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Sample records for alkaline radioactive liquid

  1. ACTINIDE-ALUMINATE SPECIATION IN ALKALINE RADIOACTIVE WASTE

    EPA Science Inventory

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

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

    EPA Science Inventory

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

  3. Radioactive liquid waste treatment facility

    SciTech Connect

    Black, R.L.

    1984-07-01

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

  4. Type A radioactive liquid sample packaging family

    SciTech Connect

    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.

  5. FINAL REPORT. ACTINIDE-ALUMINATE SPECIATION IN ALKALINE RADIOACTIVE WASTE

    EPA Science Inventory

    Investigation of behavior of actinides in alkaline media containing Al(III) showed that no aluminate complexes of actinides in oxidation states (III-VII) were formed in alkaline solutions. At alkaline precipitation (pH 10-14) of actinides in presence of Al(III) formation of alumi...

  6. Alkaline ionic liquids applied in supported ionic liquid catalyst for selective hydrogenation of citral to citronellal

    NASA Astrophysics Data System (ADS)

    Salminen, Eero; Virtanen, Pasi; Mikkola, Jyri-Pekka

    2014-02-01

    The challenge in preparation of ionic liquids containing a strong alkaline anion is to identify a suitable cation which can tolerate the harsh conditions induced by the anion. In this study, a commercial quaternary ammonium compound (quat) benzalkonium [ADBA] (alkyldimethylbenzylammonium) was used as a cation in the synthesis of different alkaline ionic liquids. In fact, the precursor, benzalkonium chloride, is a mixture of alkyldimethylbenzylammonium chlorides of various alkyl chain lengths and is commonly used in the formulation of various antiseptic products. The prepared ionic liquids were utilized as Supported Ionic Liquid Catalysts (SILCAs). Typically, a SILCA contains metal nanoparticles, enzymes or metal complexes in an ionic liquid layer which is immobilized on a solid carrier material such as an active carbon cloth (ACC). The catalysts were applied in the selective hydrogenation of citral to citronellal which is an important perfumery chemical. Interestingly, 70 % molar yield towards citronellal was achieved over a catalyst containing the alkaline ionic liquid benzalkonium methoxide.

  7. Alkaline ionic liquids applied in supported ionic liquid catalyst for selective hydrogenation of citral to citronellal.

    PubMed

    Salminen, Eero; Virtanen, Pasi; Mikkola, Jyri-Pekka

    2014-01-01

    The challenge in preparation of ionic liquids containing a strong alkaline anion is to identify a suitable cation which can tolerate the harsh conditions induced by the anion. In this study, a commercial quaternary ammonium compound (quat) benzalkonium [ADBA] (alkyldimethylbenzylammonium) was used as a cation in the synthesis of different alkaline ionic liquids. In fact, the precursor, benzalkonium chloride, is a mixture of alkyldimethylbenzylammonium chlorides of various alkyl chain lengths and is commonly used in the formulation of various antiseptic products. The prepared ionic liquids were utilized as Supported Ionic Liquid Catalysts (SILCAs). Typically, a SILCA contains metal nanoparticles, enzymes, or metal complexes in an ionic liquid layer which is immobilized on a solid carrier material such as an active carbon cloth (ACC). The catalysts were applied in the selective hydrogenation of citral to citronellal which is an important perfumery chemical. Interestingly, 70% molar yield toward citronellal was achieved over a catalyst containing the alkaline ionic liquid benzalkonium methoxide. PMID:24790972

  8. Alkaline ionic liquids applied in supported ionic liquid catalyst for selective hydrogenation of citral to citronellal

    PubMed Central

    Salminen, Eero; Virtanen, Pasi; Mikkola, Jyri-Pekka

    2014-01-01

    The challenge in preparation of ionic liquids containing a strong alkaline anion is to identify a suitable cation which can tolerate the harsh conditions induced by the anion. In this study, a commercial quaternary ammonium compound (quat) benzalkonium [ADBA] (alkyldimethylbenzylammonium) was used as a cation in the synthesis of different alkaline ionic liquids. In fact, the precursor, benzalkonium chloride, is a mixture of alkyldimethylbenzylammonium chlorides of various alkyl chain lengths and is commonly used in the formulation of various antiseptic products. The prepared ionic liquids were utilized as Supported Ionic Liquid Catalysts (SILCAs). Typically, a SILCA contains metal nanoparticles, enzymes, or metal complexes in an ionic liquid layer which is immobilized on a solid carrier material such as an active carbon cloth (ACC). The catalysts were applied in the selective hydrogenation of citral to citronellal which is an important perfumery chemical. Interestingly, 70% molar yield toward citronellal was achieved over a catalyst containing the alkaline ionic liquid benzalkonium methoxide. PMID:24790972

  9. Dynamical Correlation In Some Liquid Alkaline Earth Metals Near Melting

    NASA Astrophysics Data System (ADS)

    Thakore, B. Y.; Suthar, P. H.; Khambholja, S. G.; Gajjar, P. N.; Jani, A. R.

    2010-12-01

    The study of dynamical variables: velocity autocorrelation function (VACF) and power spectrum of liquid alkaline earth metals (Ca, Sr, and Ba) have been presented based on the static harmonic well approximation. The effective interatomic potential for liquid metals is computed using our well recognized model potential with the exchange correlation functions due to Hartree, Taylor, Ichimaru and Utsumi, Farid et al. and Sarkar et al. It is observed that the VACF computed using Sarkar et al. gives the good agreement with available molecular dynamics simulation (MD) results [Phys Rev. B 62, 14818 (2000)]. The shoulder of the power spectrum depends upon the type of local field correlation function used.

  10. Actinide-Aluminate Speciation in Alkaline Radioactive Waste

    SciTech Connect

    Dr. David L. Clark; Dr. Alexander M. Fedosseev

    2001-12-21

    Investigation of behavior of actinides in alkaline media containing AL(III) showed that no aluminate complexes of actinides in oxidation states (IIII-VIII) were formed in alkaline solutions. At alkaline precipitation IPH (10-14) of actinides in presence of AL(III) formation of aluminate compounds is not observed. However, in precipitates contained actinides (IIV)<(VI), and to a lesser degree actinides (III), some interference of components takes place that is reflected in change of solid phase properties in comparison with pure components or their mechanical mixture. The interference decreases with rise of precipitation PH and at PH 14 is exhibited very feebly. In the case of NP(VII) the individual compound with AL(III) is obtained, however it is not aluminate of neptunium(VII), but neptunate of aluminium(III) similar to neptunates of other metals obtained earlier.

  11. Method for solidifying liquid radioactive wastes

    DOEpatents

    Berreth, Julius R.

    1976-01-01

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

  12. Alkaline chemistry of transuranium elements and technetium and the treatment of alkaline radioactive wastes

    SciTech Connect

    Delegard, C.H.; Peretrukhin, V.F.; Shilov, V.P.; Pikaev, A.K.

    1995-05-01

    Goal of this survey is to generalize the known data on fundamental physical-chemical properties of TRUs and Tc, methods for their isolation, and to provide recommendations that will be useful for partitioning them from alkaline high-level wastes.

  13. Future radioactive liquid waste streams study

    SciTech Connect

    Rey, A.S.

    1993-11-01

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

  14. MODELING AN ION EXCHANGE PROCESS FOR CESIUM REMOVAL FROM ALKALINE RADIOACTIVE WASTE SOLUTIONS

    SciTech Connect

    Smith, F; Luther Hamm, L; Sebastian Aleman, S; Johnston Michael, J

    2008-08-26

    The performance of spherical Resorcinol-Formaldehyde ion-exchange resin for the removal of cesium from alkaline radioactive waste solutions has been investigated through computer modeling. Cesium adsorption isotherms were obtained by fitting experimental data using a thermodynamic framework. Results show that ion-exchange is an efficient method for cesium removal from highly alkaline radioactive waste solutions. On average, two 1300 liter columns operating in series are able to treat 690,000 liters of waste with an initial cesium concentration of 0.09 mM in 11 days achieving a decontamination factor of over 50,000. The study also tested the sensitivity of ion-exchange column performance to variations in flow rate, temperature and column dimensions. Modeling results can be used to optimize design of the ion exchange system.

  15. Conditions of accumulation of radioactive metals in the process of differentiation of ultrabasic alkaline-carbonatite rock associations

    NASA Astrophysics Data System (ADS)

    Kogarko, L. N.

    2014-07-01

    The distribution of radioactive elements in alkaline rocks from Polar Siberia and Ukraine shows that U and Th are markedly concentrated in carbonatite complex and nepheline syenite as final products of magma fractionation. Peralkaline nepheline syenites from Polar Siberia are characterized by very high contents of radioactive elements, which are close to the economic level. Radioactive elements are also concentrated in rocks of the carbonatite complex. For example, some soevites contain up to 294 × 10-4%U and 916 × 10-4% Th. In late dolomite carbonatites, the contents of radioactive elements are appreciably lower. The Th/U ratio in alkaline rocks of Polar Siberia is close to the chondrite value in primary high-Mg rocks and increases in late derivatives: phoscorite, calcite and dolomite carbonatites. The main amount of radioactive elements is contained in rare-metal accessory minerals: perovskite, pyrochlore, calzirtite, and apatite. Rock-forming minerals are distinguished by very low concentrations of radioactive elements. In alkaline series of the Chernigovka massif (Ukraine), U and Th also accumulate in the course of crystal fractionation, especially in phoscorites from the carbonatite complex. Mantle xenoliths and alkaline rocks from Ukraine reveal uranium specialization. Most likely, the discrepancy in fractionation of radioactive elements between Polar Siberia and Ukraine is caused by different geodynamic regimes of these provinces. The Mesozoic alkaline magmatism of Polar Siberia is a part of the Siberian superplume, whereas the Proterozoic alkaline complex in Ukraine is related to subduction of the oceanic crust.

  16. Reduction of INTEC Analytical Radioactive Liquid Waste

    SciTech Connect

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

    1999-06-01

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

  17. Reduction of INTEC Analytical Radioactive Liquid Wastes

    SciTech Connect

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

    1999-06-01

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

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

    PubMed

    Osmanlioglu, Ahmet Erdal

    2007-01-01

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

  19. Code System for the Radioactive Liquid Tank Failure Study.

    Energy Science and Technology Software Center (ESTSC)

    2000-01-03

    Version 01 RATAF calculates the consequences of radioactive liquid tank failures. In each of the processing systems considered, RATAF can calculate the tank isotopic concentrations in either the collector tank or the evaporator bottoms tank.

  20. Three radioactivity detectors for liquid-chromatographic systems compared

    SciTech Connect

    Frey, B.M.; Frey, F.J.

    1982-04-01

    Three radioactivity detectors coupled to a ''high-performance'' liquid-chromatography system are compared with regard to static efficiency, dynamic efficiency, background measurements, and within- and between-day variabilities. Their advantages and disadvantages are discussed.

  1. Radioactive Liquid Waste Treatment Facility: Environmental Information Document

    SciTech Connect

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

    1993-11-01

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

  2. Biodegradation of the alkaline cellulose degradation products generated during radioactive waste disposal.

    PubMed

    Rout, Simon P; Radford, Jessica; Laws, Andrew P; Sweeney, Francis; Elmekawy, Ahmed; Gillie, Lisa J; Humphreys, Paul N

    2014-01-01

    The anoxic, alkaline hydrolysis of cellulosic materials generates a range of cellulose degradation products (CDP) including α and β forms of isosaccharinic acid (ISA) and is expected to occur in radioactive waste disposal sites receiving intermediate level radioactive wastes. The generation of ISA's is of particular relevance to the disposal of these wastes since they are able to form complexes with radioelements such as Pu enhancing their migration. This study demonstrates that microbial communities present in near-surface anoxic sediments are able to degrade CDP including both forms of ISA via iron reduction, sulphate reduction and methanogenesis, without any prior exposure to these substrates. No significant difference (n = 6, p = 0.118) in α and β ISA degradation rates were seen under either iron reducing, sulphate reducing or methanogenic conditions, giving an overall mean degradation rate of 4.7 × 10(-2) hr(-1) (SE ± 2.9 × 10(-3)). These results suggest that a radioactive waste disposal site is likely to be colonised by organisms able to degrade CDP and associated ISA's during the construction and operational phase of the facility. PMID:25268118

  3. Biodegradation of the Alkaline Cellulose Degradation Products Generated during Radioactive Waste Disposal

    PubMed Central

    Rout, Simon P.; Radford, Jessica; Laws, Andrew P.; Sweeney, Francis; Elmekawy, Ahmed; Gillie, Lisa J.; Humphreys, Paul N.

    2014-01-01

    The anoxic, alkaline hydrolysis of cellulosic materials generates a range of cellulose degradation products (CDP) including α and β forms of isosaccharinic acid (ISA) and is expected to occur in radioactive waste disposal sites receiving intermediate level radioactive wastes. The generation of ISA's is of particular relevance to the disposal of these wastes since they are able to form complexes with radioelements such as Pu enhancing their migration. This study demonstrates that microbial communities present in near-surface anoxic sediments are able to degrade CDP including both forms of ISA via iron reduction, sulphate reduction and methanogenesis, without any prior exposure to these substrates. No significant difference (n = 6, p = 0.118) in α and β ISA degradation rates were seen under either iron reducing, sulphate reducing or methanogenic conditions, giving an overall mean degradation rate of 4.7×10−2 hr−1 (SE±2.9×10−3). These results suggest that a radioactive waste disposal site is likely to be colonised by organisms able to degrade CDP and associated ISA's during the construction and operational phase of the facility. PMID:25268118

  4. Membrane Treatment of Liquid Salt Bearing Radioactive Wastes

    SciTech Connect

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

    2003-02-25

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

  5. Process for encapsulating radioactive organic liquids in a resin

    SciTech Connect

    Drake, S.S.; Filter, H.E.

    1983-05-03

    Radioactive organic liquids are converted to a form suitable for burial by the process wherein the liquid is contacted with insoluble, swellable polymer particles to form swollen gelled particles which are dispersed in an unsaturated polyester, vinyl ester resin or mixture thereof which is then cured to a solid state with the gelled particles encased therein.

  6. Microbial reduction of U(VI) under alkaline conditions: implications for radioactive waste geodisposal.

    PubMed

    Williamson, Adam J; Morris, Katherine; Law, Gareth T W; Rizoulis, Athanasios; Charnock, John M; Lloyd, Jonathan R

    2014-11-18

    Although there is consensus that microorganisms significantly influence uranium speciation and mobility in the subsurface under circumneutral conditions, microbiologically mediated U(VI) redox cycling under alkaline conditions relevant to the geological disposal of cementitious intermediate level radioactive waste, remains unexplored. Here, we describe microcosm experiments that investigate the biogeochemical fate of U(VI) at pH 10-10.5, using sediments from a legacy lime working site, stimulated with an added electron donor, and incubated in the presence and absence of added Fe(III) as ferrihydrite. In systems without added Fe(III), partial U(VI) reduction occurred, forming a U(IV)-bearing non-uraninite phase which underwent reoxidation in the presence of air (O2) and to some extent nitrate. By contrast, in the presence of added Fe(III), U(VI) was first removed from solution by sorption to the Fe(III) mineral, followed by bioreduction and (bio)magnetite formation coupled to formation of a complex U(IV)-bearing phase with uraninite present, which also underwent air (O2) and partial nitrate reoxidation. 16S rRNA gene pyrosequencing showed that Gram-positive bacteria affiliated with the Firmicutes and Bacteroidetes dominated in the post-reduction sediments. These data provide the first insights into uranium biogeochemistry at high pH and have significant implications for the long-term fate of uranium in geological disposal in both engineered barrier systems and the alkaline, chemically disturbed geosphere. PMID:25231875

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

    SciTech Connect

    Osmanlioglu, Ahmet Erdal

    2012-07-01

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

  8. Process for decontaminating radioactive liquids using a calcium cyanamide-containing composition. [Patent application

    DOEpatents

    Silver, G.L.

    1980-09-24

    The present invention provides a process for decontaminating a radioactive liquid containing a radioactive element capable of forming a hydroxide. This process includes the steps of contacting the radioactive liquid with a decontaminating composition and separating the resulting radioactive sludge from the resulting liquid. The decontaminating composition contains calcium cyanamide.

  9. Alkaline extraction of polonium from liquid lead bismuth eutectic

    NASA Astrophysics Data System (ADS)

    Heinitz, S.; Neuhausen, J.; Schumann, D.

