disposal of liquid radioactive wastes in the USSR is the creation of storage places in deep aquifers in zones of stagnant regime or the slow exchange of underground water. The results of investigations and disposal practices testify to the safety and efficiency of such a method of final waste disposal which fulfils the main requirements for protecting the environment. Geological formations and stratum-collectors may be studied and selected to secure localization of liquid radioactive wastes injected into them for many tens and even hundreds of thousand years. The main requirements and criteria which must be met by geological structures and stratum-collectors to ensure safe disposal of wastes are formulated. Waste disposal is realized only after a thorough scientific appreciation of health and safety of present and future generations with regard to the regime of disposal and physico-chemical processes depending on the compatibility of the wastes with rocks and stratal waters as well as on the period of time of waste exposure up to the maximum permissible concentrations. Positive and negative factors of the method are analysed. Methods of preparing waste for disposal and chemical methods of restoring the response of the holes, ways of effective remote control of disposal and environment, etc., are briefly discussed. The results of 10-12 years experimental and industrial exploitation of storage places for liquid radioactive wastes of low- and medium-level activity are presented. The results of enlarged field tests on disposal of high-level activity liquid wastes are described. Preliminary prediction calculations are shown to be confirmed with sufficient accuracy by the data on exploitation. (author)

  17. Comparison of the accident process, radioactivity release and ground contamination between Chernobyl and Fukushima-1

    International Nuclear Information System (INIS)

    Imanaka, Tetsuji; Hayashi, Gohei; Endo, Satoru

    2015-01-01

    In this report, we have reviewed the basic features of the accident processes and radioactivity releases that occurred in the Chernobyl accident (1986) and in the Fukushima-1 accident (2011). The Chernobyl accident was a power-surge accident that was caused by a failure of control of a fission chain reaction, which instantaneously destroyed the reactor and building, whereas the Fukushima-1 accident was a loss-of-coolant accident in which the reactor cores of three units were melted by decay heat after losing the electricity supply. Although the quantity of radioactive noble gases released from Fukushima-1 exceeded the amount released from Chernobyl, the size of land area severely contaminated by 137 Cesium ( 137 Cs) was 10 times smaller around Fukushima-1 compared with around Chernobyl. The differences in the accident process are reflected in the composition of the discharged radioactivity as well as in the composition of the ground contamination. Volatile radionuclides (such as 132 Te- 132 I, 131 I, 134 Cs and 137 Cs) contributed to the gamma-ray exposure from the ground contamination around Fukishima-1, whereas a greater variety of radionuclides contributed significantly around Chernobyl. When radioactivity deposition occurred, the radiation exposure rate near Chernobyl is estimated to have been 770 μGy h −1 per initial 137 Cs deposition of 1000 kBq m −2 , whereas it was 100 μGy h −1 around Fukushima-1. Estimates of the cumulative exposure for 30 years are 970 and 570 mGy per initial deposition of 1000 kBq m −2 for Chernobyl and Fukusima-1, respectively. Of these exposures, 49 and 98% were contributed by radiocesiums ( 134 Cs + 137 Cs) around Chernobyl and Fukushima-1, respectively

  18. Disposal of liquid radioactive wastes through wells or shafts

    International Nuclear Information System (INIS)

    Perkins, B.L.

    1982-01-01

    This report describes disposal of liquids and, in some cases, suitable solids and/or entrapped gases, through: (1) well injection into deep permeable strata, bounded by impermeable layers; (2) grout injection into an impermeable host rock, forming fractures in which the waste solidifies; and (3) slurrying into excavated subsurface cavities. Radioactive materials are presently being disposed of worldwide using all three techniques. However, it would appear that if the techniques were verified as posing minimum hazards to the environment and suitable site-specific host rock were identified, these disposal techniques could be more widely used

  19. Risk of transporting plutonium dioxide and liquid plutonium nitrate by truck and rail

    International Nuclear Information System (INIS)

    Williams, L.D.; Hall, R.J.

    1976-01-01

    This paper presents an analysis method developed to assess the risk in the transportation of radioactive materials and demonstrates its application by summarizing the results of risk assessments of the shipment of plutonium dioxide powder and liquid plutonium nitrate by truck and rail in the United States. In the risk assessment, postulated release modes (series of events that could result in a release of radioactive material) are identified using the fault tree analysis method. Two categories of events that could contribute to a release during transportation are considered: accidents and substandard packaging conditions. The likelihood of basic events in these categories are determined from general accident experience, radioactive material shipping experience, engineering analysis, etc. Using the laws of combining probabilities, the likelihood of radioactive material release is then evaluated. Since accidents vary greatly in severity, the releases from these postulated events also vary in magnitude from the no release condition up through release of the entire package contents. The bases for the release quantities assigned to the postulated accidents sequences are discussed. The consequences of these postulated releases are evaluated based on the amount of plutonium estimated to be released to the environs, the probable weather conditions at the time of the accident, and population density downwind from the accident scene. The likelihood and consequence of postulated releases are coupled and expressed as a risk spectra. (author)

  20. Simulation of the distribution of radionuclides in the reservoir bed for deep-well injection disposal of acid liquid radioactive waste

    International Nuclear Information System (INIS)

    Noskov, M.D.; Istomin, A.D.; Kesler, A.G.; Zubkov, A.A.; Zakharova, E.V.; Egorov, G.F.

    2007-01-01

    A mathematical model was developed for describing the changes in the state of the reservoir bed for dee-well injection disposal of acid liquid radioactive waste. The model considers the multicomponent filtration of the solution in the heterogeneous bed, sorption-desorption of radionuclides, taking into account the dependence of the distribution coefficient on the temperature and pH, as well as radioactive decay, interaction of acids with minerals, radiation-chemical and thermochemical decomposition of the acids, and dynamics of the temperature field, taking into account the convective heat transfer, thermal conductivity, and radiogenic heat release. The results of the simulation of the migration of radionuclides were reported, as well as of the distribution of the acids and the dynamics of the temperature field in the vicinity of the injection well of the site for deep-well injection disposal of the waste from Siberian Chemical Combine. A man-caused barrier is formed in the vicinity of the injection well, hindering the spread of radionuclides in the reservoir bed [ru

  1. Analysis of release and transport of aerial radioactive materials in accident of evaporation to dryness caused by boiling of reprocessed high-level liquid waste

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Ishikawa, Jun; Abe, Hitoshi

    2015-01-01

    An accident of evaporation to dryness caused by boiling of high-level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, some amount of fission products (FPs) will be transferred to the vapor phase in the tank, and could be released to the environment. Therefore, the quantitative estimation of the transport and release behavior of FPs is one of the key issues in the assessment of the accident consequence. To resolve this issue, a systematic analysis method with computer codes has been developed on the basis of the phenomenological behavior in the accident of evaporation to dryness caused by boiling of HLLW. A simulation study demonstrated that the behavior of liquid waste temperature and the entrainment of mists were in good agreement with the experimental results during the early boiling stage, and that some issues to be resolved were pointed out for the estimation of the amount of transferred Ru to the vapor phase at the late boiling stage. (author)

  2. Treatment and disposal of radioactive waste, ch. 1

    International Nuclear Information System (INIS)

    Jonker, G.S.

    1976-01-01

    Low-level radioactive liquid wastes may be discharged into surface waters. However, discharge is rather severely restricted by internationally accepted regulations. Therefore, the liquid should be purified to an acceptable level before release into the environment. One of the techniques available for pretreatment of radioactive liquids is precipitation. Three methods have been worked out: the calcium chloride-sodium carbonate process, the barium sulphate and the calcium phosphate precipitation. The main attention was directed towards the removal of the potentially hazardous radiostrontium. Precipitation is not only suitable for the purification of radioactive waste waters but may also be used in case of an unguided contamination of the environment for the purification of surface waters before distribution as drinking water. In appendix I the chemical preparation of the calcium carbonate polymorphic forms calcite, aragonite and vaterite is described, because these play an important role in the first process. X-ray diffraction analysis of these compounds and the apparatus used are described in appendix II. Spectrophotometric analysis of phosphate and flame spectrophotometry for the analysis of calcium in presence of phosphate ions is described in appendix III. A literature survey is presented of the various precipitation methods which are already used in present day practice

  3. Fate of gaseous tritium and carbon-14 released from buried low-level radioactive waste

    International Nuclear Information System (INIS)

    Striegl, R.G.

    1988-01-01

    Microbial decomposition, chemical degradation, and volatilization of buried low-level radioactive waste results in the release of gases containing tritium ( 3 H) and carbon-14 ( 14 C) to the surrounding environment. Water vapor, carbon dioxide, and methane that contain 3 H or 14 C are primary products of microbial decomposition of the waste. Depending on the composition of the waste source, chemical degradation and volatilization of waste also may result in the production of a variety of radioactive gases and organic vapors. Movement of the gases in materials that surround waste trenches is affected by physical, geochemical, and biological mechanisms including sorption, gas-water-mineral reactions, isotopic dilution, microbial consumption, and bioaccumulation. These mechanisms either may transfer 3 H and 14 C to solids and infiltrating water or may result in the accumulation of the radionuclides in plant or animal tissue. Gaseous 3 H or 14 C that is not transferred to other forms is ultimately released to the atmosphere

  4. Electric-field triggered controlled release of bioactive volatiles from imine-based liquid crystalline phases.

    Science.gov (United States)

    Herrmann, Andreas; Giuseppone, Nicolas; Lehn, Jean-Marie

    2009-01-01

    Application of an electric field to liquid crystalline film forming imines with negative dielectric anisotropy, such as N-(4-methoxybenzylidene)-4-butylaniline (MBBA, 1), results in the expulsion of compounds that do not participate in the formation of the liquid crystalline phase. Furthermore, amines and aromatic aldehydes undergo component exchange with the imine by generating constitutional dynamic libraries. The strength of the electric field and the duration of its application to the liquid crystalline film influence the release rate of the expelled compounds and, at the same time, modulate the equilibration of the dynamic libraries. The controlled release of volatile organic molecules with different chemical functionalities from the film was quantified by dynamic headspace analysis. In all cases, higher headspace concentrations were detected in the presence of an electric field. These results point to the possibility of using imine-based liquid crystalline films to build devices for the controlled release of a broad variety of bioactive volatiles as a direct response to an external electric signal.

  5. Accumulation of radioactive cesium released from Fukushima Daiichi Nuclear Power Plant in terrestrial cyanobacteria Nostoc commune.