    2011-07-01

    The production of highly radiotoxic polonium isotopes poses serious safety concerns for the development of future nuclear systems cooled by lead bismuth eutectic (LBE). In this paper it is shown that polonium can be extracted efficiently from LBE using a mixture of alkaline metal hydroxides (NaOH + KOH) in a temperature range between 180 and 350 °C. The extraction ratio was analyzed for different temperatures, gas blankets and phase ratios. A strong dependence of the extraction performance on the redox properties of the cover gas was found. While hydrogen facilitates the removal of polonium, oxygen has a negative influence on the extraction. These findings open new possibilities to back up the safety of future LBE based nuclear facilities.

  10. Ultrasonic decontamination in perfluorinated liquids of radioactive circuit boards

    SciTech Connect

    Yam, C.S.; Harling, O.K.; Kaiser, R.

    1994-12-31

    A laboratory-scale ultrasonic decontamination system has been developed to demonstrate the application of Entropic System`s enhanced particle removal process to the radioactive decontamination of electronic circuit boards. The process uses inert perfluorinated liquids as the working media; the liquids have zero ozone depletion potential, are nontoxic, non-flammable, and are generally recognized as nonhazardous materials. The parts to be cleaned are first sonicated with a dilute solution of a high-molecular-weight fluorocarbon surfactant in an inert perfluorinated liquid. The combination of ultrasonic agitation and liquid flow promotes the detachment of the particles from the surface of the part being cleaned, their transfer from the boundary layer into the bulk liquid, and their removal from the cleaning environment, thereby reducing the probability of particle redeposition. After the cleaning process, the parts are rinsed with the pure perfluorinated liquid to remove residual surfactant. The parts are recovered after the perfluorinated liquid is evaporated into air.

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

    SciTech Connect

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

    1994-06-01

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

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

    SciTech Connect

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

    1999-10-07

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

  13. Position sensitive radioactivity detection for gas and liquid chromatography

    DOEpatents

    Cochran, Joseph L.; McCarthy, John F.; Palumbo, Anthony V.; Phelps, Tommy J.

    2001-01-01

    A method and apparatus are provided for the position sensitive detection of radioactivity in a fluid stream, particularly in the effluent fluid stream from a gas or liquid chromatographic instrument. The invention represents a significant advance in efficiency and cost reduction compared with current efforts.

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

    SciTech Connect

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

    1996-09-30

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

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

    SciTech Connect

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

    1998-06-01

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

  16. Liquids in multiorbital SU(N) magnets made up of ultracold alkaline-earth atoms

    NASA Astrophysics Data System (ADS)

    Xu, Cenke

    2010-04-01

    In this work we study one family of liquid states of k -orbital SU(N) spin systems, focusing on the case of k=2 which can be realized by ultracold alkaline-earth atoms trapped in optical lattices, with N as large as 10. Five different algebraic liquid states with selectively coupled charge, spin, and orbital quantum fluctuations are considered. The algebraic liquid states can be stabilized with large enough N and the scaling dimension of physical order parameters is calculated using a systematic 1/N expansion. The phase transitions between these liquid states are also studied and all the algebraic liquid states discussed in this work can be obtained from one “mother” state with SU(2)×U(1) gauge symmetry.

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

    SciTech Connect

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

    2011-10-01

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

  18. Selective separation of hydroxide from alkaline nuclear tank waste by liquid-liquid extraction with weak hydroxy acids.

    PubMed

    Chambliss, C Kevin; Haverlock, Tamara I; Bonnesen, Peter V; Engle, Nancy L; Moyer, Bruce A

    2002-04-15

    Recovery and recycle of caustic reagents in industrial processes offer potential means of pollution prevention, as investigated herein for particular needs related to the cleanup of alkaline nuclear waste. Specifically, the recovery of hydroxide from alkaline media by liquid-liquid extraction can be effected utilizing weak hydroxy acids, as demonstrated for NaOH utilizing a series of lipophilic fluorinated alcohols and alkylated phenols dissolved in 1-octanol. Extraction efficiency follows the expected order of acidity of the hydroxy acids, the phenols being the most efficient extractants among the compounds tested. After extraction, NaOH is effectively recoverable from the organic phase upon contact with water. The weakest hydroxy acids are the most efficiently stripped, NaOH recovery being nearly quantitative in a single contact. In competitive extraction experiments, good selectivity for hydroxide recovery over other anions such as nitrate and chloride was demonstrated. Since the order of extraction favors larger anions, the exceptional preference for hydroxide implies that the extraction occurs by deprotonation of the hydroxy acids in a cation-exchange process. Stripping therefore occurs by hydrolysis to regenerate the neutral hydroxy acid, liberating NaOH to the aqueous phase. Since hydroxide equivalents rather than actual hydroxide ions are transferred to the solvent, the process is termed "pseudohydroxide extraction." Hydroxide recovery from a simulant of alkaline nuclear tank waste (Hanford DSSF simulant) was also demonstrated in repeated extraction and stripping cycles. PMID:11993889

  19. Radioactive Liquid Waste Treatment Facility Discharges in 2011

    SciTech Connect

    Del Signore, John C.

    2012-05-16

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

  20. Interrelationship of Kaolin, Alkaline Liquid Ratio and Strength of Kaolin Geopolymer

    NASA Astrophysics Data System (ADS)

    Ramasamy, Shamala; Hussin, Kamarudin; Bakri Abdullah, Mohd Mustafa Al; Mohd Ruzaidi Ghazali, Che; Binhussain, Mohammed; Sandu, Andrei Victor

    2016-06-01

    Geopolymer is an incredible alternative green cementitious material which has ceramic-like properties, but does not require calcining that leads to reduction in processing energy usage. The purpose of this research is to study the correlation between kaolin: liquid ratio with the performance of kaolin geopolymer. Kaolin, a prominent raw geopolymer material was used to prepare enhanced geopolymer paste by mixing with alkaline activator solution. Interrelationship of kaolin to alkaline liquid ratio with hardness and flexural strength was the focus of this work. Therefore kaolin geopolymer paste with varying solid to liquid ratio ranging from 0.7 to 1.1 was prepared. Geopolymer paste was coated on low grade wood substrate prior to Vickers hardness and flexural strength. X-ray diffraction was conducted on geopolymer paste itself after 7 days to analyze the change in phase identification at early age. Kaolin geopolymer coating on wood with solid/liquid(S/L) ratio of 0.7 shows the most promising hardness and flexural strength of 15.3 Hv and 94.73MPa. X-ray diffraction test showed high existence of kaolinite on higher S/L ratio where as sodalite was observed in S/L ratio of 0.7. Microstructural studies also compliments our finding which further proves the positive dependency of S/L ratio and kaolin geopolymer strength.

  1. Clinopyroxene-liquid thermometers and barometers specific to alkaline differentiated magmas

    NASA Astrophysics Data System (ADS)

    Masotta, M.; Mollo, S.; Freda, C.; Gaeta, M.; Moore, G.

    2013-12-01

    We present new thermometers and barometers based on clinopyroxene-liquid equilibria specific to alkaline differentiated magmas. The new models were calibrated through the regression analyses of experimental datasets obtained by merging phase equilibria experiments from the literature with new experiments performed by using trachytic and phonolitic starting compositions. The regression strategy was twofold: (1) we have tested previous thermometric and barometric equations and recalibrated these models using the new datasets; (2) we have calibrated a new thermometer and a new barometer including only regression parameters that closely describe the compositional variability of the datasets. The new models yield more precise estimates than previous thermometers and barometers when used to predict temperatures and pressures of alkaline differentiated magmas. We have tested the reliability of the new equations by using clinopyroxene-liquid pairs from trachytes and phonolites erupted during major explosive eruptions at the Phlegrean Fields and Mt. Vesuvius (central Italy). The test yielded crystallization conditions comparable to those determined by means of melt and fluid inclusion analyses and phase equilibria studies; this validates the use of the proposed models for precise estimates of crystallization temperatures and pressures in differentiated alkaline magmas. Because these magmas feed some of the most voluminous, explosive, and threatening volcanic eruptions in the world, a better understanding of the environmental conditions of their reservoirs is mandatory and this is now possible with the new models provided here.

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

    PubMed

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

    2009-01-30

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

  3. [A system for decontamination of liquid radioactive waste produced in in vitro tests in nuclear medicine].

    PubMed

    Tsuchiya, T; Norimura, T; Ueno, T

    1983-06-01

    It is well known that very large storage tanks for radioactive liquids are necessary for the disposal of liquid radioactive waste. In vitro tests in radioimmunoassay in nuclear medicine are rapidly increasing for clinical examination causing marked increase in the volume of liquid radioactive waste. Thus we have developed a system for decontaminating radioactivity from liquid waste. In the first step, the liquid waste is boiled by a sterilizer and, in the second step, this sterilised liquid is filtered by a cylindrical filter (Toyo filter No. 84). After filtration, the liquid waste is passed into a beaded charcoal column and an ion exchange resin (Amberlite IRA 402) column. After these treatments, the radioactivity level of liquid waste is lowered to less than 1% of the original radioactivity. We are now in the planning stages of building an apparatus for practical use. PMID:6622764

  4. Decontamination of a radioactive waste liquid by electrodialysis

    SciTech Connect

    Lundstrom, J.E.

    1987-02-24

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

  5. Application of solvlent change techniques to blended cements used to immobilize low-level radioactive liquid waste

    SciTech Connect

    Kruger, A.A.

    1996-07-01

    The microstructures of hardened portland and blended cement pastes, including those being considered for use in immobilizing hazardous wastes, have a complex pore structure that changes with time. In solvent exchange, the pore structure is examined by immersing a saturated sample in a large volume of solvent that is miscible with the pore fluid. This paper reports the results of solvent replacement measurements on several blended cements mixed at a solution:solids ratio of 1.0 with alkaline solutions from the simulation of the off- gas treatment system in a vitrification facility treating low-level radioactive liquid wastes. The results show that these samples have a lower permeability than ordinary portland cement samples mixed at a water:solids ratio of 0.70, despite having a higher volume of porosity. The microstructure is changed by these alkaline solutions, and these changes have important consequences with regard to durability.

  6. DEVELOPMENT AND SELECTION OF IONIC LIQUID ELECTROLYTES FOR HYDROXIDE CONDUCTING POLYBENZIMIDAZOLE MEMBRANES IN ALKALINE FUEL CELLS

    SciTech Connect

    Fox, E.

    2012-05-01

    Alkaline fuel cell (AFC) operation is currently limited to specialty applications such as low temperatures and pure HO due to the corrosive nature of the electrolyte and formation of carbonates. AFCs are the cheapest and potentially most efficient (approaching 70%) fuel cells. The fact that non-Pt catalysts can be used, makes them an ideal low cost alternative for power production. The anode and cathode are separated by and solid electrolyte or alkaline porous media saturated with KOH. However, CO from the atmosphere or fuel feed severely poisons the electrolyte by forming insoluble carbonates. The corrosivity of KOH (electrolyte) limits operating temperatures to no more than 80°C. This chapter examines the development of ionic liquids electrolytes that are less corrosive, have higher operating temperatures, do not chemically bond to CO and enable alternative fuels. Work is detailed on the IL selection and characterization as well as casting methods within the polybenzimidazole based solid membrane. This approach is novel as it targets the root of the problem (the electrolyte) unlike other current work in alkaline fuel cells which focus on making the fuel cell components more durable.

  7. Disposal of liquid radioactive wastes through wells or shafts

    NASA Astrophysics Data System (ADS)

    Perkins, B. L.

    1982-01-01

    Liquids and, in some cases, suitable solids and/or entrapped gases can be disposed of by: (1) well injection into deep permeable strata, bounded by impermeable layers; (2) grout injection into an impermeable host rock, forming fractures in which the waste solidifies; and (3) slurrying into excavated subsurface cavities. Radioactive materials are presently being disposed of worldwide using all three techniques; however, it would appear that if the techniques were verified as posing minimum hazards to the environment and suitable site-specific host rock were identified, these disposal techniques could be more widely used.

  8. Disposal of liquid radioactive wastes through wells or shafts

    SciTech Connect

    Perkins, B.L.

    1982-01-01

    This report describes disposal of liquids and, in some cases, suitable solids and/or entrapped gases, through: (1) well injection into deep permeable strata, bounded by impermeable layers; (2) grout injection into an impermeable host rock, forming fractures in which the waste solidifies; and (3) slurrying into excavated subsurface cavities. Radioactive materials are presently being disposed of worldwide using all three techniques. However, it would appear that if the techniques were verified as posing minimum hazards to the environment and suitable site-specific host rock were identified, these disposal techniques could be more widely used.

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

    SciTech Connect

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

    2015-04-29

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

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

    SciTech Connect

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

    2012-07-01

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

  11. Modified microspheres for cleaning liquid wastes from radioactive nuclides

    SciTech Connect

    Danilin, Lev; Drozhzhin, Valery

    2007-07-01

    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)

  12. Early containment of high-alkaline solution simulating low-level radioactive waste stream in clay-bearing blended cement

    SciTech Connect

    Kruger, A.A.; Olson, R.A.; Tennis, P.D.

    1995-04-01

    Portland cement blended with fly ash and attapulgite clay was mixed with high-alkaline solution simulating low-level radioactive waste stream at a one-to-one weight ratio. Mixtures were adiabatically and isothermally cured at various temperatures and analyzed for phase composition, total alkalinity, pore solution chemistry, and transport properties as measured by impedance spectroscopy. Total alkalinity is characterized by two main drops. The early one corresponds to a rapid removal of phosphorous, aluminum, sodium, and to a lesser extent potassium solution. The second drop from about 10 h to 3 days is mainly associated with the removal of aluminum, silicon, and sodium. Thereafter, the total alkalinity continues descending, but at a lower rate. All pastes display a rapid flow loss that is attributed to an early precipitation of hydrated products. Hemicarbonate appears as early as one hour after mixing and is probably followed by apatite precipitation. However, the former is unstable and decomposes at a rate that is inversely related to the curing temperature. At high temperatures, zeolite appears at about 10 h after mixing. At 30 days, the stabilized crystalline composition Includes zeolite, apatite and other minor amounts of CaCO{sub 3}, quartz, and monosulfate Impedance spectra conform with the chemical and mineralogical data. The normalized conductivity of the pastes shows an early drop, which is followed by a main decrease from about 12 h to three days. At three days, the permeability of the cement-based waste as calculated by Katz-Thompson equation is over three orders of magnitude lower than that of ordinary portland cement paste. However, a further decrease in the calculated permeability is questionable. Chemical stabilization is favorable through incorporation of waste species into apatite and zeolite.

  13. Hydrogen gettering the overpressure gas from highly radioactive liquids

    SciTech Connect

    Riley, D.L.; McCoy, J.C.; 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.

  14. Comparison of liquid hot water and alkaline pretreatments of giant reed for improved enzymatic digestibility and biogas energy production.

    PubMed

    Jiang, Danping; Ge, Xumeng; Zhang, Quanguo; Li, Yebo

    2016-09-01

    Liquid hot water (LHW) and alkaline pretreatments of giant reed biomass were compared in terms of digestibility, methane production, and cost-benefit efficiency for electricity generation via anaerobic digestion with a combined heat and power system. Compared to LHW pretreatment, alkaline pretreatment retained more of the dry matter in giant reed biomass solids due to less severe conditions. Under their optimal conditions, LHW pretreatment (190°C, 15min) and alkaline pretreatment (20g/L of NaOH, 24h) improved glucose yield from giant reed by more than 2-fold, while only the alkaline pretreatment significantly (p<0.05) increased cumulative methane yield (by 63%) over that of untreated biomass (217L/kgVS). LHW pretreatment obtained negative net electrical energy production due to high energy input. Alkaline pretreatment achieved 27% higher net electrical energy production than that of non-pretreatment (3859kJ/kg initial total solids), but alkaline liquor reuse is needed for improved net benefit. PMID:27233098

  15. RECENT STUDIES OF URANIUM AND PLUTONIUM CHEMISTRY IN ALKALINE RADIOACTIVE WASTE SOLUTIONS

    SciTech Connect

    King, W; Bill Wilmarth, B; David Hobbs, D; Tommy Edwards, T

    2006-06-13

    Solubility studies of uranium and plutonium in a caustic, radioactive Savannah River Site tank waste solution revealed the existence of uranium supersaturation in the as-received sample. Comparison of the results to predictions generated from previously published models for solubility in these waste types revealed that the U model poorly predicts solubility while Pu model predictions are quite consistent with experimental observations. Separate studies using simulated Savannah River Site evaporator feed solution revealed that the known formation of sodium aluminosilicate solids in waste evaporators can promote rapid precipitation of uranium from supersaturated solutions.

  16. Conditioning of Boron-Containing Low and Intermediate Level Liquid Radioactive Waste - 12041

    SciTech Connect

    Gorbunova, Olga A.; Kamaeva, Tatiana S.

    2012-07-01

    Improved cementation of low and intermediate level radioactive waste (ILW and LLW) aided by vortex electromagnetic treatment as well as silica addition was investigated. Positive effects including accelerated curing of boron-containing cement waste forms, improve end product quality, decreased product volume and reduced secondary LRW volume from equipment decontamination were established. These results established the possibility of boron-containing LRW cementation without the use of neutralizing alkaline additives that greatly increase the volume of the final product intended for long-term storage (burial). Physical (electromagnetic) treatment in a vortex mixer can change the state of LRW versus chemical treatment. By treating the liquid phase of cement solution only, instead of the whole solution, and using fine powder and nano-particles of ferric oxides instead of separable ferromagnetic cores for the activating agents the positive effect are obtained. VET for 1 to 3 minutes yields boron-containing LRW cemented products of satisfactory quality. Silica addition at 10 % by weight will accelerate curing and solidification and to decrease radionuclide leaching rates from boron-containing cement products. (authors)

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

    SciTech Connect

    Johnson, J.D.; Whitley, C.R.; Rawool-Sullivan, M.

    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.

  18. Incineration of radioactive organic liquid wastes by underwater thermal plasma

    NASA Astrophysics Data System (ADS)

    Mabrouk, M.; Lemont, F.; Baronnet, J. M.

    2012-12-01

    This work deals with incineration of radioactive organic liquid wastes using an oxygen thermal plasma jet, submerged under water. The results presented here are focused on incineration of three different wastes: a mixture of tributylphosphate (TBP) and dodecane, a perfluoropolyether oil (PFPE) and trichloroethylene (TCE). To evaluate the plutonium behavior in used TBP/dodecane incineration, zirconium is used as a surrogate of plutonium; the method to enrich TBP/dodecane mixture in zirconium is detailed. Experimental set-up is described. During a trial run, CO2 and CO contents in the exhaust gas are continuously measured; samples, periodically taken from the solution, are analyzed by appropriate chemical methods: contents in total organic carbon (COT), phosphorus, fluoride and nitrates are measured. Condensed residues are characterized by RX diffraction and SEM with EDS. Process efficiency, during tests with a few L/h of separated or mixed wastes, is given by mineralization rate which is better than 99.9 % for feed rate up to 4 L/h. Trapping rate is also better than 99 % for phosphorous as for fluorine and chlorine. Those trials, with long duration, have shown that there is no corrosion problems, also the hydrogen chloride and fluoride have been neutralized by an aqueous solution of potassium carbonate.

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

    SciTech Connect

    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.

  20. Pilot studies to achieve waste minimization and enhance radioactive liquid waste treatment at the Los Alamos National Laboratory Radioactive Liquid Waste Treatment Facility

    SciTech Connect

    Freer, J.; Freer, E.; Bond, A.

    1996-07-01

    The Radioactive and Industrial Wastewater Science Group manages and operates the Radioactive Liquid Waste Treatment Facility (RLWTF) at the Los Alamos National Laboratory (LANL). The RLWTF treats low-level radioactive liquid waste generated by research and analytical facilities at approximately 35 technical areas throughout the 43-square-mile site. The RLWTF treats an average of 5.8 million gallons (21.8-million liters) of liquid waste annually. Clarifloculation and filtration is the primary treatment technology used by the RLWTF. This technology has been used since the RLWTF became operable in 1963. Last year the RLWTF achieved an average of 99.7% removal of gross alpha activity in the waste stream. The treatment process requires the addition of chemicals for the flocculation and subsequent precipitation of radionuclides. The resultant sludge generated during this process is solidified in drums and stored or disposed of at LANL.

  1. Biological Information Document, Radioactive Liquid Waste Treatment Facility

    SciTech Connect

    Biggs, J.

    1995-12-31

    This document is intended to act as a baseline source material for risk assessments which can be used in Environmental Assessments and Environmental Impact Statements. The current Radioactive Liquid Waste Treatment Facility (RLWTF) does not meet current General Design Criteria for Non-reactor Nuclear Facilities and could be shut down affecting several DOE programs. This Biological Information Document summarizes various biological studies that have been conducted in the vicinity of new Proposed RLWTF site and an Alternative site. The Proposed site is located on Mesita del Buey, a mess top, and the Alternative site is located in Mortandad Canyon. The Proposed Site is devoid of overstory species due to previous disturbance and is dominated by a mixture of grasses, forbs, and scattered low-growing shrubs. Vegetation immediately adjacent to the site is a pinyon-juniper woodland. The Mortandad canyon bottom overstory is dominated by ponderosa pine, willow, and rush. The south-facing slope was dominated by ponderosa pine, mountain mahogany, oak, and muhly. The north-facing slope is dominated by Douglas fir, ponderosa pine, and oak. Studies on wildlife species are limited in the vicinity of the proposed project and further studies will be necessary to accurately identify wildlife populations and to what extent they utilize the project area. Some information is provided on invertebrates, amphibians and reptiles, and small mammals. Additional species information from other nearby locations is discussed in detail. Habitat requirements exist in the project area for one federally threatened wildlife species, the peregrine falcon, and one federal candidate species, the spotted bat. However, based on surveys outside of the project area but in similar habitats, these species are not expected to occur in either the Proposed or Alternative RLWTF sites. Habitat Evaluation Procedures were used to evaluate ecological functioning in the project area.

  2. Coil spring conveyor for positioning an external radioactive standard in a liquid scintillation counter

    SciTech Connect

    Kampf, R.S.

    1986-03-04

    This patent describes a liquid scintillation counter having a counting chamber for receiving a liquid sample containing a liquid scintillator and a sample of a radioactive substance to be counted. An improved apparatus is described for positioning a radioactive source in an operating location to irradiate the liquid sample in the counting chamber comprising the combination of: (1) a flexible conveyor including an elongate, generally tubular coil spring section having an interior tubular bore configured to receive the radioactive source; (2) means for retaining the radioactive source in a predetermined axial position within the coil spring; (3) means supporting the coil spring for movement along a conveyor path between a storage location for the radioactive source remote from the counting chamber and an operating location for the radioactive source proximate the counting chamber; and (4) drive means coupled to the coil spring and operative to drive the coil spring along the conveyor path for conveying the radioactive source between the storage and operating locations.

  3. Calculation of chemical quantities for the radioactive liquid waste treatment facility

    SciTech Connect

    Del Signore, John C.; McClenahan, Robert L.

    2007-03-01

    The Radioactive Liquid Waste Treatment Facility (RLWTF) receives, stores, and treats both low-level and transuranic radioactive liquid wastes (RLW). Treatment of RLW requires the use of different chemicals. Examples include the use of calcium oxide to precipitate metals and radioactive elements from the radioactive liquid waste, and the use of hydrochloric acid to clean membrane filters that are used in the treatment process. The RL WTF is a Hazard Category 2 nuclear facility, as set forth in the LANL Final Safety Analysis Report of October 1995, and a DOE letter of March 11, 1999. A revised safety basis is being prepared for the RLWTF, and will be submitted to the NNSA in early 2007. This set of calculations establishes maximum chemical quantities that will be used in the 2007 safety basis.

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

    SciTech Connect

    Christensen, B. D.

    1999-02-15

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

  5. Industrial Technology of Decontamination of Liquid Radioactive Waste in SUE MosSIA 'Radon' - 12371

    SciTech Connect

    Adamovich, Dmitry V.; Neveykin, Petr P.; Karlin, Yuri V.; Savkin, Alexander E.

    2012-07-01

    SUE MosSIA 'RADON' - this enterprise was created more than 50 years ago, which deals with the recycling of radioactive waste and conditioning of spent sources of radiation in stationary and mobile systems in the own factory and operating organizations. Here is represented the experience SUE MosSIA 'Radon' in the field of the management with liquid radioactive waste. It's shown, that the activity of SUE MosSIA 'RADON' is developing in three directions - improvement of technical facilities for treatment of radioactive waters into SUE MosSIA 'RADON' development of mobile equipment for the decontamination of radioactive waters in other organizations, development of new technologies for decontamination of liquid radioactive wastes as part of various domestic Russian and international projects including those related to the operation of nuclear power and nuclear submarines. SUE MosSIA 'RADON' has processed more than 270 thousand m{sup 3} of radioactive water, at that more than 7000 m{sup 3} in other organizations for more than 50 years. It is shown that a number of directions, particularly, the development of mobile modular units for decontamination of liquid radioactive waste, SUE MosSIA 'RADON' is a leader in the world. (authors)

  6. Process for immobilizing radioactive boric acid liquid wastes

    SciTech Connect

    Greenhalgh, W.O.

    1986-06-17

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

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

    SciTech Connect

    White, T.L.

    1995-03-31

    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.

  8. Detection of free liquid in containers of solidified radioactive waste

    DOEpatents

    Greenhalgh, W.O.

    Nondestructive detection of the presence of free liquid within a sealed enclosure containing solidified waste is accomplished by measuring the levels of waste at two diametrically opposite locations while slowly tilting the enclosure toward one of said locations. When the measured level remains constant at the other location, the measured level at said one location is noted and any measured difference of levels indicates the presence of liquid on the surface of the solifified waste. The absence of liquid in the enclosure is verified when the measured levels at both locations are equal.

  9. Detection of free liquid in containers of solidified radioactive waste

    DOEpatents

    Greenhalgh, Wilbur O.

    1985-01-01

    A method of nondestructively detecting the presence of free liquid within a sealed enclosure containing solidified waste by measuring the levels of waste at two diametrically opposite locations while slowly tilting the enclosure toward one of said locations. When the measured level remains constant at the other location, the measured level at said one location is noted and any measured difference of levels indicates the presence of liquid on the surface of the solidified waste. The absence of liquid in the enclosure is verified when the measured levels at both locations are equal.

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

    DOEpatents

    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.

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

    DOEpatents

    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.

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

    SciTech Connect

    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.

  13. Process for immobilizing radioactive boric acid liquid wastes

    SciTech Connect

    Greenhalgh, Wilbur O.

    1986-01-01

    A method of immobilizing boric acid liquid wastes containing radionuclides by neutralizing the solution and evaporating the resulting precipitate to near dryness. The dry residue is then fused into a reduced volume, insoluble, inert, solid form containing substantially all the radionuclides.

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

    SciTech Connect

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

    1994-05-01

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

  15. Apparatus for positioning an external radioactive standard in a liquid scintillation counter

    SciTech Connect

    Horrocks, D.L.; Kampf, R.S.

    1987-07-07

    This patent describes a liquid scintillation counter having a counting chamber for receiving a sample containing a scintillator substance and a sample of a radioactive substance to be counted. The improved apparatus positions a radioactive source in an operating location to irradiate the sample in the counting chamber comprising, in combination: (1) a continuous bidirectionally flexible conveyor forming a closed loop for conveying the radioactive source through on operating location and a storage location; (2) means supporting the radioactive source at a position along the flexible conveyor for conveyance; (3) guide means for supporting the conveyor and for guiding conveyor movement along a selected path, the path transversing at spaced positions the storage location for the radioactive source remote from the counting chamber and the operating location for the radioactive source near to the counting chamber; and (4) drive means coupled to the continuous flexible conveyor to draw the conveyor around the path for conveying the radioactive source through the spaced storage and operating locations.

  16. Treatment of radioactive liquid waste by sorption on natural zeolite in Turkey.

    PubMed

    Osmanlioglu, Ahmet Erdal

    2006-09-01

    Liquid radioactive waste has been generated from the use of radioactive materials in industrial applications, research and medicine in Turkey. Natural zeolites (clinoptilolite) have been studied for the removal of several key radionuclides ((137)Cs, (60)Co, (90)Sr and (110m)Ag) from liquid radioactive waste. The aim of the present study is to investigate effectiveness of zeolite treatment on decontamination factor (DF) in a combined process (chemical precipitation and adsorption) at the laboratory tests and scale up to the waste treatment plant. In this study, sorption and precipitation techniques were adapted to decontamination of liquid low level waste (LLW). Effective decontamination was achieved when sorbents are used during the chemical precipitation. Natural zeolite samples were taken from different zeolite formations in Turkey. Comparison of the ion-exchange properties of zeolite minerals from different formations shows that Gordes clinoptilolite was the most suitable natural sorbent for radionuclides under dynamic treatment conditions and as an additive for chemical precipitation process. Clinoptilolite were shown to have a high selectivity for (137)Cs and (110m)Ag as sorbent. In the absence of potassium ions, native clinoptilolite removed (60)Co and (90)Sr very effectively from the liquid waste. In the end of this liquid waste treatment, decontamination factor was provided as 430 by using 0.5 mm clinoptilolite at 30 degrees C. PMID:16563616

  17. US and Russian innovative technologies to process low-level liquid radioactive wastes: The Murmansk initiative

    SciTech Connect

    Dyer, R.S.; Penzin, R.; Duffey, R.B.; Sorlie, A.

    1996-12-31

    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.

  18. Equipment experience in a radioactive LFCM (liquid-fed ceramic melter) vitrification facility

    SciTech Connect

    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.

  19. Liquid radioactive waste discharges from B plant to cribs

    SciTech Connect

    Williams, J.C., Westinghouse Hanford

    1996-05-29

    This engineering report compiles information on types and quantities of liquid waste discharged from B-Plant directly to cribs, ditches, reverse wells, etc., that are associated with B-Plant. Waste discharges to these cribs via overflow form 241-B, 241-BX, and 241-BY tank farms, and waste discharged to these cribs from sources other than B-Plant are discussed.Discharges from B-Plant to other cribs, unplanned releases, or waste remaining in tanks are not included in the report. Waste stream composition information is used to predict quantities of individual chemicals sent to cribs. This provides an accurate mass balance of waste streams from B-Plant to these cribs. These predictions are compared with known crib inventories as a verification of the process.

  20. Audit of the radioactive liquid waste treatment facility operations at the Los Alamos National Laboratory

    SciTech Connect

    1997-11-19

    Los Alamos National Laboratory (Los Alamos) generates radioactive and liquid wastes that must be treated before being discharged to the environment. Presently, the liquid wastes are treated in the Radioactive Liquid Waste Treatment Facility (Treatment Facility), which is over 30 years old and in need of repair or replacement. However, there are various ways to satisfy the treatment need. The objective of the audit was to determine whether Los Alamos cost effectively managed its Treatment Facility operations. The audit determined that Los Alamos` treatment costs were significantly higher when compared to similar costs incurred by the private sector. This situation occurred because Los Alamos did not perform a complete analysis of privatization or prepare a {open_quotes}make-or-buy{close_quotes} plan for its treatment operations, although a {open_quotes}make-or-buy{close_quotes} plan requirement was incorporated into the contract in 1996. As a result, Los Alamos may be spending $2.15 million more than necessary each year and could needlessly spend $10.75 million over the next five years to treat its radioactive liquid waste. In addition, Los Alamos has proposed to spend $13 million for a new treatment facility that may not be needed if privatization proves to be a cost effective alternative. We recommended that the Manager, Albuquerque Operations Office (Albuquerque), (1) require Los Alamos to prepare a {open_quotes}make-or-buy{close_quotes} plan for its radioactive liquid waste treatment operations, (2) review the plan for approval, and (3) direct Los Alamos to select the most cost effective method of operations while also considering other factors such as mission support, reliability, and long-term program needs. Albuquerque concurred with the recommendations.

  1. Magnetic crystals and helical liquids in alkaline-earth fermionic gases.

    PubMed

    Barbarino, Simone; Taddia, Luca; Rossini, Davide; Mazza, Leonardo; Fazio, Rosario

    2015-01-01

    The joint action of a magnetic field and of interactions is crucial for the appearance of exotic quantum phenomena, such as the quantum Hall effect. Owing to their rich nuclear structure, equivalent to an additional synthetic dimension, one-dimensional alkaline-earth(-like) fermionic gases with synthetic gauge potential and atomic contact repulsion may display similar related properties. Here we show the existence and the features of a hierarchy of fractional insulating and conducting states by means of analytical and numerical methods. We demonstrate that the gapped states are characterized by density and magnetic order emerging solely for gases with effective nuclear spin larger than 1/2, whereas the gapless phases can support helical modes. We finally argue that these states are related to an unconventional fractional quantum Hall effect in the thin-torus limit and that their properties can be studied in state-of-the-art laboratories. PMID:26350624

  2. Magnetic crystals and helical liquids in alkaline-earth fermionic gases

    PubMed Central

    Barbarino, Simone; Taddia, Luca; Rossini, Davide; Mazza, Leonardo; Fazio, Rosario

    2015-01-01

    The joint action of a magnetic field and of interactions is crucial for the appearance of exotic quantum phenomena, such as the quantum Hall effect. Owing to their rich nuclear structure, equivalent to an additional synthetic dimension, one-dimensional alkaline-earth(-like) fermionic gases with synthetic gauge potential and atomic contact repulsion may display similar related properties. Here we show the existence and the features of a hierarchy of fractional insulating and conducting states by means of analytical and numerical methods. We demonstrate that the gapped states are characterized by density and magnetic order emerging solely for gases with effective nuclear spin larger than 1/2, whereas the gapless phases can support helical modes. We finally argue that these states are related to an unconventional fractional quantum Hall effect in the thin-torus limit and that their properties can be studied in state-of-the-art laboratories. PMID:26350624

  3. Optimization of screening for radioactivity in urine by liquid scintillation.

    SciTech Connect

    Shanks, Sonoya Toyoko; Reese, Robert P.; Preston, Rose T.