    Science.gov (United States)

    Sasaki, Hideaki; Shirato, Susumu; Tahara, Tomoya; Sato, Kenji; Takenaka, Hiroyuki

    2013-01-01

    The Fukushima Daiichi Nuclear Power Plant accident released large amounts of radioactive substances into the environment and contaminated the soil of Tohoku and Kanto districts in Japan. Removal of radioactive material from the environment is an urgent problem, and soil purification using plants is being considered. In this study, we investigated the ability of 12 seed plant species and a cyanobacterium to accumulate radioactive material. The plants did not accumulate radioactive material at high levels, but high accumulation was observed in the terrestrial cyanobacterium Nostoc commune. In Nihonmatsu City, Fukushima Prefecture, N. commune accumulated 415,000 Bq/kg dry weight (134)Cs and 607,000 Bq kg(-1) dry weight (137)Cs. The concentration of cesium in N. commune tended to be high in areas where soil radioactivity was high. A cultivation experiment confirmed that N. commune absorbed radioactive cesium from polluted soil. These data demonstrated that radiological absorption using N. commune might be suitable for decontaminating polluted soil.

  6. Release of the radioactive patient following radiation therapy

    International Nuclear Information System (INIS)

    Powers, J.; Cancer Board, A.

    2004-01-01

    Patients walk out of medical facilities containing as much as a complete therapeutic dose of radiation on a daily basis. This presents a significant challenge to the radiation protection community, as most patients have no prior education related to radiation and may not have the aptitude to assimilate such knowledge. In the case of targeted radiation therapy in which radionuclides are used to selectively target the cancer, patients are typically released only after adequate elimination and decay of the radionuclide administered. Established modalities of targeted radiotherapy include the use of iodine for thyroid cancer, strontium for bone pain, phosphorous for haematological diseases, 131I-mIBG for neuroblastoma, and most recently Y-90 labelled monoclonal antibodies for lymphoma. In the case of permanent implants, implants of encapsulated radioactive sources are left permanently in the tissues, thus patients are released containing their complete therapeutic dose. Isotopes used in permanent implants include I-125, Pd-103 and Au-198. Radiation safety considerations for both cases, the release of a patient who has received targeted radiotherapy, and the release of a patient who has received a permanent implant, will be discussed. A summary of applicable regulations will serve as a starting point for each of the following considerations; i) Security and source control ii) Instructions to patient and family members iii) Risk to the public As the incidence of cancer increases, and the popularity of targeted radiotherapy and permanent seed implants grows, the event of having an untrained person in possession of a therapeutic dose of radiation becomes more and more common. It is essential to stop and examine the risk of this practice, whether current strategies to reduce the risk to an acceptable level are indeed effective, and whether control over these sources is even feasible. (Author)

  7. Release of the radioactive patient following radiation therapy

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J.; Cancer Board, A.

    2004-07-01

    Patients walk out of medical facilities containing as much as a complete therapeutic dose of radiation on a daily basis. This presents a significant challenge to the radiation protection community, as most patients have no prior education related to radiation and may not have the aptitude to assimilate such knowledge. In the case of targeted radiation therapy in which radionuclides are used to selectively target the cancer, patients are typically released only after adequate elimination and decay of the radionuclide administered. Established modalities of targeted radiotherapy include the use of iodine for thyroid cancer, strontium for bone pain, phosphorous for haematological diseases, 131I-mIBG for neuroblastoma, and most recently Y-90 labelled monoclonal antibodies for lymphoma. In the case of permanent implants, implants of encapsulated radioactive sources are left permanently in the tissues, thus patients are released containing their complete therapeutic dose. Isotopes used in permanent implants include I-125, Pd-103 and Au-198. Radiation safety considerations for both cases, the release of a patient who has received targeted radiotherapy, and the release of a patient who has received a permanent implant, will be discussed. A summary of applicable regulations will serve as a starting point for each of the following considerations; i) Security and source control ii) Instructions to patient and family members iii) Risk to the public As the incidence of cancer increases, and the popularity of targeted radiotherapy and permanent seed implants grows, the event of having an untrained person in possession of a therapeutic dose of radiation becomes more and more common. It is essential to stop and examine the risk of this practice, whether current strategies to reduce the risk to an acceptable level are indeed effective, and whether control over these sources is even feasible. (Author)

  8. Development of DOE complexwide authorized release protocols for radioactive scrap metals

    International Nuclear Information System (INIS)

    Chen, S. Y.

    1998-01-01

    Within the next few decades, several hundred thousand tons of metal are expected to be removed from nuclear facilities across the U.S. Department of Energy (DOE) complex as a result of decontamination and decommissioning (D and D) activities. These materials, together with large quantities of tools, equipment, and other items that are commonly recovered from site cleanup or D and D activities, constitute non-real properties that warrant consideration for reuse or recycle, as permitted and practiced under the current DOE policy. The provisions for supporting this policy are contained in the Draft Handbook for Controlling Release for Reuse or Recycle of Property Containing Residual Radioactive Material published by DOE in 1997 and distributed to DOE field offices for interim use and implementation. The authorized release of such property is intended to permit its beneficial use across the entire DOE complex. The objective of this study is to develop readily usable computer-based release protocols to facilitate implementation of the Handbook in evaluating the scrap metals for reuse and recycle. The protocols provide DOE with an effective oversight tool for managing release activities

  9. Decision no 2010-DC-0188 by the Nuclear Safety Authority on the 7. of July 2010 specifying to Electricite de France-Societe Anonyme (EDF-SA) the limits of releases in the environment of liquid and gaseous effluents for the operation of Flamanville 1 (IBN n 108), Flamanville 2 (INB n 109) and Flamanville 3 (INB n 167) reactors

    International Nuclear Information System (INIS)

    2010-01-01

    This document contains references to the different legal and official documents (codes, orders, minister's opinion, public surveys, administrative authorizations, local community opinion) at the root of this specification of limits related to releases in the environment of liquid and gaseous effluents by the three Flamanville nuclear reactors operated by Electricite de France-Societe Anonyme (EDF-SA). Tables present the limits for different gaseous and liquid radioactive emissions (carbon 14, hydrogen 3, iodine, radioactive rare earths, and other beta and gamma emitters) from these different installations, but also the maximum admitted concentrations for gaseous chemical effluents (boric acid, hydrazine hydrate, morpholine, phosphates, nitrates, various metals, and so on). The limits for thermal releases are also specified

  10. Probabilistic siting analysis of nuclear power plants emphasizing atmospheric dispersion of radioactive releases and radiation-induced health effects

    International Nuclear Information System (INIS)

    Savolainen, Ilkka

    1980-01-01

    A presentation is made of probabilistic evaluation schemes for nuclear power plant siting. Effects on health attributable to ionizing radiation are reviewed, for the purpose of assessment of the numbers of the most important health effect cases in light-water reactor accidents. The atmospheric dispersion of radioactive releases from nuclear power plants is discussed, and there is presented an environmental consequence assessment model in which the radioactive releases and atmospheric dispersion of the releases are treated by the application of probabilistic methods. In the model, the environmental effects arising from exposure to radiation are expressed as cumulative probability distributions and expectation values. The probabilistic environmental consequence assessment model has been applied to nuclear power plant site evaluation, including risk-benefit and cost-benefit analyses, and the comparison of various alternative sites. (author)

  11. Review of calculational models and computer codes for environmental dose assessment of radioactive releases

    International Nuclear Information System (INIS)

    Strenge, D.L.; Watson, E.C.; Droppo, J.G.

    1976-06-01

    The development of technological bases for siting nuclear fuel cycle facilities requires calculational models and computer codes for the evaluation of risks and the assessment of environmental impact of radioactive effluents. A literature search and review of available computer programs revealed that no one program was capable of performing all of the great variety of calculations (i.e., external dose, internal dose, population dose, chronic release, accidental release, etc.). Available literature on existing computer programs has been reviewed and a description of each program reviewed is given

  12. Review of calculational models and computer codes for environmental dose assessment of radioactive releases

    Energy Technology Data Exchange (ETDEWEB)

    Strenge, D.L.; Watson, E.C.; Droppo, J.G.

    1976-06-01

    The development of technological bases for siting nuclear fuel cycle facilities requires calculational models and computer codes for the evaluation of risks and the assessment of environmental impact of radioactive effluents. A literature search and review of available computer programs revealed that no one program was capable of performing all of the great variety of calculations (i.e., external dose, internal dose, population dose, chronic release, accidental release, etc.). Available literature on existing computer programs has been reviewed and a description of each program reviewed is given.

  13. Bioremoval of Am-241 and Cs-137 from liquid radioactive wasters by bacterial consortiums

    International Nuclear Information System (INIS)

    Ferreira, Rafael Vicente de Padua; Lima, Josenilson B. de; Gomes, Mirella C.; Borba, Tania R.; Bellini, Maria Helena; Marumo, Julio Takehiro; Sakata, Solange Kazumi

    2011-01-01

    This paper evaluates the capacity of two bacterial consortiums of impacted areas in removing the Am-241 and Cs-137 from liquid radioactive wastes.The experiments indicated that the two study consortiums were able to remove 100% of the Cs-137 and Am-241 presents in the waste from 4 days of contact. These results suggest that the bio removal with the selected consortiums, can be a viable technique for the treatment of radioactive wastes containing Am-241 and Cs-137

  14. Chemical Studies on the treatment and Conditioning of Radioactive Liquid Waste Using Combined Processes

    International Nuclear Information System (INIS)

    El-Masry, E.H.

    2004-01-01

    Natural inorganic exchanges were used to remove radioactive isotopes cesium, Cobalt and europium using coagulant zinc sulfate as coagulant from low level liquid radioactive waste. the highest percent of removal was obtained in the order asswanlly (85.5%), bentonite (82.2%) and sandstone (65.4%) for the removal of cesium . the same order of removal percent was detected for the removal of cobalt (92.5,91.2,90.6%) and europium (90.6,90.8,90.2%) for asswanlly, bentonite and sandstone respectively

  15. Malicious release of radioactive materials in urban area. Exposure of the public and emergency staff, protective measures

    International Nuclear Information System (INIS)

    Koch, Wolfgang; Lange, Florentin

    2016-01-01

    The preparedness for hypothetical radiological scenarios is part of the tasks for governmental authorities, safety and emergency organizations and the staff in case of the incident. The EURATOM guideline for radiation protection has to be implemented into national laws. According to the guidelines it is required that emergency planning has to be prepared for hypothetical radiological scenarios including terroristic or other maliciously motivated attacks using radioactive materials. The study includes assumptions on the released respirable radioactivity, restriction of the hazardous area, wind induced re-suspension of radioactive dusts and inhalation exposure, and mitigation measures.

  16. A flowrate measurement method by counting of radioactive particles suspended in a liquid

    International Nuclear Information System (INIS)

    Daniel, G.