    2007-08-01

    Numerous events have or could have resulted in the inadvertent uptake of radionuclides by fairly large populations. Should a population receive an uptake, valuable information could be obtained by using liquid scintillation counting (LSC) techniques to quickly screen urine from a sample of the affected population. This study investigates such LSC parameters as discrimination, quench, volume, and count time to yield guidelines for analyzing urine in an emergency situation. Through analyzing variations of the volume and their relationships to the minimum detectable activity (MDA), the optimum ratio of sample size to scintillating chemical cocktail was found to be 1:3. Using this optimum volume size, the alpha MDA varied from 2100 pCi/L for a 30-second count time to 35 pCi/L for a 1000-minute count time. The typical count time used by the Sandia National Laboratories Radiation Protection Sample Diagnostics program is 30 minutes, which yields an alpha MDA of 200 pCi/L. Because MDA is inversely proportional to the square root of the count time, count time can be reduced in an emergency situation to achieve the desired MDA or response time. Note that approximately 25% of the response time is used to prepare the samples and complete the associated paperwork. It was also found that if the nuclide of interest is an unknown, pregenerated discriminator settings and efficiency calibrations can be used to produce an activity value within a factor of two, which is acceptable for a screening method.

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

    SciTech Connect

    Kundari, Noor Anis Putra, Sugili; Mukaromah, Umi

    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

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

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

    2013-07-01

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

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

    SciTech Connect

    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.

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

    SciTech Connect

    Bessa Nisti, Marcelo; Godoy dos Santos, Adir Janete

    2008-08-07

    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{sup 8} Bq, 5.7x10{sup 8} Bq and 2.7xl0{sup 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.

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

    NASA Astrophysics Data System (ADS)

    Nisti, Marcelo Bessa; dos Santos, Adir Janete Godoy

    2008-08-01

    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 8.5×l08 Bq, 5.7×108 Bq and 2.7×l08 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.

  10. Compositions of magmas and carbonate silicate liquid immiscibility in the Vulture alkaline igneous complex, Italy

    NASA Astrophysics Data System (ADS)

    Solovova, I. P.; Girnis, A. V.; Kogarko, L. N.; Kononkova, N. N.; Stoppa, F.; Rosatelli, G.

    2005-11-01

    This paper presents a study of melt and fluid inclusions in minerals of an olivine-leucite phonolitic nephelinite bomb from the Monticchio Lake Formation, Vulture. The rock contains 50 vol.% clinopyroxene, 12% leucite, 10% alkali feldspars, 8% hauyne/sodalite, 7.5% nepheline, 4.5% apatite, 3.2% olivine, 2% opaques, 2.6% plagioclase, and < 1% amphibole. We distinguished three generations of clinopyroxene differing in composition and morphology. All the phenocrysts bear primary and secondary melt and fluid inclusions, which recorded successive stages of melt evolution. The most primitive melts were found in the most magnesian olivine and the earliest clinopyroxene phenocrysts. The melts are near primary mantle liquids and are rich in Ca, Mg and incompatible and volatile elements. Thermometric experiments with the melt inclusions suggested that melt crystallization began at temperatures of about 1200 °C. Because of the partial leakage of all primary fluid inclusions, the pressure of crystallization is constrained only to minimum of 3.5 kbar. Combined silicate-carbonate melt inclusions were found in apatite phenocrysts. They are indicative of carbonate-silicate liquid immiscibility, which occurred during magma evolution. Large hydrous secondary melt inclusions were found in olivine and clinopyroxene. The inclusions in the phenocrysts recorded an open-system magma evolution during its rise towards the surface including crystallization, degassing, oxidation, and liquid immiscibility processes.

  11. SOLIEX: A Novel Solid-Liquid Method of Radionuclides Extraction from Radioactive Waste Solutions - 13486

    SciTech Connect

    Shilova, E.; Viel, P.; Huc, V.

    2013-07-01

    This paper describes recent developments in new solid-liquid extraction method, called SOLIEX, to remove cesium from alkaline solutions. SOLIEX relies on the use of a reversible complexing system comprising a carbon felt bearing molecular traps (calixarenes). This complexing system exhibits a high selectivity for Cs, and is thus expected to be helpful for the treatment of highly diluted cesium wastes even with a high concentration of competing alkali metal cations. As additional advantage, this complexing system can be adapted by molecular engineering to capture other radionuclides, such as Sr, Eu, Am. Finally, this complexing system can be easily and efficiently regenerated by using a cost effective stripping procedure, which limits further generation of waste to meet 'zero liquid' discharge requirements for nuclear facilities. (authors)

  12. Liquid scintillation counting of polycarbonates: a sensitive technique for measurement of activity concentration of some radioactive noble gases.

    PubMed

    Mitev, K; Zhivkova, V; Pressyanov, D; Georgiev, S; Dimitrova, I; Gerganov, G; Boshkova, T

    2014-11-01

    This work explores the application of the liquid scintillation counting of polycarbonates for measurement of the activity concentration of radioactive noble gases. Results from experimental studies of the method are presented. Potential applications in the monitoring of radioactive noble gases are discussed. PMID:24559941

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

    DOEpatents

    Barney, Gary S.; Brownell, Lloyd E.

    1977-01-01

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

  14. [Analysis of alkaline CuO degradation products of acid detergent fiber from tobacco leaves by using liquid chromatography].

    PubMed

    Hao, Weiqiang; Wang, Leijun; Wu, Shun; Yue, Bangyi; Chen, Qiang; Zhang, Peipei

    2015-07-01

    The acid detergent fiber (ADF) from tobacco leaves was obtained by treating the sample with petroleum ether-ethanol (6:4, v/v), 30 g/L sodium dodecylsulfate and 0.5 mol/L sulphuric acid containing 20 g/L hexadecyl trimethyl ammonium bromide successively. The ADF was degraded by the alkaline CuO oxidation procedure. In this work, six samples of ADF degradation products from tobacco leaves were prepared. The samples were analyzed by using gradient liquid chromatography (LC) where an Ultimate XB C18 column was used as stationary phase, with a mixture of methanol and water as mobile phase, at a column temperature of 35 °C and a flow rate of 0.8 mL/min. Dual wavelengths of 280 nm and 320 nm were chosen for the detection. It was found that there were four characteristic peaks for the ADF degradation products. By taking these peaks as research object, the optimum time for the degradation was found to be 5 h and the sample solution could be kept stable within 7 days. The established method may provide a new approach for the studies of the differences between lignin composition in different tobacco leaves and the relationship between lignin content and the smoking quality of tobacco leaves. PMID:26672209

  15. Kinetic study on the degradation of meclophenoxate hydrochloride in alkaline aqueous solutions by high performance liquid chromatography.

    PubMed

    El-Bardicy, Mohammad Galal; Lotfy, Hayam Mahmoud; El-Sayed, Mohammad Abdalla; El-Tarras, Mohammad Fayez

    2007-01-01

    A high performance liquid chromatographic method was developed and validated for determination of meclophenoxate hydrochloride (I) in the presence of its degradation product (p-chlorophenoxy acetic acid) (II). Separation of (I) from (II) was performed using a ZORBAX ODS column with a mobile phase consisting of 0.2% triethylamine in 0.01 M ammonium carbonate: acetonitrile (70:30 v/v). The method showed high sensitivity with good linearity over the concentration range of 50 to 400 mug/ml. The method was successfully applied to the analysis of a pharmaceutical formulation containing (I) with excellent recovery. A kinetics investigation of the alkaline hydrolysis of (I) was carried out in sodium hydroxide solutions of 1, 1.5 and 2 N by monitoring the parent compound itself. The reaction order of (I) followed pseudo-first order kinetics. The activation energy could be estimated from the Arrhenius plot and it was found to be 12.331 kcal/mole. PMID:17202800

  16. Effect of alkaline earth metals on the liquid-phase hydrogenation of hydroquinone over Ru-based catalysts

    NASA Astrophysics Data System (ADS)

    Li, Hongwei; Ji, Dong; Li, Yu; Liang, Yalan; Li, Gui Xian

    2015-12-01

    A series of Ru-based catalysts modified by alkaline earth metals were prepared by the impregnation-precipitation method and characterized using transmission electron microscopy, X-ray diffraction, ICP optical emission spectroscopy, Infrared Spectroscopy of adsorbed pyridine analysis and surface area analysis. The performance of the catalysts was measured via liquid-phase hydroquinone hydrogenation reaction. Results show that the Ru-Sr/NaY catalyst has the best activity and selectivity among those Ru-based catalysts. The conversion of hydroquinone and the selectivity to 1,4-cyclohexanediol reached up to 99.6% and 89.6% at optimum reaction condition (700 r/min, 423 K and 5 MPa pressure of H2 in 3 h). This may be attributed to the fact that the right amount of Strontium is beneficial to the good dispersion of the ruthenium nanoclusters on the surface of NaY and modify the acidic properties of the catalyst. Moreover, IR of adsorbed pyridine analysis suggested the proper ratio of L/B acid of the catalysts played an important role in the performance of the hydroquinone hydrogenation reaction.

  17. Transient thermal analysis for radioactive liquid mixing operations in a large-scaled tank

    DOE PAGESBeta

    Lee, S. Y.; Smith, III, F. G.

    2014-07-25

    A transient heat balance model was developed to assess the impact of a Submersible Mixer Pump (SMP) on radioactive liquid temperature during the process of waste mixing and removal for the high-level radioactive materials stored in Savannah River Site (SRS) tanks. The model results will be mainly used to determine the SMP design impacts on the waste tank temperature during operations and to develop a specification for a new SMP design to replace existing longshaft mixer pumps used during waste removal. The present model was benchmarked against the test data obtained by the tank measurement to examine the quantitative thermalmore » response of the tank and to establish the reference conditions of the operating variables under no SMP operation. The results showed that the model predictions agreed with the test data of the waste temperatures within about 10%.« less

  18. Transient thermal analysis for radioactive liquid mixing operations in a large-scaled tank

    SciTech Connect

    Lee, S. Y.; Smith, III, F. G.

    2014-07-25

    A transient heat balance model was developed to assess the impact of a Submersible Mixer Pump (SMP) on radioactive liquid temperature during the process of waste mixing and removal for the high-level radioactive materials stored in Savannah River Site (SRS) tanks. The model results will be mainly used to determine the SMP design impacts on the waste tank temperature during operations and to develop a specification for a new SMP design to replace existing longshaft mixer pumps used during waste removal. The present model was benchmarked against the test data obtained by the tank measurement to examine the quantitative thermal response of the tank and to establish the reference conditions of the operating variables under no SMP operation. The results showed that the model predictions agreed with the test data of the waste temperatures within about 10%.

  19. Supported liquid membranes in radioactive waste treatment processes: Recent experience and perspective

    SciTech Connect

    Nechaev, A.F.; Projaev, V.V.; Kapranchik, V.P.

    1995-12-31

    Recent experience in practical application of Supported Liquid Membranes (SLM or SUPLIM) both in the hydrometallurgy and nuclear technology has been analyzed. The results obtained allow one to consider SUPLIM as a promising technology for radioactive waste treatment. This statement is based on the evaluation of integrated socioeconomic effects, including quantity of additional chemicals, the volume of secondary technological streams and secondary wastes, simplicity and the low costs of equipment used, potential possibility to organize in situ process, and the level of the harmful impact on personnel. 35 refs.

  20. Detection of free liquid in drums of radioactive waste. [Patent application

    DOEpatents

    Not Available

    1979-10-16

    A nondestructive thermal imaging method for detecting the presence of a liquid such as water within a sealed container is described. The process includes application of a low amplitude heat pulse to an exterior surface area of the container, terminating the heat input and quickly mapping the resulting surface temperatures. The various mapped temperature values can be compared with those known to be normal for the container material and substances in contact. The mapped temperature values show up in different shades of light or darkness that denote different physical substances. The different substances can be determined by direct observation or by comparison with known standards. The method is particularly applicable to the detection of liquids above solidified radioactive wastes stored in sealed containers.

  1. Removal of Radioactive Nuclides from Mo-99 Acidic Liquid Waste - 13027

    SciTech Connect

    Hsiao, Hsien-Ming; Pen, Ben-Li

    2013-07-01

    About 200 liters highly radioactive acidic liquid waste originating from Mo-99 production was stored at INER (Institute of Nuclear Energy Research). A study regarding the treatment of the radioactive acidic liquid waste was conducted to solve storage-related issues and allow discharge of the waste while avoiding environmental pollution. Before discharging the liquid waste, the acidity, NO{sub 3}{sup -} and Hg ions in high concentrations, and radionuclides must comply with environmental regulations. Therefore, the treatment plan was to neutralize the acidic liquid waste, remove key radionuclides to reduce the dose rate, and then remove the nitrate and mercury ions. Bench tests revealed that NaOH is the preferred solution to neutralize the high acidic waste solution and the pH of solution must be adjusted to 9∼11 prior to the removal of nuclides. Significant precipitation was produced when the pH of solution reached 9. NaNO{sub 3} was the major content in the precipitate and part of NaNO{sub 3} was too fine to be completely collected by filter paper with a pore size of approximately 3 μm. The residual fine particles remaining in solution therefore blocked the adsorption column during operation. Two kinds of adsorbents were employed for Cs-137 and a third for Sr-90 removal to minimize cost. For personnel radiation protection, significant lead shielding was required at a number of points in the process. The final process design and treatment facilities successfully treated the waste solutions and allowed for environmentally compliant discharge. (authors)

  2. Partitioning of K, U, and Th between sulfide and silicate liquids - Implications for radioactive heating of planetary cores

    NASA Technical Reports Server (NTRS)

    Murrell, M. T.; Burnett, D. S.

    1986-01-01

    Experimental partitioning studies are reported of K, U, and Th between silicate and FeFeS liquids designed to test the proposal that actinide partitioning into sulfide liquids is more important then K partitioning in the radioactive heating of planetary cores. For a basaltic liquid at 1450 C and 1.5 GPa, U partitioning into FeFeS liquids is five times greater than K partitioning. A typical value for the liquid partition coefficient for U from a granitic silicate liquid at one atmosphere at 1150 C and low fO2 is about 0.02; the coefficient for Th is similar. At low fO2 and higher temperature, experiments with basaltic liquids produce strong Ca and U partitioning into the sulfide liquid with U coefficient greater than one. The Th coefficient is less strongly affected.

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

    SciTech Connect

    OSMANLIOGLU, Ahmet Erdal

    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)

  4. Investigation of three-dimensional localisation of radioactive sources using a fast organic liquid scintillator detector

    NASA Astrophysics Data System (ADS)

    Gamage, K. A. A.; Joyce, M. J.; Taylor, G. C.

    2013-04-01

    In this paper we discuss the possibility of locating radioactive sources in space using a scanning-based method, relative to the three-dimensional location of the detector. The scanning system comprises an organic liquid scintillator detector, a tungsten collimator and an adjustable equatorial mount. The detector output is connected to a bespoke fast digitiser (Hybrid Instruments Ltd., UK) which streams digital samples to a personal computer. A radioactive source has been attached to a vertical wall and the data have been collected in two stages. In the first case, the scanning system was placed a couple of metres away from the wall and in the second case it moved few centimetres from the previous location, parallel to the wall. In each case data were collected from a grid of measurement points (set of azimuth angles for set of elevation angles) which covered the source on the wall. The discrimination of fast neutrons and gamma rays, detected by the organic liquid scintillator detector, is carried out on the basis of pulse gradient analysis. Images are then produced in terms of the angular distribution of events for total counts, gamma rays and neutrons for both cases. The three-dimensional location of the neutron source can be obtained by considering the relative separation of the centres of the corresponding images of angular distribution of events. The measurements have been made at the National Physical Laboratory, Teddington, Middlesex, UK.

  5. 1H and 23Na MAS NMR spectroscopy of cationic species in CO2 selective alkaline earth metal porous silicoaluminophosphates prepared via liquid and solid state ion exchange

    NASA Astrophysics Data System (ADS)

    Arévalo-Hidalgo, Ana G.; Dugar, Sneha; Fu, Riqiang; Hernández-Maldonado, Arturo J.

    2012-07-01

    The location of extraframework cations in Sr2+ and Ba2+ ion-exchanged SAPO-34 was estimated by means of 1H and 23Na MAS NMR spectroscopy and spectral deconvolution. Incorporation of the alkaline earth metal cations onto the SAPO framework was achieved via liquid state ion exchange, coupled partial detemplation/solid-state ion exchange, and combination of both techniques. MAS NMR revealed that the level of ion exchange was limited by the presence of protons and sodium cations near hexagonal prisms (site SI), which are relatively difficult to exchange with the alkaline earth metal due to steric and charge repulsion criteria. In addition, the presence of ammonium cations in the supercages facilitated the exchange of otherwise tenacious hydrogen as corroborated by unit cell compositional data as well as enhanced CO2 adsorption at low partial pressures. The extraframework ammonium species were produced from partial detemplation of the structure-directing agent employed for the SAPO-34 synthesis, tetraethylammonium.

  6. Ambient formic acid in southern California air: A comparison of two methods, Fourier transform infrared spectroscopy and alkaline trap-liquid chromatography with UV detection

    SciTech Connect

    Grosjean, D. ); Tuazon, E.C. ); Fujita, E. )

    1990-01-01

    Formic acid is an ubiquitous component of urban smog. Sources of formic acid in urban air include direct emissions from vehicles and in situ reaction of ozone with olefins. Ambient levels of formic acid in southern California air were first measured some 15 years ago by Hanst et al. using long-path Fourier transform infrared spectroscopy (FTIR). All subsequent studies of formic acid in the Los Angeles area have involved the use of two methods, either FTIR or collection on alkaline traps followed by gas chromatography, ion chromatography, or liquid chromatography analysis with UV detection, ATLC-UV. The Carbon Species Methods Comparison Study (CSMCS), a multilaboratory air quality study carried out in August 1986 at a southern California smog receptor site, provided an opportunity for direct field comparison of the FTIR and alkaline trap methods. The results of the comparison are presented in this brief report.

  7. Survey: utilization of zeolites for the removal of radioactivity from liquid waste streams. [178 references

    SciTech Connect

    Roddy, J.W.

    1981-08-01

    A survey was made of the literature and of experience at selected nuclear installations to provide information on the stability of inorganic ion exchangers when used for the decontamination of both low-level and high-level radioactive liquids. Results of past campaigns at the Savannah River Plant, Oak Ridge National Laboratory, and Rockwell Hanford Operations were examined. In addition, the performance of zeolites used for controlling water quality in nuclear fuel storage basins was evaluated. The literature survey served as a guide for identifying relevant material from foreign sources and supplemented the information obtained by direct contact of domestic researchers. The study included a brief review of the physical and chemical properties of zeolites. A secondary objective of the study was to compile data on the corrosion resistance of containment materials for zeolites.

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

    SciTech Connect

    Martynov, B.V.

    1995-10-01

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

  9. Design features of the radioactive Liquid-Fed Ceramic Melter system

    SciTech Connect

    Holton, L.K. Jr.

    1985-06-01

    During 1983, the Pacific Northwest Laboratory (PNL), at the request of the Department of Energy (DOE), undertook a program with the principal objective of testing the Liquid-Fed Ceramic Melter (LFCM) process in actual radioactive operations. This activity, termed the Radioactive LFCM (RLFCM) Operations is being conducted in existing shielded hot-cell facilities in B-Cell of the 324 Building, 300 Area, located at Hanford, Washington. This report summarizes the design features of the RLFCM system. These features include: a waste preparation and feed system which uses pulse-agitated waste preparation tanks for waste slurry agitation and an air displacement slurry pump for transferring waste slurries to the LFCM; a waste vitrification system (LFCM) - the design features, design approach, and reasoning for the design of the LFCM are described; a canister-handling turntable for positioning canisters underneath the RLFCM discharge port; a gamma source positioning and detection system for monitoring the glass fill level of the product canisters; and a primary off-gas treatment system for removing the majority of the radionuclide contamination from the RLFCM off gas. 8 refs., 48 figs., 6 tabs.

  10. Utilization of natural hematite as reactive barrier for immobilization of radionuclides from radioactive liquid waste.

    PubMed

    El Afifi, E M; Attallah, M F; Borai, E H

    2016-01-01

    Potential utilization of hematite as a natural material for immobilization of long-lived radionuclides from radioactive liquid waste was investigated. Hematite ore has been characterized by different analytical tools such as Fourier transformer infrared (FTIR), X-ray fluorescence (XRF), powder X-ray diffraction (XRD), thermogravimetry (TG) and differential thermal (DT) analysis, scanning electron microscopy (SEM) and BET-surface area. In this study, europium was used as REEs(III) and as a homolog of Am(III)-isotopes (such as (241)Am of 432.6 y, (242m)Am of 141 y and (243)Am of 7370 y). Micro particles of the hematite ore were used for treatment of radioactive waste containing (152+154)Eu(III). The results indicated that 96% (4.1 × 10(4) Bq) of (152+154)Eu(III) was efficiently retained onto hematite ore. Kinetic experiments indicated that the processes could be simulated by a pseudo-second-order model and suggested that the process may be chemisorption in nature. The applicability of Langmuir, Freundlich and Temkin models was investigated. It was found that Langmuir isotherm exhibited the best fit with the experimental results. It can be concluded that hematite is an economic and efficient reactive barrier for immobilization of long-lived radio isotopes of actinides and REEs(III). PMID:26465672

  11. Best available technology for the Los Alamos National Laboratory Radioactive Liquid Waste Treatment Facility

    SciTech Connect

    Midkiff, W.S.; Romero, R.L.; Suazo, I.L.; Garcia, R.; Parsons, R.M.

    1993-10-15

    The existing Los Alamos National Laboratory TA-50 liquid radioactive waste treatment plant RLWP has been in service for over thirty years, during this period many technical, regulatory, and processing changes have occurred. The existing facility can no longer comply with the demands and requirements for continued operation, and would not be able to comply with anticipated stringent future contaminant discharge limitations. Either a major upgrading or replacement of the existing facility is required. In order to assess the most appropriate means of providing an adequate facility to comply with predicted requirements for Ta-50, this Best Available Technology (BAT) Study was conducted to compare feasible technical and economic alternatives in order to define the most favorable technology configuration. This report consists of eleven sections. Section 1 provides a general introduction and background of the TA-50 operations and the basis for this study. Section 2 provides a technical discussion of the unit processes at TA-50 and several other comparable operations at other DOE sites. Section 3 addresses the evaluation and selection of appropriate treatment processes. Section 4 provides an analysis of environmental issues and concerns. Section 5 presents the rationale for the selection of preferred process configurations. Section 6 is the evaluation of operational issues. Section 7 addresses energy and resource use topics. Section 8 provides an economic analysis, and Section 9 summarizes the evaluation and the identification of the BAT. These sections are augmented by appendices. The report identifies the construction of a new radioactive liquid waste treatment facility as the BAT. Based on the information analyzed for this study, this option appears to provide the best combination of environmental compliance, operability, and economic value.

  12. Process for the removal of radioactive iodine from a liquid, especially urine, and apparatus to carry out the process

    SciTech Connect

    Benes, I.; Heinzel, F.; Muller-Duysing, W.

    1980-10-21

    Apparatus for removing radioactive iodine from a liquid, especially urine, having a collecting means, a treating tank with a mixer-pump, a filter, and a catch basin in which radiation is detected. The apparatus includes various control means and magnetic values to provide automatic operation.

  13. M558 radioactive tracer diffusion. [diffusion coefficients of Zn-65 in liquid zinc under weightlessness conditions

    NASA Technical Reports Server (NTRS)

    Ukanwa, A. O.

    1974-01-01

    This experiment was performed in Skylab 3 with two objectives in mind. First, the experimental self-diffusion coefficients for liquid zinc were to be determined in a convection-free environment. Secondly the reduction in convective mixing in earth gravity by going into the zero-gravity environment of space was to be estimated. The experiment was designed to utilize high temperatures and linear thermal gradients provided by the M518 Multipurpose Electric Furnace, and the radioactivity of zinc-65 of 245-day half-life to investigate self-diffusion in liquid zinc. The distribution of zinc-65 tracer, after melting, maintaining at soak temperature for 1 hour of soak time and then resolidifying, was obtained by sample sectioning. The concentration of activity of each section (microcurie-gram) was plotted against positions along the sample axial and radial position. Experimental data and theoretical results from solution of Fick's law of diffusion in one dimensional were compared. Samples tested on earth showed very rapid diffusion. Diffusion coefficient in unit gravity was 50 times the zero-gravity diffusion coefficient of Skylab.

  14. Modeling and risk assessment of a 30-Year-old subsurface radioactive-liquid drain field

    NASA Astrophysics Data System (ADS)

    Dawson, Lon A.; Pohl, Phillip I.

    1997-11-01

    The contamination from a 30-year-old radioactive liquid drain field was assessed for movement in the subsurface and potential risks to humans. This assessment included determining field concentrations of cesium 137 (137Cs) and other inorganic contaminants and modeling of the flow and transport of the liquid waste that was sent to the drain field. The field investigation detected no contamination deeper than 15 feet (4.6 m) from the bottom of the drain field. Prediction of the water content of the vadose zone showed no saturated conditions for times greater than 10 years after the known infiltration. Sensitivity analysis of the modeling parameters showed the equilibrium sorption coefficient to be the most important factor in predicting the contaminant plumes. Calibration of modeling results with field data gave a 137Cs sorption coefficient that is within the range of values found in the literature. The risk assessment for the site showed that the contamination poses no significant risk to human health.

  15. Reduction of N2O and NO generation in anaerobic-aerobic (low dissolved oxygen) biological wastewater treatment process by using sludge alkaline fermentation liquid.

    PubMed

    Zhu, Xiaoyu; Chen, Yinguang

    2011-03-15

    This paper reported an efficient method to significantly reduce nitrous oxide (N(2)O) and nitric oxide (NO) generation in anaerobic-aerobic (low dissolved oxygen) processes. It was found that by the use of waste-activated sludge alkaline fermentation liquid as the synthetic wastewater-carbon source, compared with the commonly used carbon source in the literature (e.g., acetic acid), the generation of N(2)O and NO was reduced by 68.7% and 50.0%, respectively, but the removal efficiencies of total phosphorus (TP) and total nitrogen (TN) were improved. Both N(2)O and NO were produced in the low dissolved oxygen (DO) stage, and the use of sludge fermentation liquid greatly reduced their generation from the denitrification. The presences of Cu(2+) and propionic acid in fermentation liquid were observed to play an important role in the reduction of N(2)O and NO generation. The analysis of the activities of denitrifying enzymes suggested that sludge fermentation liquid caused the significant decrease of both nitrite reductase activity to NO reductase activity ratio and NO reductase activity to N(2)O reductase activity ratio, which resulted in the lower generation of NO and N(2)O. Fluorescence in situ hybridization analysis indicated that the number of glycogen accumulating bacteria, which was reported to be relevant to nitrous oxide generation, in sludge fermentation liquid reactor was much lower than that in acetic acid reactor. The quantitative detection of the nosZ gene, encoding nitrous oxide reductase, showed that the use of fermentation liquid increased the number of bacteria capable of reducing N(2)O to N(2). The feasibility of using sludge fermentation liquid to reduce NO and N(2)O generation in an anaerobic-low DO process was finally confirmed for a municipal wastewater. PMID:21322643

  16. A NEW, SMALL DRYING FACILITY FOR WET RADIOACTIVE WASTE AND LIQUIDS

    SciTech Connect

    Oldiges, Olaf; Blenski, Hans-Juergen

    2003-02-27

    Due to the reason, that in Germany every Waste, that is foreseen to be stored in a final disposal facility or in a long time interim storage facility, it is necessary to treat a lot of waste using different drying technologies. In Germany two different drying facilities are in operation. The GNS Company prefers a vacuum-drying-technology and has built and designed PETRA-Drying-Facilities. In a lot of smaller locations, it is not possible to install such a facility because inside the working areas of that location, the available space to install the PETRA-Drying-Facility is too small. For that reason, GNS decided to design a new, small Drying-Facility using industrial standard components, applying the vacuum-drying-technology. The new, small Drying-Facility for wet radioactive waste and liquids is presented in this paper. The results of some tests with a prototype facility are shown in chapter 4. The main components of that new facility are described in chapter 3.

  17. EXPLORING ENGINEERING CONTROL THROUGH PROCESS MANIPULATION OF RADIOACTIVE LIQUID WASTE TANK CHEMICAL CLEANING

    SciTech Connect

    Brown, A.

    2014-04-27

    One method of remediating legacy liquid radioactive waste produced during the cold war, is aggressive in-tank chemical cleaning. Chemical cleaning has successfully reduced the curie content of residual waste heels in large underground storage tanks; however this process generates significant chemical hazards. Mercury is often the bounding hazard due to its extensive use in the separations process that produced the waste. This paper explores how variations in controllable process factors, tank level and temperature, may be manipulated to reduce the hazard potential related to mercury vapor generation. When compared using a multivariate regression analysis, findings indicated that there was a significant relationship between both tank level (p value of 1.65x10{sup -23}) and temperature (p value of 6.39x10{sup -6}) to the mercury vapor concentration in the tank ventilation system. Tank temperature showed the most promise as a controllable parameter for future tank cleaning endeavors. Despite statistically significant relationships, there may not be confidence in the ability to control accident scenarios to below mercury’s IDLH or PAC-III levels for future cleaning initiatives.

  18. Solvent extraction of [sup 99]Tc from radioactive intermediate liquid waste by dibenzo-18-crown-6.

    SciTech Connect

    Paviet-Hartmann, P.

    2001-01-01

    Technetium is one of the most prominent problems since its most stable specie in the environment, the pertechnetate ion, is highly mobile and considered as a long-term hazard in nuclear waste disposal. Because of the small activities of 99Tc relative to other fission products 137Cso r 90Sr,a nd its long half-life time (t1/2 = 2.1{center_dot}10{sup 5} yrs), 99Tci s one of the key isotopes that should always be analyzed in the radioactive liquid waste streams from the reprocessing industry where the largest concentrations are to be expected. Furthermore, as a pure beta-emitter, Tc has to be isolated from the intermediate level waste (ILW) stream prior to any measurement in such complex media. We have developed a method for 99Tc extraction providing recommendations that will be useful for extracting it from acid and basic ILW. The extraction of 99Tc from L W by dibenzo-18-crown-6 (DB18C6) has been investigated and a simplex optimization of key parameters involved in the procedure has allowed us to set up their best values. Experiments have been carried out on synthetic and real effluents from La Hague reprocessing plant, France, and results show that DB18C6 is highly selective towards 99Tc. The application of this procedure has been successfully demonstrated through the analysis of actual waste streams coming from two reprocessing plants at La Hague and Marcoule, France.

  19. SOLVENT EXTRACTION OF 99Tc FROM RADIOACTIVE INTERMEDIATE LIQUID WASTE BY DIBENZO-18-CROWN-6

    SciTech Connect

    Paviet-Hartmann, P.

    2002-02-25

    Technetium is one of the most prominent problems since its most stable specie in the environment, the pertechnetate ion, is highly mobile and considered as a long-term hazard in nuclear waste disposal. Because of the small activities of 99Tc relative to other fission products 137Cs or 90Sr, and its long half-life time (t1/2 = 2.1 x 105 yrs), 99Tc is one of the key isotopes that should always be analyzed in the radioactive liquid waste streams from the reprocessing industry where the largest concentrations are to be expected. Furthermore, as a pure beta-emitter, 99Tc has to be isolated from the intermediate level waste (ILW) stream prior to any measurement in such complex media. We have developed a method for 99Tc extraction providing recommendations that will be useful for extracting it from acid and basic ILW. The extraction of 99Tc from ILW by dibenzo-18-crown-6 (DB18C6) has been investigated and a simplex optimization of key parameters involved in the procedure has allowed us to set u p their best values. Experiments have been carried out on synthetic and real effluents from La Hague reprocessing plant, France, and results show that DB18C6 is highly selective towards 99Tc. The application of this procedure has been successfully demonstrated through the analysis of actual waste streams coming from two reprocessing plants at La Hague and Marcoule, France.