    1983-04-01

    By external counting of fine #betta# emitting radioactive particles suspended in a liquid, the flowrate in a system of pipes can be measured. The study comprises three phases: 1. - The hydraulic validity of the method is demonstrated in laminar as well as in turbulent flow under certain conditions of particles size and density and of liquid viscosity. 2. - Radioactive labelling of microspheres of serumalbumin or ion exchange resins with indium 113m delivered by a generator Tin 113 → Indium 113m. 3. - Counting with a scintillation detector: a method of threshold overstepping is experimented with a mechanical or electronic simulator; the statistical study of particle superposition under the detector enables a correction for the resulting counting losses to be proposed. The method provides absolute measurements, but is particularly suitable to measure relative flowrates in a hydraulic network. It can be continuous and does not perturb the flow and the network. The accuracy of the method is analysed in details [fr

  17. Development of a process to reduce the uranium concentration of liquid radioactive waste

    International Nuclear Information System (INIS)

    Fuentealba Toro, Edgardo David

    2015-01-01

    The purpose of radioactive waste management is to prevent the discharge of waste into the biosphere, a function carried out in Chile by the Chilean Nuclear Energy Commission (CCHEN), which stores around 500 [L] of these organic and inorganic waste in cans coming from research of Universities and CCHEN' laboratories. Within the inorganic liquid waste are concentrations of Uranyl salts with sulfates, chlorides and phosphates. The purpose of this work is to develop at laboratory level a process to concentrate and precipitate uranium salts (Sulfate and Uranyl Chloride) present in radioactive liquid effluents, because in the case of these very long period wastes in liquid state, the most widely used processes are aimed at concentrating or extracting radioactive compounds through separation processes, for their conditioning and final storage under conditions whose radiological risk is minimized. The selected process is liquid-liquid extraction, being evaluated solvents such as benzene and kerosene with the following extractants: tri-n-octylphosphine oxide (TOPO), di-2-ethylhexyl phosphoric acid (DEHPA) and Cyanex© 923. To determine the extraction conditions, which allow to reduce the concentration of uranium to values lower than 10 ppm, the extractant concentration was modified from 0.05 to 0.41 [M] with solvent volume / residue (VO/VA) ratios of 0.2 to 0.5, at an initial concentration of 8,446 [gU/L] and subsequent precipitation of uranium extracted by a reaction with ammonium carbonate. From these experimental tests the maximum extraction conditions were determined. To the generated effluents, a second stage of extraction was necessary in order to reduce its concentration below 10 [mg / L]. The experimental tests allowed to reduce the concentration under 2.5 [mgU/L], equivalent to 99.97% extraction efficiency. The tests with Cyanex© 923 in replacement of the TOPO, allowed to obtain similar results and even better in some cases, due to the fact that final

  18. Water - The radiological health of rivers: releases are very much controlled downstream power plants. What do hospital releases represent? The Seine reserves a surprise

    International Nuclear Information System (INIS)

    Anon.

    2015-01-01

    After a brief presentation of the role of the IRSN in the control of the radioactivity present in waters and in the control and follow-up of all sources of radioactivity, a first article briefly present the hydro-collector network, indicates that some point samplings of sediment and aquatic species are performed, that a national network of beacons for a continuous radioactivity measurement is installed in the main French rivers, downstream nuclear installations, and that advanced measurement techniques are used to detect very small level of tritium. Maps giving a brief indication of the radiological condition of the Loire and Rhone are provided. A second article addresses the control of releases downstream power plants, and evokes the legal context and the associated objectives and produced documents. The third article discusses the risk associated with hospital wastes and releases (liquid and solid effluents), how radioactivity is controlled between the hospital and tap water distribution. The last article reports and comments the results obtained by an analysis of historical pollutions trapped in the sediments of the Seine: 40 year-old traces of plutonium have been discovered, due to an accidental release from a CEA installation in Fontenay-aux-Roses, with no detrimental impact on population or on sewer workers

  19. Development of an immobilisation technique by cementation for non-radioactive simulated liquid waste, from Mo-99 production process

    International Nuclear Information System (INIS)

    Arva, E A; Marabini, S G; Varani, J L

    2012-01-01

    The Argentine Atomic Energy Commission (CNEA) is the responsible for developing a management nuclear waste disposal programme. This programme contemplates the strictly environmental safe and efficient management of the radioactive waste from different sources. Since 1985, CNEA has been producing commercially Mo-99 for medical use. In this process two types of liquid waste are produced. One of them has high alkaline (NaOH 3,5M) and aluminate contents. Since Mo-99 production started, such liquid waste was stored in specially designed containers during production, and after a decay period in smaller containers in interim storage conditions. As this waste is still a liquid, development of an immobilisation technique is required. Immobilisation of radioactive liquid waste by cementation is a frequently used technique, and will be studied in the present work using Mo-99 non-radioactive simulated liquid waste. In this second stage, a full scale (200 liters drum) cementation test using simulated non radioactive waste was carried out. Such test included: using the BEBA 201 mixing machine - the same that will be used with real waste in the future for 'tuning up' the process, construction of a specially designed temperature sensor for measuring the maximum temperature value (five different positions, four inside the drum and one outside) and the time elapsed after all components mixing. Finally, standard specimens (IRAM 1622) were made for mechanical resistance tests after cement setting at 28 days. The results show values of temperature not above 40 o C with the maximum at 12 hours before component mixing and compression strength of 14 MPa. Such values are compatible for a waste immobilisation process by cementation (author)

  20. Summarization of radioactive effluent monitoring in Daya Bay Nuclear Power Station 1994-2002

    International Nuclear Information System (INIS)

    Ouyang Junjie; Chen Yue

    2004-01-01

    This paper introduces the radioactive effluent monitoring systems, measurement and quality control methods used in Daya Bay Nuclear Power Station from its commercial operation in 1994. The main work and experiences for the management of effluent release are discussed and analyzed. The radwaste data appear declining trend and are far blow the annual limit approved by the national Environmental Protection Bureau since 1995. The normalized release (unit GBq/GWa) of 9 years is as follows: liquid nuclides (except tritium) 11.1, liquid tritium 1.91 x 10 4 , noble gas 2.03 x 10 4 , halogen 0.13, aerosol 7.57 x 10 -3 . For 110m Ag, the average release from 1998 to 2002 has been reduced to 1/7 of the quantity in 1997

  1. Population dose commitments due to radioactive releases from nuclear power plant sites in 1975

    International Nuclear Information System (INIS)

    Baker, D.A.; Soldat, J.K.; Watson, E.C.

    1977-10-01

    Population radiation dose commitments were estimated from reported radionuclide releases from commercial power reactors operating during 1975. Fifty-year dose commitments from one year exposure were calculated from both liquid and atmospheric releases for four population groups (infant, child, teenager and adult) residing between 2 and 80 km from each site. Results are given in the form of tables giving the dose commitments for both liquid and airborne pathways for each age group and organ. Also included for each site is a histogram showing the fraction of the total population within the 2 to 80-km region around each site receiving various average dose commitments from the airborne pathways. The total dose commitment from both liquid and airborne pathways ranged from a high of 750 person-rem to a low of 0.008 person-rem with an arithmetic mean of 34 person-rem

  2. Management of radioactive low level liquid, gaseous, and solid wastes in the 200 areas

    International Nuclear Information System (INIS)

    White, A.T.

    1976-01-01

    The practices which are currently used for handling radioactive waste are outlined. These include burial of solid waste, scrubbing of off gas streams, and routing liquid effluents (mostly cooling water) to open ponds where the water percolates to the water table

  3. Process for producing zeolite adsorbent and process for treating radioactive liquid waste with the zeolite adsorbent

    International Nuclear Information System (INIS)

    Motojima, K.; Kawamura, F.

    1984-01-01

    Zeolite is contacted with an aqueous solution containing at least one of copper, nickel, cobalt, manganese and zinc salts, preferably copper and nickel salts, particularly preferably copper salt, in such a form as sulfate, nitrate, or chloride, thereby adsorbing the metal on the zeolite in its pores by ion exchange, then the zeolite is treated with a water-soluble ferrocyanide compound, for example, potassium ferrocyanide, thereby forming metal ferrocyanide on the zeolite in its pores. Then, the zeolite is subjected to ageing treatment, thereby producing a zeolite adsorbent impregnated with metal ferrocyanide in the pores of zeolite. The adsorbent can selectively recover cesium with a high percent cesium removal from a radioactive liquid waste containing at least radioactive cesium, for example, a radioactive liquid waste containing cesium and such coexisting ions as sodium, magnesium, calcium and carbonate ions at the same time at a high concentration. The zeolite adsorbent has a stable adsorbability for a prolonged time

  4. Management of radioactive liquid and solid wastes at the Research Reactor Institute, Kyoto University, (3)

    International Nuclear Information System (INIS)

    Tsutsui, Tenson; Shimoura, K.; Koyama, A.

    1977-11-01

    In this report, the management of radioactive liquid and solid wastes at the Research Reactor Institute, Kyoto University during past 6 years, from April in 1971 to March in 1977 are reviewed. (auth.)

  5. Liquid Effluent Monitoring Information System test plans release 1.2

    International Nuclear Information System (INIS)

    Adams, R.T.

    1994-01-01

    The Liquid Effluent Monitoring Information System (LEMIS) is being developed as the organized information repository facility in support of the liquid effluent monitoring requirements of the Tri-Party Agreement. It is necessary to provide an automated repository into which the results from liquid effluent sampling will be placed. This repository must provide for effective retention, review, and retrieval of selected sample data by authorized persons and organizations. This System Architecture document is the aggregation of the DMR P+ methodology project management deliverables. Together they represent a description of the project and its plan through four Releases, corresponding to the definition and prioritization of requirements defined by the user

  6. Application of ion exchange in liquid radioactive waste management of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Puskar; Chopra, S K; Sharma, P D [Nuclear Power Corporation, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    The operation of nuclear power plants would necessarily result in generation of gaseous, liquid and solid radioactive wastes. The wastes are treated/conditioned to ensure that the permissible discharge limits laid down by Atomic Energy Regulatory Board of India are complied with. The wastes are segregated on activity levels, types of radioisotopes present and chemical nature of liquid streams. The basic philosophy of various treatment techniques is to concentrate and contain as much activity as possible. It is of utmost importance that the wastes are effectively treated by proven methods/processes. The radiochemical nature of waste generated is one of the parameters to select a treatment/conditioning method. The paper presents an outline of various processes adopted for treatment of liquid waste and ion exchange processes, their application in liquid waste management in detail. Projected quantities of liquid wastes for the current designs are included. (author). 2 tabs.

  7. A cost/benefit analysis of methods for controlling the release of radioactive materials in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Blanco, R.E.; Dahlman, R.C.; Davis, W. Jr.; Finney, B.C.; Groenier, W.S.; Hill, G.S.; Kibbey, A.H.; Kitts, F.G.; Lindauer, R.B.; Moore, R.E.; Pechin, W.H.; Roddy, J.W.; Ryon, A.D.; Seagren, R.D.; Sears, M.B.; Witherspoon, J.P.