  20. Utilization of coal fly ash in solidification of liquid radioactive waste from research reactor.

    PubMed

    Osmanlioglu, Ahmet Erdal

    2014-05-01

    In this study, the potential utilization of fly ash was investigated as an additive in solidification process of radioactive waste sludge from research reactor. Coal formations include various percentages of natural radioactive elements; therefore, coal fly ash includes various levels of radioactivity. For this reason, fly ashes have to be evaluated for potential environmental implications in case of further usage in any construction material. But for use in solidification of radioactive sludge, the radiological effects of fly ash are in the range of radioactive waste management limits. The results show that fly ash has a strong fixing capacity for radioactive isotopes. Specimens with addition of 5-15% fly ash to concrete was observed to be sufficient to achieve the target compressive strength of 20 MPa required for near-surface disposal. An optimum mixture comprising 15% fly ash, 35% cement, and 50% radioactive waste sludge could provide the solidification required for long-term storage and disposal. The codisposal of radioactive fly ash with radioactive sludge by solidification decreases the usage of cement in solidification process. By this method, radioactive fly ash can become a valuable additive instead of industrial waste. This study supports the utilization of fly ash in industry and the solidification of radioactive waste in the nuclear industry. PMID:24638274

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

    SciTech Connect

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

    2011-01-12

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

  2. Alkaline Leaching of Key, Non-Radioactive Components from Simulants and Hanford Tank Sludge 241-S-110: Results of FY01 Studies

    SciTech Connect

    Rapko, Brian M.; Vienna, John D.; Sinkov, Serguei I.; Kim, Jinseong; Cisar, Alan J.

    2002-09-10

    This study addressed three aspects in selected alkaline leaching: first, the use of oxidants persulfate, permanganate, and ferrate as selective chromium-leaching agents from washed Hanford Tank S-110 solids under varying conditions of hydroxide concentration, temperature, and time was investigated. Second, the selective dissolution of solids containing mercury(II) oxide under alkaline conditions was examined. Various compounds were studied for their effectiveness in dissolving mercury under varying conditions of time, temperature, and hydroxide concentration in the leachate. Three compounds were studied: cysteine, iodide, and diethyldithiophosphoric acid (DEDTPA). Finally, the possibility of whether an oxidant bound to an anion-exchange resin can be used to effectively oxidize chromium(III) in alkaline solutions was addressed. The experimental results remain ambiguous to date; further work is required to reach any definitive conclusions as to the effectiveness of this approach.

  3. LOW LEVEL LIQUID RADIOACTIVE WASTE TREATMENT AT MURMANSK, RUSSIA: FACILITY UPGRADE AND EXPANSION

    SciTech Connect

    BOWERMAN,B.; CZAJKOWSKI,C.; DYER,R.S.; SORLIE,A.

    2000-03-01

    Today there exist many almost overfilled storage tanks with liquid radioactive waste in the Russian Federation. This waste was generated over several years by the civil and military utilization of nuclear power. The current waste treatment capacity is either not available or inadequate. Following the London Convention, dumping of the waste in the Arctic seas is no longer an alternative. Waste is being generated from today's operations, and large volumes are expected to be generated from the dismantling of decommissioned nuclear submarines. The US and Norway have an ongoing co-operation project with the Russian Federation to upgrade and expand the capacity of a treatment facility for low level liquid waste at the RTP Atomflot site in Murmansk. The capacity will be increased from 1,200 m{sup 3}/year to 5,000 m{sup 3} /year. The facility will also be able to treat high saline waste. The construction phase will be completed the first half of 1998. This will be followed by a start-up and a one year post-construction phase, with US and Norwegian involvement for the entire project. The new facility will consist of 9 units containing various electrochemical, filtration, and sorbent-based treatment systems. The units will be housed in two existing buildings, and must meet more stringent radiation protection requirements that were not enacted when the facility was originally designed. The US and Norwegian technical teams have evaluated the Russian design and associated documentation. The Russian partners send monthly progress reports to US and Norway. Not only technical issues must be overcome but also cultural differences resulting from different methods of management techniques. Six to eight hour time differentials between the partners make real time decisions difficult and relying on electronic age tools becomes extremely important. Language difficulties is another challenge that must be solved. Finding a common vocabulary, and working through interpreters make the

  4. DEMONSTRATION SOLIDIFICATION TESTS CONDUCTED ON RADIOACTIVELY CONTAMINATED ORGANIC LIQUIDS AT THE AECL WHITESHELL LABORATORIES

    SciTech Connect

    Ryz, R. A.; Brunkow, W. G.; Govers, R.; Campbell, D.; Krause, D.

    2002-02-25

    The AECL, Whiteshell Laboratory (WL) near Pinawa Manitoba, Canada, was established in the early 1960's to carry out AECL research and development activities for higher temperature versions of the CANDU{reg_sign} reactor. The initial focus of the research program was the Whiteshell Reactor-1 (WR-1) Organic Cooled Reactor (OCR) that began operation in 1965. The OCR program was discontinued in the early 1970's in favor of the successful heavy-water-cooled CANDU system. WR-1 continued to operate until 1985 in support of AECL nuclear research programs. A consequence of the Federal government's recent program review process was AECL's business decision to discontinue research programs and operations at the Whiteshell Laboratories and to consolidate its' activities at the Chalk River Laboratories. As a result, AECL received government concurrence in 1998 to proceed to plan actions to achieve closure of WL. The planning actions now in progress address the need to safely and effectively transition the WL site from an operational state, in support of AECL's business, to a shutdown and decommissioned state that meets the regulatory requirements for a licensed nuclear site. The decommissioning program that will be required at WL is unique within AECL and Canada since it will need to address the entire research site rather than individual facilities declared redundant. Accordingly, the site nuclear facilities are being systematically placed in a safe shutdown state and planning for the decommissioning work to place the facilities in a secure monitoring and surveillance state is in progress. One aspect of the shutdown activities is to deal with the legacy of radioactively contaminated organic liquid wastes. Use of a polymer powder to solidify these organic wastes was identified as one possibility for improved interim storage of this material pending final disposition.

  5. Spanish LLW and MLW disposal: durability of cemented materials in (Na, K)Cl simulated radioactive liquid waste.

    PubMed

    Goñi, S; Guerrero, A; Hernández, M S

    2001-01-01

    The microstructural stability or durability of a specific backfilling pozzolanic-cement mortar, which is employed in Spain, in concrete containers for the storage of low level liquid wastes (LLW) and medium level liquid wastes (MLW), has been studied by means of the Koch-Steinegger test at the temperatures of 20 and 40 degrees C during a period of 365 days. Mortar samples were immersed in salt solutions of 3.46 M NaCl and 3.46 M KCl to simulate the salinity of some radioactive liquid waste matrices. The resistance of the mortar to the saline solution attack is evaluated by the development of the relative flexural strength. The changes of the microstructure were followed by mercury intrusion porosimetry (MIP), scanning electron microscopy (SEM) and X-ray diffraction (XRD). Pore solution was extracted and analyzed at different periods of time to know the possible diffusion of sodium, chloride and potassium inside the microstructure. PMID:11150135

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

    SciTech Connect

    McVey, C.B.

    1994-10-13

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

  7. Calculation of releases of radioactive materials in gaseous and liquid effluents from pressurized water reactors (PWR-GALE Code). Revision 1

    SciTech Connect

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

    1985-04-01

    This report revises the original issuance of NUREG-0017, ''Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE-Code)'' (April 1976), to incorporate more recent operating data now available as well as the results of a number of in-plant measurement programs at operating pressurized water reactors. The PWR-GALE Code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents (i.e., the gaseous and liquid source terms). The US Nuclear Regulatory Commission uses the PWR-GALE Code to determine conformance with the requirements of Appendix I to 10 CFR Part 50.

  8. Sampling and analysis of radioactive liquid wastes and sludges in the Melton Valley and evaporator facility storage tanks at ORNL

    SciTech Connect

    Sears, M.B.; Botts, J.L.; Ceo, R.N.; Ferrada, J.J.; Griest, W.H.; Keller, J.M.; Schenley, R.L.

    1990-09-01

    The sampling and analysis of the radioactive liquid wastes and sludges in the Melton Valley Storage Tanks (MVSTs), as well as two of the evaporator service facility storage tanks at ORNL, are described. Aqueous samples of the supernatant liquid and composite samples of the sludges were analyzed for major constituents, radionuclides, total organic carbon, and metals listed as hazardous under the Resource Conservation and Recovery Act (RCRA). Liquid samples from five tanks and sludge samples from three tanks were analyzed for organic compounds on the Environmental Protection Agency (EPA) Target Compound List. Estimates were made of the inventory of liquid and sludge phases in the tanks. Descriptions of the sampling and analytical activities and tabulations of the results are included. The report provides data in support of the design of the proposed Waste Handling and Packaging Plant, the Liquid Low-Level Waste Solidification Project, and research and development activities (R D) activities in developing waste management alternatives. 7 refs., 8 figs., 16 tabs.

  9. 222-S radioactive liquid waste line replacement and 219-S secondary containment upgrade, Hanford Site, Richland, Washington

    SciTech Connect

    1995-01-01

    The U.S. Department of Energy (DOE) is proposing to: (1) replace the 222-S Laboratory (222-S) radioactive liquid waste drain lines to the 219-S Waste Handling Facility (219-S); (2) upgrade 219-S by replacing or upgrading the waste storage tanks and providing secondary containment and seismic restraints to the concrete cells which house the tanks; and (3) replace the transfer lines from 219-S to the 241-SY Tank Farm. This environmental assessment (EA) has been prepared in compliance with the National Environmental Policy Act (NEPA) of 1969, as amended, the Council on Environmental Quality Regulations for Implementing the Procedural Provisions of NEPA (40 Code of Federal Regulations [CFR] 1500-1508), and the DOE Implementing Procedures for NEPA (10 CFR 1021). 222-S is used to perform analytical services on radioactive samples in support of the Tank Waste Remediation System and Hanford Site environmental restoration programs. Activities conducted at 222-S include decontamination of analytical processing and support equipment and disposal of nonarchived radioactive samples. These activities generate low-level liquid mixed waste. The liquid mixed waste is drained through pipelines in the 222-S service tunnels and underground concrete encasements, to two of three tanks in 219-S, where it is accumulated. 219-S is a treatment, storage, and/or disposal (TSD) unit, and is therefore required to meet Washington Administrative Code (WAC) 173-303, Dangerous Waste Regulations, and the associated requirements for secondary containment and leak detection. The service tunnels are periodically inspected by workers and decontaminated as necessary to maintain as low as reasonably achievable (ALARA) radiation levels. Although no contamination is reaching the environment from the service tunnels, the risk of worker exposure is present and could increase. 222-S is expected to remain in use for at least the next 30 years to serve the Hanford Site environmental cleanup mission.

  10. Characterization of biocenoses in the storage reservoirs of liquid radioactive wastes of Mayak PA. Initial descriptive report.

    PubMed

    Pryakhin, E A; Mokrov, Yu G; Tryapitsina, G A; Ivanov, I A; Osipov, D I; Atamanyuk, N I; Deryabina, L V; Shaposhnikova, I A; Shishkina, E A; Obvintseva, N A; Egoreichenkov, E A; Styazhkina, E V; Osipova, O F; Mogilnikova, N I; Andreev, S S; Tarasov, O V; Geras'kin, S A; Trapeznikov, A V; Akleyev, A V

    2016-01-01

    As a result of operation of the Mayak Production Association (Mayak PA), Chelyabinsk Oblast, Russia, an enterprise for production and separation of weapon-grade plutonium in the Soviet Union, ecosystems of a number of water bodies have been radioactively contaminated. The article presents information about the current state of ecosystems of 6 special industrial storage reservoirs of liquid radioactive waste from Mayak PA: reservoirs R-3, R-4, R-9, R-10, R-11 and R-17. At present the excess of the radionuclide content in the water of the studied reservoirs and comparison reservoirs (Shershnyovskoye and Beloyarskoye reservoirs) is 9 orders of magnitude for (90)Sr and (137)Cs, and 6 orders of magnitude for alpha-emitting radionuclides. According to the level of radioactive contamination, the reservoirs of the Mayak PA could be arranged in the ascending order as follows: R-11, R-10, R-4, R-3, R-17 and R-9. In 2007-2012 research of the status of the biocenoses of these reservoirs in terms of phytoplankton, zooplankton, bacterioplankton, zoobenthos, aquatic plants, ichthyofauna, avifauna parameters was performed. The conducted studies revealed decrease in species diversity in reservoirs with the highest levels of radioactive and chemical contamination. This article is an initial descriptive report on the status of the biocenoses of radioactively contaminated reservoirs of the Mayak PA, and is the first article in a series of publications devoted to the studies of the reaction of biocenoses of the fresh-water reservoirs of the Mayak PA to a combination of natural and man-made factors, including chronic radiation exposure. PMID:26094572

  11. MD simulations of the formation of stable clusters in mixtures of alkaline salts and imidazolium-based ionic liquids.

    PubMed

    Méndez-Morales, Trinidad; Carrete, Jesús; Bouzón-Capelo, Silvia; Pérez-Rodríguez, Martín; Cabeza, Óscar; Gallego, Luis J; Varela, Luis M

    2013-03-21

    Structural and dynamical properties of room-temperature ionic liquids containing the cation 1-butyl-3-methylimidazolium ([BMIM](+)) and three different anions (hexafluorophosphate, [PF6](-), tetrafluoroborate, [BF4](-), and bis(trifluoromethylsulfonyl)imide, [NTf2](-)) doped with several molar fractions of lithium salts with a common anion at 298.15 K and 1 atm were investigated by means of molecular dynamics simulations. The effect of the size of the salt cation was also analyzed by comparing these results with those for mixtures of [BMIM][PF6] with NaPF6. Lithium/sodium solvation and ionic mobilities were analyzed via the study of radial distribution functions, coordination numbers, cage autocorrelation functions, mean-square displacements (including the analysis of both ballistic and diffusive regimes), self-diffusion coefficients of all the ionic species, velocity and current autocorrelation functions, and ionic conductivity in all the ionic liquid/salt systems. We found that lithium and sodium cations are strongly coordinated in two different positions with the anion present in the mixture. Moreover, [Li](+) and [Na](+) cations were found to form bonded-like, long-lived aggregates with the anions in their first solvation shell, which act as very stable kinetic entities within which a marked rattling motion of salt ions takes place. With very long MD simulation runs, this phenomenon is proved to be on the basis of the decrease of self-diffusion coefficients and ionic conductivities previously reported in experimental and computational results. PMID:23480174

  12. Effects of alkaline or liquid-ammonia treatment on crystalline cellulose: changes in crystalline structure and effects on enzymatic digestibility

    PubMed Central

    2011-01-01

    Background In converting biomass to bioethanol, pretreatment is a key step intended to render cellulose more amenable and accessible to cellulase enzymes and thus increase glucose yields. In this study, four cellulose samples with different degrees of polymerization and crystallinity indexes were subjected to aqueous sodium hydroxide and anhydrous liquid ammonia treatments. The effects of the treatments on cellulose crystalline structure were studied, in addition to the effects on the digestibility of the celluloses by a cellulase complex. Results From X-ray diffractograms and nuclear magnetic resonance spectra, it was revealed that treatment with liquid ammonia produced the cellulose IIII allomorph; however, crystallinity depended on treatment conditions. Treatment at a low temperature (25°C) resulted in a less crystalline product, whereas treatment at elevated temperatures (130°C or 140°C) gave a more crystalline product. Treatment of cellulose I with aqueous sodium hydroxide (16.5 percent by weight) resulted in formation of cellulose II, but also produced a much less crystalline cellulose. The relative digestibilities of the different cellulose allomorphs were tested by exposing the treated and untreated cellulose samples to a commercial enzyme mixture (Genencor-Danisco; GC 220). The digestibility results showed that the starting cellulose I samples were the least digestible (except for corn stover cellulose, which had a high amorphous content). Treatment with sodium hydroxide produced the most digestible cellulose, followed by treatment with liquid ammonia at a low temperature. Factor analysis indicated that initial rates of digestion (up to 24 hours) were most strongly correlated with amorphous content. Correlation of allomorph type with digestibility was weak, but was strongest with cellulose conversion at later times. The cellulose IIII samples produced at higher temperatures had comparable crystallinities to the initial cellulose I samples, but achieved

  13. Combined capillary electrophoresis and high performance liquid chromatography studies on the kinetics and mechanism of the hydrogen peroxide-thiocyanate reaction in a weakly alkaline solution.

    PubMed

    Hu, Ying; Song, Yanan; Horváth, Attila K; Cui, Yin; Ji, Chen; Zhao, Yuemin; Gao, Qingyu

    2014-03-01

    The hydrogen peroxide-thiocyanate reaction has been reinvestigated by means of capillary electrophoresis and high performance liquid chromatography under weakly alkaline conditions at 25.0±0.1 °C. Concentration-time series of thiocyanate, sulfate and cyanate have been followed by capillary electrophoresis as well as that of thiocyanate and hydrogen peroxide by HPLC. It has been clearly demonstrated that OxSCN(-) (where x=1, 2 and 3) cannot be accumulated in detectable amount in contrast to the results of Christy and Egeberg, hence these species can only be regarded as short-lived intermediates. It has been shown that the overall rate law is first-order with respect to both reactants, but no pH-dependence was observed within the pH range of 8.86-10.08. A simple kinetic model has been proposed to fit all the concentration-time curves simultaneously at five different pHs demonstrating the powerful combination of the experimental techniques CE and HPLC with simultaneous evaluation of kinetic curves. It is also enlightened that the quality of the buffer strongly affects the rate of the overall reaction that increases in the order of application of ammonia, phosphate, carbonate and borate, respectively at a constant ionic strength and pH. PMID:24468335

  14. Orbital free ab initio simulations of liquid alkaline earth metals: from pseudopotential construction to structural and dynamic properties.

    PubMed

    Rio, Beatriz G del; González, Luis E

    2014-11-19

    We have performed a comprehensive study of the properties of liquid Be, Ca and Ba, through the use of orbital free ab initio simulations. To this end we have developed a force-matching method to construct the necessary local pseudopotentials from standard ab initio calculations. The structural magnitudes are analyzed, including the average and local structures and the dynamic properties are studied. We find several common features, like an asymmetric second peak in the structure factor, a large amount of local structures with five-fold symmetry, a quasi-universal behaviour of the single-particle dynamic properties and a large degree of positive dispersion in the propagation of collective density fluctuations, whose damping is dictated by slow thermal relaxations and fast viscoelastic ones. Some peculiarities in the dynamic properties are however observed, like a very high sound velocity and a large violation of the Stokes-Einstein relation for Be, or an extremely high positive dispersion and a large slope in the dispersion relation of shear waves at the onset of the wavevector region where they are supported for Ba. PMID:25347355

  15. Development of a New Thermal HF Plasma Reactor for the Destruction of Radioactive Organic Halogen Liquid Wastes

    SciTech Connect

    Bournonville, B.; Meillot, E.; Girold, C.

    2006-07-01

    A newly patented process employing thermal plasma for destruction of radioactive organic halogen liquid wastes is proposed. This studied safe system can destroy a great variety of wastes, even mixed together, using plasma torch as high temperature source. At the exit of the process, only non-toxic products are formed as atmospheric gases, liquid water and halogen sodium salt. The process has been built with the help of thermodynamic and kinetic simulations. A good atomic stoichiometry is necessary for avoiding the formation of solid carbon (soot) or toxic COCl{sub 2}. That why liquid water is added to the waste in the plasma flow. Then, an introduction of air cools and dilutes the formed gases and adds oxidant agent achieving oxidation of explosive H{sub 2} and toxic CO. Due to the high concentration of hydrochloric acid, an efficient wet treatment using soda traps it. Subsequently, the exhaust gases are only composed of Ar, O{sub 2}, N{sub 2}, CO{sub 2} and H{sub 2}O. In the first experimental step, pure organic molecules, mixed or not, without halogen have been destroyed. The experimental results show that all the compounds have been completely destroyed and only CO{sub 2} and H{sub 2}O have been formed without formation of any toxic compound or soot. After these encouraging results, chlorinated compounds as dichloromethane or chloroform have been destroyed by the process. In this case, the results are close to the previous one with an important formation of hydrochloric acid, as expected, which was well trapped by the soda to respect the French norm of rejection. A specific parameter study has been done with dichloromethane for optimising the operating condition to experimentally observe the influence of different parameters of the process as the stoichiometry ratio between waste and water, the air addition flow, the waste flow. The final aim of this study is to develop a clean process for treatment of radioactive organic halogen compounds. A small scale reactor

  16. Evaluation of polyacrylonitrile (PAN) as a binding polymer for absorbers used to treat liquid radioactive wastes

    SciTech Connect

    Sebesta, F.; John, J.; Motl, A.; Stamberg, K.

    1995-11-01

    The chemical and radiation stability of polyacrylonitrile (PAN) in the form of beads (B-PAN), similar to the beads of composite absorbers, and one selected composite absorber (ammonium molybdophosphate, the active component in PAN binder [AMP-PAN], a prospective candidate for the treatment of acidic wastes) were studied. Aqueous 1M HNO{sub 3} + 1M NaNO{sub 3}, 1M NaOH + 1M NaNO{sub 3}, and 1M NaOH were chosen as simulants of DOE acidic and alkaline wastes. In addition,radiation stability was determined indistilled water. The chemical stability of B-PAN and AMP-PAN beads was tested for a period up to one month of contact with the solution at ambient temperature. The radiation stability of the beads was checked in a radiation dose range 10{sup 3}--10{sup 6} Gy (10{sup 5}--10{sup 8} rads). In acidic solutions the stability of PAN binder was proved not to be limited by either chemical or radiation decomposition. PAN binder may thus be used for preparing composite absorbers for treatment of acid wastes from DOE facilities. The same conclusion is valid for alkaline solutions with pH up to 13. In highly alkaline solutions (concentration of NAOH higher than I M) and in the presence of NaNO{sub 3}, the stability of the tested polyacrylonitrile polymer was sufficient for applications not extending over 10 days. Cross-linking of the polymer caused by ionizing radiation was found to have a positive influence on chemical stability. This effect enables a longer period of applicability of PAN-based composite absorbers. Because of the high sorption rate achievable with PAN-based absorbers, the stability achieved is sufficient for most applications in the DOE complex. The chemical stability of binding polymer may also be further improved by testing another, more suitable type of polymer from the broad family of polyacrylonitrile polymers.

  17. HWMA/RCRA Closure Plan for the CPP-648 Radioactive Solid and Liquid Waste Storage Tank System (VES-SFE-106)

    SciTech Connect

    S. K. Evans

    2006-08-15

    This Hazardous Waste Management Act/Resource Conservation and Recovery Act closure plan for the Radioactive Solid and Liquid Waste Storage Tank System located in the adjacent to the Sludge Tank Control House (CPP-648), Idaho Nuclear Technology and Engineering Center, Idaho National Laboratory, was developed to meet the interim status closure requirements for a tank system. The system to be closed includes a tank and associated ancillary equipment that were determined to have managed hazardous waste. The CPP-648 Radioactive Solid and Liquid Waste Storage Tank System will be "cleaned closed" in accordance with the requirements of the Hazardous Waste Management Act/Resource Conservation and Recovery Act as implemented by the Idaho Administrative Procedures Act and 40 Code of Federal Regulations 265. This closure plan presents the closure performance standards and methods of acheiving those standards for the CPP-648 Radioactive Solid and Liquid Waste Storage Tank System.

  18. Separation, Concentration, and Immobilization of Technetium and Iodine from Alkaline Supernate Waste

    SciTech Connect

    James Harvey; Michael Gula

    1998-12-07

    Development of remediation technologies for the characterization, retrieval, treatment, concentration, and final disposal of radioactive and chemical tank waste stored within the Department of Energy (DOE) complex represents an enormous scientific and technological challenge. A combined total of over 90 million gallons of high-level waste (HLW) and low-level waste (LLW) are stored in 335 underground storage tanks at four different DOE sites. Roughly 98% of this waste is highly alkaline in nature and contains high concentrations of nitrate and nitrite salts along with lesser concentrations of other salts. The primary waste forms are sludge, saltcake, and liquid supernatant with the bulk of the radioactivity contained in the sludge, making it the largest source of HLW. The saltcake (liquid waste with most of the water removed) and liquid supernatant consist mainly of sodium nitrate and sodium hydroxide salts. The main radioactive constituent in the alkaline supernatant is cesium-137, but strontium-90, technetium-99, and transuranic nuclides are also present in varying concentrations. Reduction of the radioactivity below Nuclear Regulatory Commission (NRC) limits would allow the bulk of the waste to be disposed of as LLW. Because of the long half-life of technetium-99 (2.1 x 10 5 y) and the mobility of the pertechnetate ion (TcO 4 - ) in the environment, it is expected that technetium will have to be removed from the Hanford wastes prior to disposal as LLW. Also, for some of the wastes, some level of technetium removal will be required to meet LLW criteria for radioactive content. Therefore, DOE has identified a need to develop technologies for the separation and concentration of technetium-99 from LLW streams. Eichrom has responded to this DOE-identified need by demonstrating a complete flowsheet for the separation, concentration, and immobilization of technetium (and iodine) from alkaline supernatant waste.

  19. Alkaline quinone flow battery.

    PubMed

    Lin, Kaixiang; Chen, Qing; Gerhardt, Michael R; Tong, Liuchuan; Kim, Sang Bok; Eisenach, Louise; Valle, Alvaro W; Hardee, David; Gordon, Roy G; Aziz, Michael J; Marshak, Michael P

    2015-09-25

    Storage of photovoltaic and wind electricity in batteries could solve the mismatch problem between the intermittent supply of these renewable resources and variable demand. Flow batteries permit more economical long-duration discharge than solid-electrode batteries by using liquid electrolytes stored outside of the battery. We report an alkaline flow battery based on redox-active organic molecules that are composed entirely of Earth-abundant elements and are nontoxic, nonflammable, and safe for use in residential and commercial environments. The battery operates efficiently with high power density near room temperature. These results demonstrate the stability and performance of redox-active organic molecules in alkaline flow batteries, potentially enabling cost-effective stationary storage of renewable energy. PMID:26404834

  20. Separation of alkali, alkaline earth and rare earth cations by liquid membranes containing macrocyclic carriers. Third progress report, September 1, 1980-April 1, 1981

    SciTech Connect

    Christensen, J.J.

    1981-04-15

    The overall objective of this project is to study the use of liquid membrane systems employing macrocyclic ligand carriers in making separations among metal cations. During the third year of the project, work continued in the development of a mathematical model to describe cation transport. The model was originally developed to describe the relationship between cation transport rate (J/sub M/) and the cation-macrocycle stability constant (K). The model was tested by determining the rates of transport of alkali and alkaline earth cations through chloroform membranes containing carrier ligands where the stability constants for their reaction with cations in methanol were known. From the results, it is clear that the model correctly describes the dependence of J/sub M/ on log K. The model also correctly describes the effect of cation concentration and carrier concentration on cation transport rates, as detailed in the previous progress report. During the third year of the project, the transport model was expanded so as to apply to competitive transport of cations from mixtures of two cations in the source aqueous phase. Data were collected under these conditions and the ability of the model to predict the flux of each cation was tested. Representative data of this type are presented along with corresponding data which were obtained when each cation was transported by the same carrier from a source phase containing only that cation. Comparison of transport rates determined under the two experimental conditions indicates that the relationship between the two sets of data is complex. To date, a few of these data involving transport from binary cation mixtures have been tested against the transport model. It was found that the model correctly predicts the cation fluxes from cation mixtures. These preliminary results indicate that the transport model can successfully predict separation factors when cation mixtures are used.

  1. Corrosion Control Measures For Liquid Radioactive Waste Storage Tanks At The Savannah River Site

    SciTech Connect

    Wiersma, B. J.; Subramanian, K. H.

    2012-11-27

    The Savannah River Site has stored radioactive wastes in large, underground, carbon steel tanks for approximately 60 years. An assessment of potential degradation mechanisms determined that the tanks may be vulnerable to nitrate- induced pitting corrosion and stress corrosion cracking. Controls on the solution chemistry and temperature of the wastes are in place to mitigate these mechanisms. These controls are based upon a series of experiments performed using simulated solutions on materials used for construction of the tanks. The technical bases and evolution of these controls is presented in this paper.

  2. Treatment Options for Liquid Radioactive Waste. Factors Important for Selecting of Treatment Methods

    SciTech Connect

    Dziewinski, J.J.

    1998-09-28

    The cleanup of liquid streams contaminated with radionuclides is obtained by the selection or a combination of a number of physical and chemical separations, processes or unit operations. Among those are: Chemical treatment; Evaporation; Ion exchange and sorption; Physical separation; Electrodialysis; Osmosis; Electrocoagulation/electroflotation; Biotechnological processes; and Solvent extraction.

  3. Process for the denitrification of high-level radioactive liquid wastes

    NASA Astrophysics Data System (ADS)

    Gattys, F. J.

    1981-07-01

    Commercial paraformaldehyde with a low water content and mean particle size smaller than 60 microns is introduced as denitrifying agent at about 110 C into the liquid waste. A buffer of citric acid and phosphate is added during the denitrification reaction. The escaping gases contain nitrogen oxides which are recombined and concentrated to about 40 percent by weight nitric acid.

  4. Report on the flowsheet model for the electrochemical treatment of liquid radioactive wastes

    SciTech Connect

    Hobbs, D.T.

    1995-04-11

    The objective of this report is to describe the modeling and optimization procedure for the electrochemical removal of nitrates and nitrites from low level radioactive wastes. The simulation is carried out in SPEEDUP{trademark}, which is a state of the art flowsheet modeling package. The flowsheet model will provide a better understanding of the process and aid in the scale-up of the system. For example, the flowsheet model has shown that the electrochemical cell must be operated in batch mode to achieve 95% destruction. The present status of the flowsheet model is detailed in this report along with a systematic description of the batch optimization of the electrochemical cell. Results from two batch runs and one optimization run are also presented.

  5. Flowsheet model for the electrochemical treatment of liquid radioactive wastes. Final report

    SciTech Connect

    Hobbs, D.T.; Prasad, S.; Farell, A.E.; Weidner, J.W.; White, R.E.

    1995-12-31

    The objective of this report is to describe the modeling and optimization procedure for the electrochemical removal of nitrates and nitrites from low level radioactive wastes. The simulation is carried out in SPEEDUP{trademark}, which is a state of the art flowsheet modeling package. The flowsheet model will provide a better understanding of the process and aid in the scale-up of the system. For example, the flowsheet model has shown that the electrochemical cell must be operated in batch mode to achieve 95 percent destruction. The flowsheet model is detailed in this report along with a systematic description of the batch optimization of the electrochemical cell. Results from two batch runs and one optimization run are also presented.

  6. Analysis by high-performance liquid chromatography of radioactively labeled carbohydrate components of proteoglycans

    SciTech Connect

    Lohmander, L.S.

    1986-04-01

    Methods were developed for the separation of radioactively labeled carbohydrate components of proteoglycans by isocratic ion-moderated partition HPLC. Neutral sugars were separated after hydrolysis in trifluoroacetic acid with baseline separation between glucose, xylose, galactose, fucose, and mannose. N-Acetylneuraminic acid, N-acetylated hexosamines, glucose, galactose, and xylitol were likewise well separated from each other under isocratic elution conditions. Glucuronic acid, iduronic acid, and their lactones were separated after hydrolysis in formic acid and sulfuric acid. Glucosamine, galactosamine, galactosaminitol, and glucosaminitol were separated by HPLC on a cation exchanger with neutral buffer after hydrolysis in hydrochloric acid. THe separation techniques also proved useful in fractionation of exoglycosidase digests of O- and N-linked oligosaccharides. Separations of aldoses, hexosamines, and uronic acids were adapted to sensitive photometric detection.

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

    SciTech Connect

    Ehst, D.; Nuclear Engineering Division

    2010-08-04

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

  8. Determination of vapor-liquid equilibrium data and decontamination factors needed for the development of evaporator technology for use in volume reduction of radioactive waste streams

    SciTech Connect

    Betts, S.E.

    1993-10-01

    A program is currently in progress at Argonne National Laboratory to evaluate and develop evaporator technology for concentrating radioactive waste streams. By concentrating radioactive waste streams, disposal costs can be significantly reduced. To effectively reduce the volume of waste, the evaporator must achieve high decontamination factors so that the distillate is sufficiently free of radioactive material. One technology that shows a great deal of potential for this application is being developed by LICON, Inc. In this program, Argonne plans to apply LICON`s evaporator designs to the processing of radioactive solutions. Concepts that need to be incorporated into the design of the evaporator include, criticality safety, remote operation and maintenance, and materials of construction. To design an effective process for concentrating waste streams, both solubility and vapor-liquid equilibrium data are needed. The key issue, however, is the high decontamination factors that have been demonstrated by this equipment. Two major contributions were made to this project. First, a literature survey was completed to obtain available solubility and vapor-liquid equilibrium data. Some vapor-liquid data necessary for the project but not available in the literature was obtained experimentally. Second, the decontamination factor for the evaporator was determined using neutron activation analysis (NAA).

  9. Subsurface disposal of liquid low-level radioactive wastes at Oak Ridge, Tennessee

    SciTech Connect

    Stow, S.H.; Haase, C.S.