    1977-01-01

    Cost/benefit surveys were made to determine the cost (in dollars) and effectiveness of radwaste treatment systems for decreasing the release of radioactive materials from model fuel cycle facilities, and to determine the benefits in terms of reduction in radiological dose commitment to individuals and populations in the surrounding areas. The studies include milling of uranium ores, conversion of virgin uranium and recycle uranium to UF 6 , fabrication of light-water reactor (LWR) fuels containing enriched uranium or enriched uranium and plutonium, fabrication of high-temperature gas-cooled reactor (HTGR) fuels containing 233 U and thorium, and reprocessing of LWR and HTGR fuels. Conceptual flowsheets were prepared for each model facility illustrating the treatment methods for gaseous and liquid effluents. The ''base'' case represents the lowest treatment cost, current treatment technology, and highest radiological dose. In succeeding cases, increasingly efficient radwaste treatment equipment is added to the ''base'' plant to reduce the amounts of radioactive materials released. The technology ranges from that currently available to that which may be developed over the next 30 years. The status of development for these technologies is discussed. The dose estimates are for maximum individual total body and organ doses at the plant boundary and for population total-body and organ doses out to 89 km. Comparisons of the doses vs annual costs in dollars are presented. In summary, they indicate that (1) the annual doses can be reduced to very low fractions of the natural background dose by the successful development and application of the radwaste treatment methods; and (2) excluding mills, the capital costs for the treatment methods vary from 0.02 to 8% of the capital cost of the base plants and the total annual operating costs (fixed charges plus operating costs) vary from 0.009 to 7.0% of the capital costs for the plant

  8. Study about the integrated treatment of chemical and radioactive effluents, introducing the zero release concept

    International Nuclear Information System (INIS)

    Mierzwa, Jose Carlos

    1996-01-01

    An Integrated System to the treatment of Chemical and Radioactive Effluents to the Centro Experimental Aramar is proposed and evaluated, introducing the Effluent Zero Release concept, where factors related to the environmental regulation in vigor in the country, as well as the availability of hydrological resources in the place where CEA have been implanted, are considered. Through a literature analysis of the main effluents treatment techniques available nowadays and after a case of study selection, take into account two industrial installations that will be implanted at CEA, it was defined an arrangement to compose the Integrated System to the Treatment of Chemicals and Radioactive Effluents, focusing the Zero Release concept consolidation. A defined arrangement uses a combination among three treatment processes, it means chemical precipitation, reverse osmosis and evaporation, that were experimentally evaluated. The proposed arrangement was evaluated using synthetic effluents, that were prepared based on data from literature and conception documents of the installation considered in this work. Three kinds of effluents were simulated, one arising from a nuclear reactor laundry, one arising from the water refrigeration system and demineralized water production to the nuclear reactor and the other one arising from a nuclear material production laboratory. Each effluent were individually submitted to the selected treatment processes, to get the best operational conditions for each treatment process. The results got during the laboratory assays show that the proposed Integrated System to the Treatment of Chemicals and Radioactive Effluents is feasible, consolidating the Effluent Zero Release concept, which is the proposition of this work. (author)

  9. Method of electrolytic processing for radioactive liquid waste

    International Nuclear Information System (INIS)

    Otsuka, Katsuyuki; Takahashi, Yoshiharu; Tamai, Hideaki.

    1989-01-01

    Radioactive liquid wastes containing sodium compounds are electrolized using mercury as a cathode. As a result, they are separated into sodium-containing metal amalgam and residues. Metals containing sodium are separated from amalgam, purified and re-utilized, while mercury is recycled to the electrolysis vessel. The foregoing method can provide advantageous effect such as: (1) volume of the wastes to be processed can be reduced, (2) since processing can be carried out at a relatively low temperature, low boiling elements can be handled with no evaporization, (3) useful elements can be recovered and (4) other method than glass solidification can easily be employed remarkable volume-reduction of solidification products can be expected. (K.M.)

  10. Oxidizing purification of liquid radioactive waste from organic substances and radionuclides by K permanganate

    International Nuclear Information System (INIS)

    Rudenko, L.I.; Dzhuzha, O.V.; Khan, V.E.

    2007-01-01

    The basic opportunity of the oxidizing purification of liquid radioactive waste (LRW) with the use of a water solution of potassium permanganate for the preliminary preparation of LRW at a stage prior to the evaporating devices of the Chernobyl NPP is shown

  11. Boron Removal in Radioactive Liquid Waste by Forward Osmosis Membrane

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dooseong; Choi, Hei Min; Lee, Kune Woo; Moon Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    These wastes contain about 0.3-0.8 wt% boric acid and have been concentrated through an evaporation treatment. Boric acid tends to crystallize owing to its solubility, and to plug the evaporator. The volume reduction obtained through evaporation is limited by the amount of boric acid in the waste. As an emerging technology, forward osmosis (FO) has attracted growing interest in wastewater treatment and desalination. FO is a membrane process in which water flows across a semi-permeable membrane from a feed solution of lower osmotic pressure to a draw solution of higher osmotic pressure. However, very few studies on the removal of boron by FO have been performed. The objective of this study is to evaluate the possibility of boron separation in radioactive liquid waste by FO. In this study, the performance of FO was investigated to separate boron in the simulated liquid waste under the factors such as pH, osmotic pressure, ionic strength of the solution, and membrane characteristic. The boron separation in radioactive borate liquid waste was investigated with an FO membrane. When the feed solution containing boron is treated by the FO membrane, the boron permeation depends on the type of membrane, membrane orientation, pH of the feed solution, salt and boron concentration in the feed solution, and osmotic pressure of the draw solution. The boron flux begins to decline from pH 7, and increases with an increase in the osmotic driving force. The boron flux of the CTA-ES and ALFD membrane orientation is higher than those of the CTA-NW and ALFF orientation, respectively. The boron permeation rate is constant regardless of the osmotic pressure and membrane orientation. The boron flux decreases slightly with the salt concentration, but it is not heavily influenced at a low salt concentration.

  12. Melton Valley liquid low-level radioactive waste storage tanks evaluation

    International Nuclear Information System (INIS)

    1995-06-01

    The Melton Valley Liquid Low-Level Radioactive Waste Storage Tanks (MVSTs) store the evaporator concentrates from the Liquid Low-Level Radioactive Waste (LLLW) System at the Oak Ridge National Laboratory (ORNL). The eight stainless steel tanks contain approximately 375,000 gallons of liquid and sludge waste. These are some of the newer, better-designed tanks in the LLLW System. They have been evaluated and found by the US Environmental Protection Agency (EPA) and the Tennessee Department of Environment and Conservation to comply with all Federal Facility Agreement requirements for double containment. The operations and maintenance aspects of the tanks were also reviewed by the Defense Nuclear Facilities Safety Board (DNFSB) in September 1994. This document also contains an assessment of the risk to the public and ORNL workers from a leak in one of the MVSTs. Two primary scenarios were investigated: (1) exposure of the public to radiation from drinking Clinch River water contaminated by leaked LLLW, and (2) exposure of on-site workers to radiation by inhaling air contaminated by leaked LLLW. The estimated frequency of a leak from one of the MVSTs is about 8 x 10 -4 events per year, or about once in 1200 years (with a 95% confidence level). If a leak were to occur, the dose to a worker from inhalation would be about 2.3 x 10 -1 mrem (with a 95% confidence level). The dose to a member of the public through the drinking water pathway is estimated to be about 7 x 10 -1 mrem (with a 95% confidence level). By comparison with EPA Safe Drinking Water regulations, the allowable lifetime radiation dose is about 300 mrem. Thus, a postulated LLLW leak from the MVSTs would not add appreciably to an individual's lifetime radiation dose

  13. Boron Removal in Radioactive Liquid Waste by Forward Osmosis Membrane

    International Nuclear Information System (INIS)

    Hwang, Dooseong; Choi, Hei Min; Lee, Kune Woo; Moon Jeikwon

    2014-01-01

    These wastes contain about 0.3-0.8 wt% boric acid and have been concentrated through an evaporation treatment. Boric acid tends to crystallize owing to its solubility, and to plug the evaporator. The volume reduction obtained through evaporation is limited by the amount of boric acid in the waste. As an emerging technology, forward osmosis (FO) has attracted growing interest in wastewater treatment and desalination. FO is a membrane process in which water flows across a semi-permeable membrane from a feed solution of lower osmotic pressure to a draw solution of higher osmotic pressure. However, very few studies on the removal of boron by FO have been performed. The objective of this study is to evaluate the possibility of boron separation in radioactive liquid waste by FO. In this study, the performance of FO was investigated to separate boron in the simulated liquid waste under the factors such as pH, osmotic pressure, ionic strength of the solution, and membrane characteristic. The boron separation in radioactive borate liquid waste was investigated with an FO membrane. When the feed solution containing boron is treated by the FO membrane, the boron permeation depends on the type of membrane, membrane orientation, pH of the feed solution, salt and boron concentration in the feed solution, and osmotic pressure of the draw solution. The boron flux begins to decline from pH 7, and increases with an increase in the osmotic driving force. The boron flux of the CTA-ES and ALFD membrane orientation is higher than those of the CTA-NW and ALFF orientation, respectively. The boron permeation rate is constant regardless of the osmotic pressure and membrane orientation. The boron flux decreases slightly with the salt concentration, but it is not heavily influenced at a low salt concentration

  14. Occupational radiation protection experience in radioactive waste management at Bhabha Atomic Research Centre, Trombay, Mumbai, India

    International Nuclear Information System (INIS)

    Ramchandran, V.; Jauhri, G.S.

    2000-01-01

    Waste Management Facilities, Trombay (WMFT) comprises Radioactive Solid waste Management Site (RSMS), an Effluent Treatment Plant (ETP), and a Decontamination Centre (DC). Radioactive wastes from the plants and laboratories in Mumbai are handled here. The wastes are categorized and classified as per International Atomic Energy Agency (IAEA) and Atomic Energy Regulatory Board (AERB) guidelines. RSMS is a near surface disposal facility, where assorted beta gamma solid waste is disposed off in appropriate disposal facilities. ETP is a centralized low level liquid waste treatment facility, where liquid effluent is chemically treated to remove the radionuclides present in it, monitored for radioactivity, and discharged into the Mumbai Harbour Bay. In DC, plant and laboratory used clothings and personnel protective wears are decontaminated, monitored and sent for reuse. A comprehensive radiation monitoring programme is in place in these facilities from the beginning of radioactive waste management operations at BARC. The per capita radiation dose of the occupational workers and individual maximum dose has been low. Radioactivity release through liquid effluent from ETP has been kept well below Authorized Limits (AL). There has been no safety related unusual occurrences during the facility operation, that had any significant radiological impact. (author)

  15. Application of biosorbents in treatment of the radioactive liquid waste

    International Nuclear Information System (INIS)

    Ferreira, Rafael Vicente de Padua

    2014-01-01

    Radioactive liquid waste containing organic compounds need special attention, because the treatment processes available are expensive and difficult to manage. The biosorption is a potential treatment technique that has been studied in simulated wastes. The biosorption term is used to describe the removal of metals, non-metals and/or radionuclides by a material from a biological source, regardless of its metabolic activity. Among the potential biomasses, agricultural residues have very attractive features, as they allow for the removal of radionuclides present in the waste using a low cost biosorbent. The aim of this study was to evaluate the potential use of different biomass originating from agricultural products (coconut fiber, coffee husk and rice husk) in the treatment of real radioactive liquid organic waste. Experiments with these biomass were made including 1) Preparation, activation and characterization of biomasses; 2) Conducting biosorption assays; and 3) Evaluation of the product of immobilization of biomasses in cement. The biomasses were tested in raw and activated forms. The activation was carried out with diluted HNO 3 and NaOH solutions. Biosorption assays were performed in polyethylene bottles, in which were added 10 mL of radioactive waste or waste dilutions in deionized water with the same pH and 2% of the biomass (w/v). At the end of the experiment, the biomass was separated by filtration and the remaining concentration of radioisotopes in the filtrate was determined by ICP-OES and gamma spectrometry. The studied waste contains natural uranium, americium-241 and cesium-137. The adopted contact times were 30 min, 1, 2 and 4 hours and the concentrations tested ranged between 10% and 100%. The results were evaluated by maximum experimental sorption capacity and isotherm and kinetics ternary models. The highest sorption capacity was observed with raw coffee husk, with approximate values of 2 mg/g of U (total), 40 x 10 -6 mg/g of Am-241 and 50 x10 -9

  16. Intercomparison of numerical simulations on oceanic dispersion of the radioactive cesium released because of the Fukushima disaster

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, H.; Kobayshi, T.; Furuno, A. [Japan Atomic Energy Agency, Ibaraki (Japan); Usui, N.; Kamachi, M. [Japan Meteorological Agency, Meteorological Research Inst., Ibaraki (Japan); Nishikawa, S.; Ishikawa, Y. [Japan Agency for Marine-Earth Science and Tech., Kanagawa (Japan)

    2014-07-01

    We conducted numerical simulations on oceanic dispersion of the radioactive cesium released because of the Fukushima disaster in the North Pacific. Two independent oceanic reanalysis data were used in the simulations. Both simulations suggested that the {sup 137}Cs concentration had been reduced to the pre-Fukushima level around 2.5 years after the disaster. The intercomparison revealed that meso-scale eddies accompanied by the Kuroshio Extension may have efficiently diluted the radioactive cesium at the sea surface. The meso-scale eddies also played an important role in transporting the surface radioactive cesium into the intermediate layer. (author)

  17. Decontaminating method for radioactive contaminant

    International Nuclear Information System (INIS)

    Suzuki, Ken-ichi.

    1994-01-01

    After decontamination of radioactive contaminates with d-limonene, a radioactive material separating agent not compatible with liquid wastes caused by decontamination is added to the liquid wastes. Then after stirring, they are stood still to be separated into two phases, and the radioactive materials in the liquid waste phase caused by decontamination are transferred to the phase of the radioactive material separating agent. With such procedures, they can satisfactorily be separated into two phases of d-limonene and the radioactive material separating agent. Further, d-limonene remaining after the separation can be used again as a decontaminating agent for radioactive contaminates. Therefore, the amount of d-limonene to be used can be reduced, to lower the cost for cleaning, thereby enabling to reduce the amount of radioactive wastes formed. (T.M.)

  18. Interpretation of the concepts of ALARA and bat for radioactive effluent releases from nuclear installation

    International Nuclear Information System (INIS)

    Chen Xiaoqiu

    2009-01-01

    Based on the understanding of the important concepts of both ALARA and BAT associated with the characteristics of effluent releases from the existing nuclear installations and the abatement techniques for effluents, this paper elaborates the principle of controlling radioactive effluent concentration from nuclear installation, that is based on the BAT focusing on the abatement techniques for effluents, introduces the good practice in the projects, and optimize the effluent releases with account taken of external factors such as the site condition. (authors)

  19. Limitation of releases of radioactive effluents for nuclear power plants in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Tolksdorf, P.; Buehling, A.

    1981-01-01

    Empirical values relating to the effluents of nuclear power plants in the Federal Republic of Germany are now available. These values cover a period of several years of operation. The measured emissions of radioactive substances are often very much below the maximum permissible values, based on the dose limits for the environment stipulated in the legal regulations. Extensive technical and administrative measures contribute to the reduction of radioactive effluents. Furthermore, additional possibilities for improvement are mentioned which may lead to a further reduction of radioactive effluents. These are derived from investigations into the release of radioactive substances in nuclear power plants. The licensing procedure in the Federal Republic of Germany in fixing discharge limits is outlined. Proposals are made concerning licence values which may be determined for the radioactive effluents in modern standardized nuclear power plants with light-water reactors. The resulting radiation exposures are quoted for a typical nuclear power plant site. (author)

  20. Specific equipment for the transfer and transport of radioactive liquids

    International Nuclear Information System (INIS)

    Leblais, R.

    1983-01-01

    Safety regulations impose a system of transport high-level radioactive liquids which excludes all risk of accidental projections. Ets. LEMER have collaborated with the A.E.C. for the industrial manufacture of more than 250 pieces of equipment for FRANCE and for 13 nuclear centres abroad. This equipment can be classified in two categories: - CENDRILLON containers which must be placed in special shock-proof and fire-proof shells for transport on public roads; - containers mounted on trailors provided with their own shock-proof and fire-proof protection

  1. Evaluation 2000 and regulation and method. Release monitoring and environmental surveillance around Cea centers

    International Nuclear Information System (INIS)

    2001-06-01

    This publication counts for the year 2000 for the evaluation of liquid and gaseous radioactive effluents releases and the radioactivity levels measured in the vicinity of Cea centers, through the air, water, vegetation and milk surveillance. An analysis of the results from 1996 to 2000 allows to follow their evolution. A second booklet develops the sampling and measurement methods made on effluents in environment. It present besides the regulation applied to effluents monitoring. (N.C.)

  2. Compliance of the Vaalputs national radioactive waste disposal facility to a frequency-magnitude release criterion as required for licensing

    International Nuclear Information System (INIS)

    Adrian, H.W.W.; Gerber, H.H.; Kruger, J.; Weygand, J.

    1986-01-01

    Accidental releases of radioactivity from the Vaalputs nuclear waste repository have been quantified and release frequencies have been attached to a number of accident scenarios of human or natural origin. These have then been compared to a frequency-magnitude release criterion according to South African licensing requirements. It was shown that the criterion was applicable in three release bands. In two of these the criterion was met by some orders of magnitude. In the third band the permitted release frequency was a factor 55 below the limit in spite of pessimistic release assumptions

  3. New Approaches to Cleaning Liquid Radioactive Waste

    Directory of Open Access Journals (Sweden)

    Zabulonov, Yu.L.

    2015-05-01

    Full Text Available The industrial cleaning methods of liquid radioactive waste and technologically contaminated solutions, which contain heavy metals and radionuclides, are considered. It is shown that in the case when heavy metal ions exclusively exist in ionic form, the cleaning method with highest efficiency is electrodialysis. In the case when components, which must be removed, are in ionic and colloidal forms at the same time, the previous destruction of colloidal and organic matter (method of hydrodynamic cavitation, lowtemperature plasma etc is necessary. The developed «PTANK» method enables an effective purification of multicomponent metalcontaining man-made solutions, which contain additionally organic substances and complexes. Development of advanced membrane technologies, creation of complex recycling schemes and their synergistic combination will provide an opportunity to achieve deep cleaning of technologically contaminated solutions and minimize the amount of secondary wastes.

  4. A process for solidifying radioactive liquid waste

    International Nuclear Information System (INIS)

    Mergan, L.M.; Cordier, J.-P.

    1981-01-01

    In a process for solidifying radioactive liquid waste, its pH is adjusted, solids precipitated and then it is concentrated to about 50% solids content using a thin film evaporator, the concentrate then being dried to powder in a heated mixer. The mixer has a heated wall and working means, e.g. a rotor and helical screw, to shear the dried concentrate from the internal walls, subdivide it into a dry particulate powder, and advance the powder to the mixer outlet. The dried particles are then encapsulated in a suitable matrix. Vapour from the mixer and evaporator is condensed and recycled after any particles have been removed from it. The mixer may both dry the concentrate and mix the dry particles with the encapsulating matrix, and possibly, part of the mixer may be used for pH adjustment and precipitation. (author)

  5. Mathematical modeling of radionuclide release through a borehole in a radioactive waste repository

    International Nuclear Information System (INIS)

    Choi, Heui Joo

    1996-02-01

    The effects of inadvertent human intrusion as a form of direct drilling into a radioactive waste repository are discussed in this thesis. It has been mentioned that the inadvertent direct drilling into the repository could provide a release pathway for radionuclides even with its low occurrence probability. The following analyses are carried out regarding the problem. The maximum concentration in a water-filled borehole penetrating a repository is computed with a simple geometry. The modeling is based upon the assumption of the diffusive mass transfer in the waste forms and the complete mixing in the borehole. It is shown that the maximum concentrations of six radionuclides in the borehole could exceed the Maximum Permissible Concentration. Also, the diffusive mass transport in a water-filled borehole is investigated with a solubility-limited boundary condition. An analytic solution is derived for this case. Results show that the diffusive mass transport is fast enough to justify the assumption of the complete mixing compared with the considered time span. The axial diffusive mass transport along a water-filled borehole is modeled to compute the release rate taking account of the rock matrix diffusion. The results show that the release of short-lived radionuclides are negligible due to the low concentration gradient in early time and the rock matrix diffusion. The release rates of four long-lived radionuclides are computed. It is also shown that the model developed could be applied to a borehole at a non-cylindrically shaped repository and the off-center drilling of a cylindrical repository. The release rates of long-lived nuclides through a porous material-filled borehole are computed. The results show that the release of all the long-lived nuclides is negligible up to half million years in the case that the borehole is filled with the porous material. The radiological effects of the nuclides released through the borehole penetrating the repository are computed

  6. Regulatory analysis on criteria for the release of patients administered radioactive material. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, S.; McGuire, S.A.

    1997-02-01

    This regulatory analysis was developed to respond to three petitions for rulemaking to amend 10 CFR parts 20 and 35 regarding release of patients administered radioactive material. The petitions requested revision of these regulations to remove the ambiguity that existed between the 1-millisievert (0.1-rem) total effective dose equivalent (TEDE) public dose limit in Part 20, adopted in 1991, and the activity-based release limit in 10 CFR 35.75 that, in some instances, would permit release of individuals in excess of the current public dose limit. Three alternatives for resolution of the petitions were evaluated. Under Alternative 1, NRC would amend its patient release criteria in 10 CFR 35.75 to match the annual public dose limit in Part 20 of 1 millisievert (0.1 rem) TEDE. Alternative 2 would maintain the status quo of using the activity-based release criteria currently found in 10 CFR 35.75. Under Alternative 3, the NRC would revise the release criteria in 10 CFR 35.75 to specify a dose limit of 5 millisieverts (0.5 rem) TEDE.

  7. Regulatory analysis on criteria for the release of patients administered radioactive material. Final report

    International Nuclear Information System (INIS)

    Schneider, S.; McGuire, S.A.

    1997-02-01

    This regulatory analysis was developed to respond to three petitions for rulemaking to amend 10 CFR parts 20 and 35 regarding release of patients administered radioactive material. The petitions requested revision of these regulations to remove the ambiguity that existed between the 1-millisievert (0.1-rem) total effective dose equivalent (TEDE) public dose limit in Part 20, adopted in 1991, and the activity-based release limit in 10 CFR 35.75 that, in some instances, would permit release of individuals in excess of the current public dose limit. Three alternatives for resolution of the petitions were evaluated. Under Alternative 1, NRC would amend its patient release criteria in 10 CFR 35.75 to match the annual public dose limit in Part 20 of 1 millisievert (0.1 rem) TEDE. Alternative 2 would maintain the status quo of using the activity-based release criteria currently found in 10 CFR 35.75. Under Alternative 3, the NRC would revise the release criteria in 10 CFR 35.75 to specify a dose limit of 5 millisieverts (0.5 rem) TEDE

  8. System for disposing of radioactive waste

    International Nuclear Information System (INIS)

    Gablin, K.A.; Hansen, L.J.

    1979-01-01

    A system is described for disposing of radioactive waste material from nuclear reactors by solidifying the liquid components to produce an encapsulated mass adapted for disposal by burial. The method contemplates mixing of radioactive waste materials, with or without contained solids, with a setting agent capable of solidifying the waste liquids into a free standing hardened mass, placing the resulting liquid mixture in a container with a proportionate amount of a curing agent to effect solidification under controlled conditions, and thereafter burying the container and contained solidified mixture. The setting agent is a water-extendable polymer consisting of a suspension of partially polymerized particles of urea formaldehyde in water, and the curing agent is sodium bisulfate. Methods are disclosed for dewatering slurry-like mixtures of liquid and particulate radioactive waste materials, such as spent ion exchange resin beads, and for effecting desired distribution of non-liquid radioactive materials in the central area of the container prior to solidification, so that the surrounding mass of lower specific radioactivity acts as a partial shield against higher radioactivity of the non-liquid radioactive materials. The methods also provide for addition of non-radioactive filler materials to dilute the mixture and lower the overall radioactivity of the hardened mixture to desired Lowest Specific Activity counts. An inhibiting agent is added to the liquid mixture to adjust the solidification time, and provision is made for adding additional amounts of setting agent and curing agent to take up any free water and further encapsulate the hardened material within the container. 30 claims

  9. System for disposing of radioactive waste

    International Nuclear Information System (INIS)

    Gablin, K.A.; Hansen, L.J.

    1977-01-01

    A system is described for disposing of radioactive waste material from nuclear reactors by solidifying the liquid components to produce an encapsulated mass adapted for disposal by burial. The method contemplates mixing of radioactive waste materials, with or without contained solids, with a setting agent capable of solidifying the waste liquids into a free standing hardened mass, placing the resulting liquid mixture in a container with a proportionate amount of a curing agent to effect solidification under controlled conditions, and thereafter burying the container and contained solidified mixture. The setting agent is a water-extendable polymer consisting of a suspension of partially polymerized particles of urea formaldehyde in water, and the curing agent is sodium bisulfate. Methods are disclosed for dewatering slurry-like mixtures of liquid and particulate radioactive waste materials, such as spent ion exchange resin beads, and for effecting desired distribution of non-liquid radioactive materials in the central area of the container prior to solidification, so that the surrounding mass of lower specific radioactivity acts as a partial shield against higher radioactivity of the non-liquid radioactive materials. The methods also provide for addition of non-radioactive filler materials to dilute the mixture and lower the overall radioactivity of the hardened mixture to desired Lowest Specific Activity counts. An inhibiting agent is added to the liquid mixture to adjust the solidification time, and provision is made for adding additional amounts of setting agent and curing agent to take up any free water and further encapsulate the hardened material within the container

  10. Chemical treatment of radioactive liquid wastes from medical applications; Tratamiento de desechos radiactivos liquidos de aplicaciones medicas

    Energy Technology Data Exchange (ETDEWEB)

    Castillo A, J

    1996-12-31

    This work is a study about the treatment of the most important radioactive liquid wastes from medical usages, generated in medical institutions with nuclear medicine services. The radionuclides take in account are {sup 32} P, {sup 35} S, {sup 125} I. The treatments developed and improved were specific chemical precipitations for each one of the radionuclides. This work involve to precipitate the radionuclide from the liquid waste, making a chemical compound insoluble in the aqueous phase, for this process the radionuclide stay in the precipitate, lifting the aqueous phase with a very low activity than the begin. The {sup 32} P precipitated in form of Ca{sub 3} {sup 32} P O{sub 4} and Ca{sub 2} H {sup 32} P O{sub 4} with a value for Decontamination Factor (DF) at the end of the treatment of 32. The {sup 35} S was precipitated in form of Ba{sup 35} SO{sub 4} with a DF of 26. The {sup 125} I was precipitated in Cu {sup 125} I to obtain a DF of 24. The results of the treatments are between the limits given for the International Atomic Energy Agency and the 10 Code of Federal Regulation 20, for the safety release at the environment. (Author).

  11. The Separation Of The Uranium On The Liquid Radioactive Wastes By Using Zeolites

    International Nuclear Information System (INIS)

    Budiyono, ME; Peornomo, Herry; S, Djoko

    2000-01-01

    The experimental investigation on the separation of uranium of liquid radioactive wastes by using zeolite has been carried out. The aims of this investigation was to separate uranium and to determine the solid content of the liquid radioactive waste. This investigation was carried out because zeolite could be used as a good sorption material. The investigation parameters were the number of zeolites, the pH of wastes and the grain size of zeolites. The number of zeolites was varied from 10-100 g, the waste pH was varied from 1-12 and the grain size of zeolites was varied from 20/40-100/200 mesh, each parameters to included in the waste. The conclusion that could be drawn from this investigation were that the best result of the 80 g zeolites used, pH was 9, and the grain size of zeolites was 60/80 mesh, the solid content was 119,46 g/l and efficiency of separation was 81,74 %. As a rule, the solid content in the evaporator process should be about 200-300 g/l

  12. Using copper hexacyanoferrate (II) impregnated zeolite for cesium removal from radioactive liquid waste

    International Nuclear Information System (INIS)

    Fumio, K.; Kenji, M.

    1982-01-01

    Experiments were performed to obtain fundamental data on cesium ion removal characteristics of metal hexacyanoferrate (II) impregnated zeolite in radioactive liquid waste containing a large amount of sodium sulfate. Copper hexacyanoferrate (II) impregnated zeolite (CuFZ) was prepared and showed a high selectivity for cesium ion. The material was suitable for use in an ion exchange column. This exchanger could selectively and efficiently remove the cesium even if there is 15 wt% Na 2 SO 4 in the solution. Cesium removal ability and stability of CuFZ were excellent over a wide pH range between 1.5 and 10. The cesium ion exchange ability was not influenced by the presence of the alkali metal ions, calcium and magnesium, and carbonate ions even at concentrations 25 times greater than the cesium ion. However, since ammonium ion behaves similarly to cesium ion and interrupts latter ion adsorption, the presence of ammonium ion is not desirable. The CuFZ offers the possibility of separating and removing cesium from liquid wastes produced in facilities handling radioactive materials

  13. Complete release from regulatory control via the density of radioactive contamination of soil of the Chornobyl exclusion zone

    International Nuclear Information System (INIS)

    Bondarenko, Oleg; Fadeev, Mykhaylo; Kireev, Serhiy; Proskura, Mykola

    2008-01-01

    Full text: In this work a general procedure of establishment the criterion of complete release from regulatory control via the density of radioactive contamination of soil of alone areas of the Chornobyl exclusion zone is represented by the following three stages, namely: 1) Justification of the dose criterion of complete release by applying the fundamental approaches of regulating the prolonged exposure of the public on a basement of ICRP Publication No. 82; 2) Justification of a procedure for establishment of dose constraint through evaluation of the statistical distribution of a controlled radiation value through determination of the high boundary of the confidence interval; 3) Generalization of conversion coefficients (via the density of radioactive contamination of soil) and coverage coefficients for the dose forming factors of the public (inhalation intake of transuranium radionuclides both at natural and technogenic dust resuspension, peroral intake of 137 Cs and 90 Sr via food stuff, external exposure from 137 Cs); on a basis of these coefficients average doses and dose constraint of a critical group of the public are calculated. As it is shown in the work, the generalized criterion of complete release from regulatory control via the density of radioactive contamination of soil of the Chornobyl exclusion zone can be defined by dividing the recommended ICRP dose clearance level for situation of the post-accidental prolonged exposure of a critical group of the public, namely, 0.3 mSv·year -1 to the dose conversion coefficient (i.e. transfer from the superficial soil contamination of radioactivity to the dose constraint for the Chornobyl exclusion zone), namely, 12.9 μSv·year -1 /((kBq·M -2 ). Thus, a level of the density of radioactive contamination of soil, that provides the complete release from regulatory control makes 23.3 kBq·M -2 . For completion of the work on justification of the criterion of complete territory release in the conditions of

  14. Applicability of Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) to releases of radioactive substances

    International Nuclear Information System (INIS)

    Miller, S.R.

    1987-01-01

    The Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA), commonly called Superfund, provided a $1.6 billion fund (financed by a tax on petrochemical feedstocks and crude oil and by general revenues) for the cleanup of releases of hazardous substances, including source, special nuclear or byproduct material, and other radioactive substances, from mostly inactive facilities. The US Environmental Protection Agency (EPA) is authorized to require private responsible parties to clean up releases of hazardous substances, or EPA, at its option, may undertake the cleanup with monies from the Fund and recover the monies through civil actions brought against responsible parties. CERCLA imposes criminal penalties for noncompliance with its reporting requirements. This paper will overview the key provisions of CERCLA which apply to the cleanup of radioactive materials

  15. Principles for establishing limits for the release of radioactive materials into the environment

    International Nuclear Information System (INIS)

    1978-01-01

    The document provides a basic consideration of concepts and principles for use by national authorities in setting limits for planned releases of radioactive material. The following topics are discussed general concepts, assessment of dose to the critical group, assessment of collective dose commitments, application of optimization techniques to the determination of discharge limits, explanation and application of the concept of collective dose commitment, discharge limitations based on concentration indices

  16. Determining the level of gross alpha and beta radioactivity of water from Marilao river using liquid scintillation counter

    International Nuclear Information System (INIS)

    Cruz, J.; Magtaka, J.; Balisi, R.; Castaneda, Soledad; De Vero, J.

    2009-01-01

    This study aims to determine the level of gross alpha- and beta- radioactivity present in the Marilao River. Liquid Scintillation Counter was used to detect samples radioactivity. Water samples were taken along the different spots of Marilao River. The results showed that the radioactivity are below the Philippine National Standard for Drinking Water (PNSDW) which is 0.1 Bq/L for gross alpha activity and 1.0 Bq/L for gross beta activity set by the Dept. of Health. Hence water samples from Marilao River, is safe in terms of the level of radioactivity levels.(author)

  17. Radiation protection requirements in the limitation of the release of radioactive effluents

    International Nuclear Information System (INIS)

    Beninson, D.

    1978-01-01

    The paper summarizes the requirements of radiation protection, as presented in the new ICRP recommendation, in relation to the limitation of the release of radioactive effluents. The concepts of effective dose equivalent and collective dose equivalent are used in the presentation of the optimization procedures and the dose limitation to individuals. The dose commitment is used for the procedures applied to control future exposures. An appendix describes the basic concepts and quantities used in assessments of human exposures and risks. (author)

  18. Updated synthesis of knowledge related to the impact of radioactive releases from the damaged nuclear site of Fukushima Dai-Ichi on the marine environment - 13 July 2012

    International Nuclear Information System (INIS)

    2012-01-01

    Illustrated by tables, figures and graphs, this report first comments the evolution of the radioactive pollution of sea water: main radionuclides notices in sea water since the 21 March 2011 (with two main origins: atmospheric fallouts, liquid releases from the site), evolution of sea water contamination along the coast at the vicinity of the power station, simulation of caesium 137 dispersion in sea water off Japan. Then, it comments the evolution of the presence of radionuclides in sediments and in marine species (species with contamination level exceeding the admissible level for consumption, concentrations noticed by sea animals, expected evolution for marine species, other published data)

  19. Order of the 15. of September 2010 approving the decision no 2010-DC-0188 by the Nuclear Safety Authority on the 7. of July 2010 specifying to Electricite de France-Societe Anonyme (EDF-SA) the limits of releases in the environment of liquid and gaseous effluents for the operation of Flamanville 1 (INB n 108), Flamanville 2 (INB n 109) and Flamanville 3 (INB n 167) reactors

    International Nuclear Information System (INIS)

    2010-01-01

    This legal publication contains references to the different legal and official documents (codes, orders, minister's opinion, public surveys, administrative authorizations, local community opinion) at the root of this specification of limits related to releases in the environment of liquid and gaseous effluents by the three Flamanville nuclear reactors operated by Electricite de France-Societe Anonyme (EDF-SA). Tables present the limits for different gaseous and liquid radioactive emissions (carbon 14, hydrogen 3, iodine, radioactive rare earths, and other beta and gamma emitters) from these different installations, but also the maximum admitted concentrations for gaseous chemical effluents (boric acid, hydrazine hydrate, morpholine, phosphates, nitrates, various metals, and so on). The limits for thermal releases are also specified

  20. U.S. Department of Energy's initiatives for proliferation prevention in Russia: results of radioactive liquid waste treatment project, year 2

    International Nuclear Information System (INIS)

    Pokhitonov, Y.; Kamachev, V.; Kelley, D.

    2010-10-01

    The objective of the project is to engage weapons scientists with training and research programs at selected nuclear sites in Russia and apply high technology polymers to immobilize legacy ILW and HLW liquids that have posed environmental challenges over the years. One compelling advantage of the projects is that V.G. Khlopin Radium Institute and Pacific Nuclear Solutions have been engaged in applied research for seven years to validate the performance and effectiveness of the polymer technology for use with radioactive liquids. With conclusive results of the research work on sixty active and simulant waste streams, the project can focus on actual applications of the technology at Ozersk (Mayak), Sever sk (SCC) Zheleznogorsk (MCC) and Gatchyna rather that on pure research. The long term objective of the project is find viable waste management solutions for serious radioactive and chemical contamination that has existed in Russia and the U. S. for several decades. The polymer technologies may be applied to all radioactive liquid. This paper summarizes the experimental work of the immobilization process and data definition of the most effective polymer compositions in addition to determining the optimum polymer to liquid ratios for economic considerations. (Author)

  1. Liquid crystalline systems for transdermal delivery of celecoxib: in vitro drug release and skin permeation studies.

    Science.gov (United States)

    Estracanholli, Eder André; Praça, Fabíola Silva Garcia; Cintra, Ana Beatriz; Pierre, Maria Bernadete Riemma; Lara, Marilisa Guimarães

    2014-12-01

    Liquid crystalline systems of monoolein/water could be a promising approach for the delivery of celecoxib (CXB) to the skin because these systems can sustain drug release, improve drug penetration into the skin layers and minimize side effects. This study evaluated the potential of these systems for the delivery of CXB into the skin based on in vitro drug release and skin permeation studies. The amount of CXB that permeated into and/or was retained in the skin was assayed using an HPLC method. Polarizing light microscopy studies showed that liquid crystalline systems of monoolein/water were formed in the presence of CXB, without any changes in the mesophases. The liquid crystalline systems decreased drug release when compared to control solution. Drug release was independent of the initial water content of the systems and CXB was released from cubic phase systems, irrespective of the initial water content. The systems released the CXB following zero-order release kinetics. In vitro drug permeation studies showed that cubic phase systems allowed drug permeation and retention in the skin layers. Cubic phase systems of monoolein/water may be promising vehicles for the delivery of CXB in/through the skin because it improved CXB skin permeation compared with the control solution.

  2. Liquid Effluent Monitoring Information System (LEMIS) test plans release 1.0

    International Nuclear Information System (INIS)

    Adams, R.T.

    1994-01-01

    The Liquid Effluent Monitoring Information System (LEMIS) is being developed as the organized information repository facility in support of the liquid effluent monitoring requirements of the Tri-Party Agreement. It is necessary to provide an automated repository into which the results from liquid effluent sampling will be placed. This repository must provide for effective retention, review, and retrieval of selected sample data by authorized persons and organizations. This System Architecture document is the aggregation of the DMR P+ methodology project management deliverables. Together they represent a description of the project and its plan through four Releases, corresponding to the definition and prioritization of requirements defined by the user

  3. Liquid Effluent Monitoring Information System (LEMIS) test plans release 1.1

    International Nuclear Information System (INIS)

    Adams, R.T.

    1994-01-01

    The Liquid Effluent Monitoring Information System (LEMIS) is being developed as the organized information repository facility in support of the liquid effluent monitoring requirements of the Tri-Party Agreement. It is necessary to provide an automated repository into which the results from liquid effluent sampling will be placed. This repository must provide for effective retention, review, and retrieval of selected sample data by authorized persons and organizations. This System Architecture document is the aggregation of the DMR P+ methodology project management deliverables. Together they represent a description of the project and its plan through four Releases, corresponding to the definition and prioritization of requirements defined by the user

  4. Microbiological aspects of safety in radioactive waste management

    International Nuclear Information System (INIS)

    Ershov, B.G.; Safonov, A.V.; Nazina, T.N.; Gorbunova, O.A.

    2012-01-01

    In long-term storage and/or disposal of radioactive waste, microbiological processes play an important, and in some cases a vital role. The article discusses the issues of microbiological processes in underground liquid LLW repositories and microbiological destruction of cemented radwaste. It is shown that biological additives to cement matrices can be used to effectively prevent the occurrence of microbiological processes, increasing reliability of engineering barriers that block release of radionuclides into the areas adjacent to the repositories [ru

  5. Drug Release and Skin Permeation from Lipid Liquid Crystalline Phases

    Science.gov (United States)

    Costa-Balogh, F. O.; Sparr, E.; Sousa, J. J. S.; Pais, A. A. C. C.

    We have studied drug release and skin permeation from several different liquid crystalline lipid formulations that may be used to control the respective release rates. We have studied the release and permeation through human skin of a water-soluble and amphiphilic drug, propranolol hydrochloride, from several formulations prepared with monoolein and phytantriol as permeation enhancers and controlled release excipients. Diolein and cineol were added to selected formulations. We observed that viscosity decreases with drug load, wich is compatible with the occurrence of phase changes. Diolein stabilizes the bicontinuous cubic phases leading to an increase in viscosity and sustained release of the drug. The slowest release was found for the cubic phases with higher viscosity. Studies on skin permeation showed that these latter formulations also presented lower permeability than the less viscous monoolein lamellar phases. Formulations containing cineol originated higher permeability with higher enhancement ratios. Thus, the various formulations are adapted to different circumstances and delivery routes. While a slow release is usually desired for drug sustained delivery, the transdermal route may require a faster release. Lamellar phases, which are less viscous, are more adapted to transdermal applications. Thus, systems involving lamellar phases of monoolein and cineol are good candidates to be used as skin permeation enhancers for propranolol hydrochloride.

  6. Method of transporting radioactive slurry-like wastes

    Energy Technology Data Exchange (ETDEWEB)

    Kamiya, K; Yusa, H; Sugimoto, Y

    1975-06-30

    The object is to prevent blockage of a transporting tube to positively and effectively transport radioactive slurry wastes. A method of transporting radioactive slurry-like wastes produced in an atomic power plant, wherein liquid wastes produced in the power plant are diluted to form into a driving liquid, by which said radioactive slurry-like wastes are transported within the pipe, and said driving liquid is recovered as the liquid waste.

  7. Spectrophotometer as a tool for analysis of various species in radioactive liquid wastes (Preprint No. RA-27)

    International Nuclear Information System (INIS)

    Yeotikar, R.G.; Kaushik, C.P.; Raj, Kanwar

    1988-02-01

    For the immobilisation of high level radioactive liquid waste in a glass matrix, the analysis of different species in the waste is the primary requirement for fixing the base composition of glass. The commercially available spectrophotometer has been remotised to facilitate the analysis of different radioactive streams. Methods for analysis of constituents such as Fe, Mo, Cr, U, total lanthanides, Zr and PO4 have been established. (author)

  8. Population dose commitments due to radioactive releases from nuclear-power-plant sites in 1978

    International Nuclear Information System (INIS)

    Peloquin, R.A.; Schwab, J.D.; Baker, D.A.

    1982-06-01

    Population radiation dose commitments have been estimated from reported radionuclide releases from commercial power reactors operating during 1978. Fifty-year dose commitments from a one-year exposure were calculated from both liquid and atmospheric releases for four population groups (infant, child, teen-ager and adult) residing between 2 and 80 km from each site. This report tabulates the results of these calculations, showing the dose commitments for both liquid and airborne pathways for each age group and organ. Also included for each site is a histogram showing the fraction of the total population within 2 to 80 km around each site receiving various average dose commitments from the airborne pathways. The total dose commitment from both liquid and airborne pathways ranged from a high of 200 person-rem to a low of 0.0004 person-rem with an arithmetic mean of 14 person-rem. The total population dose for allsites was estimated at 660 person-rem for the 93 million people considered at risk. The average individual dose commitment from all pathways on a site basis ranged from a low of 3 x 10 -6 mrem to a high of 0.08 mrem. No attempt was made in this study to determine the maximum dose commitment received by any one individual from the radionuclides released at any of the sites

  9. Population dose commitments due to radioactive releases from Nuclear-Power-Plant Sites in 1979

    International Nuclear Information System (INIS)

    Baker, D.A.; Peloquin, R.A.

    1982-12-01

    Population radiation dose commitments have been estimated from reported radionuclide releases from commercial power reactors operating during 1979. Fifty-year dose commitments from a one-year exposure were calculated from both liquid and atmospheric releases for four population groups (infant, child, teen-ager and adult) residing between 2 and 80 km from each site. This report tabulates the results of these calculations, showing the dose commitments for both liquid and airborne pathways for each age group and organ. Also included for each site is a histogram showing the fraction of the total population within 2 to 80 km around each site receiving various average dose commitments from the airborne pathways. The total dose commitment from both liquid and airborne pathways ranged from a high of 1300 person-rem to a low of 0.0002 person-rem with an arithmetic mean of 38 person-rem. The total population dose for all sites was estimated at 1800 person-rem for the 94 million people considered at risk. The average individual dose commitment from all pathways on a site basis ranged from a low of 2 x 10 - 6 mrem to a high of 0.7 mrem. No attempt was made in this study to determine the maximum dose commitment received by any one individual from the radionuclides released at any of the sites

  10. Characterization of radioactive organic liquid wastes; Caracterizacion de desechos liquidos organicos radiactivos

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez A, I.; Monroy G, F.; Quintero P, E.; Lopez A, E.; Duarte A, C., E-mail: ivonne-arce@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    With the purpose of defining the treatment and more appropriate conditioning of radioactive organic liquid wastes, generated in medical establishments and research centers of the country (Mexico) and stored in drums of 208 L is necessary to characterize them. This work presents the physical-chemistry and radiological characterization of these wastes. The samples of 36 drums are presented, whose registrations report the presence of H-3, C-14 and S-35. The following physiochemical parameters of each sample were evaluated: ph, conductivity, density and viscosity; and analyzed by means of gamma spectrometry and liquid scintillation, in order to determine those contained radionuclides in the same wastes and their activities. Our results show the presence of H-3 (61%), C-14 (13%) and Na-22 (11%) and in some drums low concentrations of Co-60 (5.5%). In the case of the registered drums with S-35 (8.3%) does not exist presence of radioactive material, so they can be liberated without restriction as conventional chemical wastes. The present activities in these wastes vary among 5.6 and 2312.6 B g/g, their ph between 2 and 13, the conductivities between 0.005 and 15 m S, the densities among 1.05 and 1.14, and the viscosities between 1.1 and 39 MPa. (Author)

  11. Ion exchanger material based on Titanium phosphate for liquid radioactive waste treatment

    Energy Technology Data Exchange (ETDEWEB)

    Maslova, M.; Gerasimova, L. [Tananaev Institute of Chemistry and Technology of Rare Elements and Mineral Resources, Kola Scientific Center, Russian Academy of Sciences, Apatity (Russian Federation)

    2013-07-01

    A comparative study of the physicochemical and service properties of samples of Ti(OH) 1.36(HPO4) 1.32 * 2.3H2O sorbent in the finely dispersed and granulated forms, mastered for commercial production, was made. The sorption of Cs and Sr cations from solutions of various, compositions was studied in batch experiments, and the diffusion coefficients of the exchanging ions were determined. The hydrolytic stability of the. sorbents was examined with the aim to determine the optimal operation conditions. Experiments showed that the cation exchangers based on titanium phosphate are the most efficient in removal from liquid radioactive waste of induced radioactive isotopes of corrosion products, which is due to formation of weakly dissociating compounds of nonferrous metal ions with functional groups of the ion exchangers in the sorbent phase. (author)

  12. Ecological solution of the problem of handling liquid radioactive wastes - Lr (by the example of Flue SSC RF RIA R)

    International Nuclear Information System (INIS)

    Polyakov, V.I.; Bukvich, B.A.

    2006-01-01

    A sharp reduction of nuclear waste amounts is possible if their elements are considered as source material of atomic complexes - SMAC. The prospect of their possible salvaging will require technological changes and ensuring safety of storage of the material till the need arises. Long experience in deep liquid radioactive waste disposal and accounting, calculations, and motivations demonstrate that a corresponding choice of geological formations makes it possible to abandon liquid radioactive waste solidification and ensure their isolation from environment when the most rigid radiation safety requirements are fulfilled. (author)

  13. Hazard of radioactive releases resulted from coal burning; Opasnost` vysvobozhdeniya radioaktivnykh produktov pri szhiganii uglya

    Energy Technology Data Exchange (ETDEWEB)

    Gabbard, V

    1995-09-01

    Consideration is given to the data, pointing to the fact, that coal-burning power plants release of radioactive substances, contained in gaseous wastes, is not less, than the same one of nuclear power plants. The necessity of regulating emission of these substance in atmosphere by analogy with nuclear power industry is shown. 1 fig.

  14. Methodology for assessing the radiological consequences of radioactive releases from the BPX Facility at PPPL

    International Nuclear Information System (INIS)

    McKenzie-Carter, M.A.; Lyon, R.E.; Rope, S.K.

    1991-04-01

    This report contains information to support the Environmental Assessment for the Burning Plasma Experiment (BPX) Project proposed for the Princeton Plasma Physics Laboratory (PPPL). The assumptions and methodology used to assess the impact to members of the public from operational and accidental releases of radioactive material from the proposed BPX during the operational period of the project are described. A description of the tracer release tests conducted at PPPL by NOAA is included; dispersion values from these tests are used in the dose calculations. Radiological releases, doses, and resulting health risks are calculated and summarized. The computer code AIRDOS- EPA, which is part of the computer code system CAP-88, is used to calculate the individual and population doses for routine releases; FUSCRAC3 is used to calculate doses resulting from off-normal releases where direct application of the NOAA tracer test data is not practical. Where applicable, doses are compared to regulatory limits and guideline values. 48 refs., 16 tabs

  15. Conceptual design and development of high-activity radioactive liquid packaging (summary)

    International Nuclear Information System (INIS)

    Riley, D.L.; McCoy, J.C.; Edwards, W.S.

    1994-08-01

    Environmental remediation and disposal of US Department of Energy radioactive liquid waste require analytical support, characterization, process development, testing, demonstration, and stabilization. In support of these diverse activities, there is a need to transport varying quantities of Type B high-activity liquid (HAL). To date, except for quantities of 50 ml (1.7 oz), there has never been, a US, Nuclear Regulatory Commission-licensed liquid Type B package available to support these remediation activities. In an effort to develop suitable packaging for large volumes of HAL, an investigation into packaging alternatives that would facilitate such transfers is under way. In, past and present studies, a spent fuel shipping cask fitted with a high-integrity pressure vessel has been determined to be the most viable concept for large volume HAL shipments. One concept that was investigated utilized the Pacific Nuclear 125-B shipping container and has been shown to meet the strUctural, thermal, shielding, and criticality conditions for HAL. The results of these investigations are being extended to develop the concept into the HAL packaging system

  16. Final Treatment Center Project for Liquid and Wet Radioactive Waste in Slovakia

    International Nuclear Information System (INIS)

    Kravarik, K.; Stubna, M.; Pekar, A.; Krajc, T.; Zatkulak, M.; Holicka, Z.; Slezak, M.

    2006-01-01

    The Final Treatment Center (FTC) for Mochovce nuclear power plant (NPP) is designed for treatment and final conditioning of radioactive liquid and wet waste produced from plant operation. Mochovce NNP uses a Russian VVER-440 type reactor. Treated wastes comprise radioactive concentrates, spent resin and sludge. VUJE Inc. as an experienced company in field of treatment of radioactive waste in Slovakia has been chosen as main contractor for technological part of FTC. This paper describes the capacity, flow chart, overall waste flow and parameters of the main components in the FTC. The initial project was submitted for approval to the Slovak Electric plc. in 2003. The design and manufacture of main components were performed in 2004 and 2005. FTC construction work started early in 2004. Initial non-radioactive testing of the system is planned for summer 2006 and then radioactive tests are to be followed. A one-year trial operation of facility is planned for completion in 2007. SE - VYZ will be operates the FTC during trial operation and after its completion. SE - VYZ is subsidiary company of Slovak Electric plc. and it is responsible for treatment with radioactive waste and spent fuel in the Slovak republic. SE - VYZ has, besides of other significant experience with operation of Jaslovske Bohunice Treatment Centre. The overall capacity of the FTC is 870 m 3 /year of concentrates and 40 m 3 /year of spent resin and sludge. Bituminization and cementation were provided as main technologies for treatment of these wastes. Treatment of concentrate is performed by bituminization. Concentrate and bitumen are metered into a thin film evaporator with rotating wiping blades. Surplus water is evaporated and concentrate salts are embedded in bitumen. Bitumen product is discharged into 200 l steel drums. Spent resin and sludge are decanted, dried and mixed with bitumen. These mixtures are also discharged into 200 l steel drums. Drums are moved along bituminization line on a roller

  17. Device for sampling radioactive and aggressive liquid and vaporous media

    International Nuclear Information System (INIS)

    Przibram, E.; Halm, G.

    1974-01-01

    The equipment enables the taking of samples even of radioactive media from a main pipeline in the through-flow in a closed system. A tap device is attached to the main pipeline which branches into two parts. The one branch contains the actual tap which is closed to both sides with snap closure coupling. It is only used for taking samples. The other branch bridges the tap position as a bypass so that a representative sample is always available. Both branches join up again and lead back to the main pipeline. The sampling can be used in a nuclear power plant for the determination of O 2 , CI, SiO 2 , and Cu. A millilitre collecting cylinder and a millipore filtration device can be connected to the tap for liquid sampling and solid analysis, respectively. The system can be extended to several tap positions. Permanent measuring equipment is attached to the bypass pipe to control the sample liquid. (DG) [de

  18. Development of Characterization Protocol for Mixed Liquid Radioactive Waste Classification

    International Nuclear Information System (INIS)

    Norasalwa Zakaria; Syed Asraf Wafa; Wo, Y.M.; Sarimah Mahat; Mohamad Annuar Assadat Husain

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Zakaria, Norasalwa, E-mail: norasalwa@nuclearmalaysia.gov.my [Waste Technology Development Centre, Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia); Wafa, Syed Asraf [Radioisotop Technology and Innovation, Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia); Wo, Yii Mei [Radiochemistry and Environment, Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia); Mahat, Sarimah [Material Technology Group, Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia)

    2015-04-29

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

  20. Radioactive waste management at a Liquid Metal Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Abrams, C.S.; Fryer, R.H.; Witbeck, L.C.

    1979-01-01

    This paper presents the radioactive waste production and management at a Liquid Metal Fast Breeder Reactor-II (EBR-II), which is operated for the US Department of Energy by the Argonne National Laboratory at the Idaho National Engineering Laboratory (INEL). Since this facility, in addition to supplying power has been used to demonstrate the breeder, fuel cycling, and recently operations with defective fuel elements, various categories of waste have been handled safely over some 14 years of operation. Liquid wastes are processed such that the resulting effluent can be discharged to an uncontrolled area. Solid wastes up to 10,000 R/hr are packaged and shipped contamination-free to a disposal site or interim storage with exposures to personnel approximately 10 mrem. Gaseous waste discharges are low such as 143 Ci of noble gases in 1978 and do not have a significant effect on the environment even with operations with breached fuel