    1986-01-01

    At Oak Ridge National Laboratory (ORNL) subsurface injection has been used to dispose of low-level liquid nuclear waste for the last two decades. The process consists of mixing liquid waste with cement and other additives to form a slurry that is injected under pressure through a cased well into a low-permeability shale at a depth of 300 m (1000 ft). The slurry spreads from the injection well along bedding plane fractures and forms solid grout sheets of up to 200 m (660 ft) in radius. Using this process, ORNL has disposed of over 1.5 x 10/sup 6/ Ci of activity; the principal nuclides are /sup 90/Sr and /sup 137/Cs. In 1982, a new injection facility was put into operation. Each injection, which lasts some two days, results in the emplacement of approximately 750,000 l (180,000 gal) of slurry. Disposal cost per liter is approximately $0.30, including capital costs of the facility. This subsurface disposal process is fundamentally different from other operations. Wastes are injected into a low-permeability aquitard, and the process is designed to isolate nuclides, preventing dispersion in groundwaters. The porosity into which wastes are injected is created by hydraulically fracturing the host formation along bedding planes. The site is in the structurally complex Valley and Ridge Province. The stratigraphy consists of lower Paleozoic rocks. Investigations are under way to determine the long-term hydrologic isolation of the injection zone and the geochemical impact of saline groundwater on nuclide mobility. Injections are monitored by gamma-ray logging of cased observation wells to determine grout sheet orientation after an injection. Recent monitoring work has involved the use of tiltmeters, surface uplift surveys, and seismic arrays. 26 refs., 7 figs.

  10. Subsurface disposal of liquid low-level radioactive wastes at Oak Ridge, Tennessee

    SciTech Connect

    Stow, S.H.; Haase, C.S.

    1986-01-01

    At Oak Ridge National Laboratory (ORNL) subsurface injection has been used to dispose of low-level liquid nuclear waste for the last two decades. The process consists of mixing liquid waste with cement and other additives to form a slurry that is injected under pressure through a cased well into a low-permeability shale at a depth of 300 m. The slurry spreads from the injection well along bedding plane fractures and forms solid grout sheets of up to 200 m in radius. Using this process, ORNL has disposed of over 1.5 x 10/sup 6/ Ci of activity; the principal nuclides are /sup 90/Sr and /sup 137/Cs. In 1982, a new injection facility was put into operation. Each injection, which lasts some two days, results in the emplacement of approximately 750,000 liters of slurry. Disposal cost per liter is about $0.30, including capital costs of the facility. This subsurface disposal process is fundamentally different from other operations. Wastes are injected into a low-permeability aquitard, and the process is designed to isolate nuclides, preventing dispersion in groundwaters. The porosity into which wastes are injected is created by hydraulically fracturing the host formation along bedding planes. Investigations are under way to determine the long-term hydrologic isolation of the injection zone and the geochemical impact of saline groundwater on nuclide mobility. Injections are monitored by gamma-ray logging of cased observation wells to determine grout sheet orientation after an injection. Recent monitoring work has involved the use of tiltmeters, surface uplift surveys, and seismic arrays. Recent regulatory constraints may cause permanent cessation of the operation. Federal and state statutes, written for other types of injection facilities, impact the ORNL facility. This disposal process, which may have great applicability for disposal of many wastes, including hazardous wastes, may not be developed for future use.

  11. Removal of Mercury from the Off-Gas from Thermal Treatment of Radioactive Liquid Waste

    SciTech Connect

    Deldebbio, John Anthony; Olson, Lonnie Gene

    2001-05-01

    Acidic, radioactive wastes with a high nitrate concentration, and containing mercury are currently being stored at the Idaho Nuclear Technology and Engineering Center (INTEC). In the past, these wastes were converted into a dry, granular solid by a high temperature fluidized-bed calcination process. In the future, the calcined solids may be immobilized by a vitrification process prior to disposal. It has been proposed that a vitrification facility be built to treat the acidic wastes, as well as the calcined solids. As was the case with the calcination process, NOx levels in the vitrification off-gas are expected to be high, and mercury emissions are expected to exceed the Maximum Control Technology (MACT) limits. Mitigation of mercury emissions by wet scrubbing, followed by adsorption onto activated carbon is being investigated. Scoping tests with sulfur-impregnated activated carbon, KCl-impregnated activated carbon and non-impregnated activated carbon were conducted with a test gas containing1% NO2, 28% H2O, 4% O2 and 67% N2. Average removal efficiencies for Hgo and HgCl2 were 100 ± 2.5% and 99 ± 3.6% respectively, for sulfur-impregnated carbon. The KCl-impregnated carbon removed 99 ± 4.6% HgCl2. The removal efficiency of the non-impregnated carbon was 99 ± 3.6% for HgCl2. No short-term detrimental effects due to NO2 and H2O were observed. These results indicate that, placed downstream of a wet scrubber, an activated carbon adsorption bed has the potential of reducing mercury levels sufficiently to enable compliance with the MACT limit. Long-term exposure tests, and bed size optimization studies are planned for the future.

  12. Ocean Circulation Modeling for Aquatic Dispersion of Liquid Radioactive Effluents from Nuclear Power Plants

    SciTech Connect

    Chung, Y.G.; Lee, G.B.; Bang, S.Y.; Choi, S.B.; Lee, S.U.; Yoon, J.H.; Nam, S.Y.; Lee, H.R.

    2006-07-01

    Recently, three-dimensional models have been used for aquatic dispersion of radioactive effluents in relation to nuclear power plant siting based on the Notice No. 2003-12 'Guideline for investigating and assessing hydrological and aquatic characteristics of nuclear facility site' of the Ministry of Science and Technology (MOST) in Korea. Several nuclear power plants have been under construction or planed, which are Shin-Kori Unit 1 and 2, Shin-Wolsong Unit 1 and 2, and Shin-Ulchin Unit 1 and 2. For assessing the aquatic dispersion of radionuclides released from the above nuclear power plants, it is necessary to know the coastal currents around sites which are affected by circulation of East Sea. In this study, a three dimensional hydrodynamic model for the circulation of the East Sea of Korea has been developed as the first phase, which is based on the RIAMOM (Research Institute of Applied Mechanics' Ocean Model, Kyushu University, Japan). The model uses the primitive equation with hydrostatic approximation, and uses Arakawa-B grid system horizontally and Z coordinate vertically. Model domain is 126.5 deg. E to 142.5 deg. E of east longitude and 33 deg. N and 52 deg. N of the north latitude. The space of the horizontal grid was 1/12 deg. to longitude and latitude direction and vertical level was divided to 20. This model uses Generalized Arakawa Scheme, Slant Advection, and Mode-Splitting Method. The input data were from JODC (Japan Oceanographic Data Center), KNFRDI (Korea National Fisheries Research and Development Institute), and ECMWF (European Center for Medium-Range Weather Forecasts). The modeling results are in fairly good agreement with schematic patterns of the surface circulation in the East Sea/Japan Sea. The local current model and aquatic dispersion model of the coastal region will be developed as the second phase. The oceanic dispersion experiments will be also carried out by using ARGO Drifter around a nuclear power plant site. (authors)

  13. The Polymers for Liquid Radioactive Waste Solidification: a Lost Chapter in the History of Engineering or a Step Forward? - 13529

    SciTech Connect

    Pokhitonov, Yury; Kelley, Dennis

    2013-07-01

    Ideas on the application of polymers for the liquid radioactive waste immobilization go a way back, and the first studies in the area were published 30-40 years ago. One should admit that regardless of the fairly large number of publications appeared in the past years currently the interest in this work came down greatly. It was the successful assimilation and worldwide implementation of the LRW cementation technology caused a slump in the interest in polymers. But today it's safe to say that the situation slowly changes, particularly due to the market appearance of the high-tech polymers manufactured by Nochar Company, and unique properties of these polymers gradually raise the demand in various countries. The results of multiple experiments performed with the simulated solutions have passed the comprehensive tests with actual waste. The economic effect from the implementation of the new technology is defined by the volume reduction of waste coming onto the repository, by the decline in the cost of transportation and of the repository construction on account of cutting down the construction volume. Interesting results have been obtained during the search for the technical decisions that would allow using the polymer materials in the processing technology of the industrial toxic waste. One more promising area of the possible application of polymers should be pointed out. It is the application of polymer materials as the assets for the emergency damage control when the advantages of the polymers become obvious. (authors)

  14. Flow and transport modeling of liquid radioactive waste injection using data from the Siberian Chemical Plant Injection Site

    NASA Astrophysics Data System (ADS)

    Shestakov, V.; Kuvaev, A.; Lekhov, A.; Pozdniakov, S.; Rybalchenko, A.; Zubkov, A.; Davis, P.; Kalinina, E.

    2002-06-01

    The focus of our investigation was simulating pre-injection and post-injection subsurface conditions at the waste disposal site of liquid radioactive wastes at the Siberian Chemical Complex (SCC). The main environmental and human safety concern posed by this site is related to the potential radionuclide discharge into the nearby Tom River and into the existing public-water-supply well fields located 10-13 km away. Even though (within the site) the two lower injection aquifers are isolated from the upper aquifers by a relatively continuous aquitard, in terms of regional flow they represent one hydrogeologic system that is affected by injection as well as by groundwater withdrawal from the upper aquifers and groundwater discharge into the river. Groundwater flow and transport models were developed to simulate regional flow and waste migration. Even after 1,000 years, none of the simulations indicated that there is any serious potential of high-concentration contamination of water supply wells and the discharge zone. In that time frame, simulation indicated a potential for upward movement of some amounts of contaminants through the heterogeneous sandy-clay aquitard. That is why a conceptual model incorporating heterogeneity of the clay aquitard with the possibility of preferential flow via sandy windows needs to be developed. Additional field characterization of the aquitard properties should be performed along the potential contaminant migration pathways that lead to the groundwater discharge zone.

  15. Assay for inorganic pyrophosphate in chondrocyte culture using anion-exchange high-performance liquid chromatography and radioactive orthophosphate labeling

    SciTech Connect

    Prins, A.P.; Kiljan, E.; v.d. Stadt, R.J.; v.d. Korst, J.K.

    1986-02-01

    A method is described for determination of inorganic pyrophosphate (PPi) in cell culture medium and in rabbit articular chondrocytes grown in the presence of radioactive orthophosphate (/sup 32/Pi). Intra- and extracellular /sup 32/PPi formed was measured using high-performance liquid chromatographic (HPLC) separation of the PPi from orthophosphate (Pi) and other phosphate-containing compounds. The chromatographic separation on a weak anion-exchange column is based on the extent to which various phosphate compounds form complexes with Mg2+ at low pH and the rate at which such formation occurs. These complexes are eluted more readily than the uncomplexed compounds. Best results were obtained using a simultaneous gradient of Mg2+ ions and ionic strength. In this case separation of small amounts of PPi from a large excess of Pi was possible without prior removal of Pi or extraction of the PPi fraction. The assay is also useful for measurement of inorganic pyrophosphatase activity. The sensitivity of the assay depends on the specific activity of the added /sup 32/Pi and on the culture conditions, but is comparable with the most sensitive of the enzymatic assays. Sample preparation, particularly deproteinization, proved to be of importance. The losses of PPi which occur during procedures of this sort due to hydrolysis and coprecipitation were quantitated.

  16. Reconstruction of the radionuclide spectrum of liquid radioactive waste released into the Techa river in 1949-1951.

    PubMed

    Mokrov, Yuri G

    2003-04-01

    The major part of the liquid radioactive waste released by the Mayak Production Association (PA) radiochemical plant into the Techa river occurred in 1949-1951, but there is information on only one single radiochemical analysis of a water sample taken on 24 and 25 September 1951. These data are here used to assess the spectrum of radionuclides that were released between 1949 and 1951. For this purpose, details of the radiochemical methods of radionuclide extraction and radiometric measurements of beta-activity used at Mayak PA in the 1950s have been taken into account. It is concluded that the data from the radiochemical measurements agree with the theoretical composition of fission products in uranium after exposure times in the reactor (120 days) and subsequent hold times (35 days) that were typical for the procedures at that time. The results of the analysis are at variance with assumptions that underlie the current Techa river dosimetry system. They confirm the conclusion that the external doses to the Techa river residents in the critical period up to 1952 were predominantly due to short-lived fission products. PMID:12720000

  17. [Transport processes of low-level radioactive liquid effluent of nuclear power station in closed water body].

    PubMed

    Wu, Guo-Zheng; Xu, Zong-Xue

    2012-07-01

    The transport processes of low-level radioactive liquid effluent of Xianning nuclear power station in the closed water body Fushui Reservoir are simulated using the EFDC model. Six nuclides concentration distribution with different half-lives in the reservoir are analyzed under the condition of 97% guarantee rate incoming water and four-running nuclear power units. The results show that the nuclides concentration distribution is mainly affected by the flow field of the reservoir and the concentration is decided by the half-lives of nuclide and the volume of incoming water. In addition, the influence region is enlarged as increasing of half-life and tends to be stable when the half-life is longer than 5 years. Moreover, the waste water discharged from the outlet of the nuclear power plant has no effect on the water-intake for the outlet located at the upstream of the water-intake and the flow field flows to the dam of the reservoir. PMID:23002624

  18. Barium borosilicate glass a potential matrix for immobilization of sulfate bearing high-level radioactive liquid waste

    NASA Astrophysics Data System (ADS)

    Kaushik, C. P.; Mishra, R. K.; Sengupta, P.; Kumar, Amar; Das, D.; Kale, G. B.; Raj, Kanwar

    2006-11-01

    Borosilicate glass formulations adopted worldwide for immobilization of high-level radioactive liquid waste (HLW) is not suitable for sulphate bearing HLW, because of its low solubility in such glass. A suitable glass matrix based on barium borosilicate has been developed for immobilization of sulphate bearing HLW. Various compositions based on different glass formulations were made to examine compatibility with waste oxide with around 10 wt% sulfate content. The vitrified waste product obtained from barium borosilicate glass matrix was extensively evaluated for its characteristic properties like homogeneity, chemical durability, glass transition temperature, thermal conductivity, impact strength, etc. using appropriate techniques. Process parameters like melt viscosity and pour temperature were also determined. It is found that SB-44 glass composition (SiO 2: 30.5 wt%, B 2O 3: 20.0 wt%, Na 2O: 9.5 wt% and BaO: 19.0 wt%) can be safely loaded with 21 wt% waste oxide without any phase separation. The other product qualities of SB-44 waste glass are also found to be on a par with internationally adopted waste glass matrices. This formulation has been successfully implemented in plant scale.

  19. Spectrophotometric determination of plutonium in highly radioactive liquid waste using an internal standardization technique with neodymium(III).

    PubMed

    Surugaya, Naoki; Taguchi, Shigeo; Sato, Soichi; Watahiki, Masaru; Hiyama, Toshiaki

    2008-03-01

    A simple and rapid spectrophotometric method has been developed for the determination of Pu in highly radioactive liquid waste. This method uses Nd(III) as an internal standard, which enables us to determine the concentration of Pu and to authenticate the whole analytical scheme as well. A Nd(III) standard mixed with a sample solution and Pu was quantitatively oxidized to Pu(VI) with Ce(IV) in a nitric acid medium, having the maximum absorbance at 830 nm. A spectrophotometric measurement of Pu(VI) was subsequently performed to determine the concentration compared with the maximum absorbance of Nd(III) at 795 nm. It was estimated that the relative expanded uncertainty for a real sample is less than 10%. The limit of detection was calculated to be 1.8 mg/L (3 sigma). The proposed method was also validated through comparison experiments with isotope dilution mass spectrometry, and was successfully applied to analysis for nuclear waste management at spent nuclear fuel reprocessing plants. PMID:18332546

  20. [Development of a simple quantitative method for the strontium-89 concentration of radioactive liquid waste using the plastic scintillation survey meter for beta rays].

    PubMed

    Narita, Hiroto; Tsuchiya, Yuusuke; Hirase, Kiyoshi; Uchiyama, Mayuki; Fukushi, Masahiro

    2012-11-01

    Strontium-89 (89Sr: pure beta, E; 1.495 MeV-100%, halflife: 50.5 days) chloride is used as pain relief from bone metastases. An assay of 89Sr is difficult because of a pure beta emitter. For management of 89Sr, we tried to evaluate a simple quantitative method for the 59Sr concentration of radioactive liquid waste using scintillation survey meter for beta rays. The counting efficiency of the survey meter with this method was 35.95%. A simple 30 minutes measurement of 2 ml of the sample made the quantitative measurement of 89Sr practical. Reducing self-absorption of the beta ray in the solution by counting on the polyethlene paper improved the counting efficiency. Our method made it easy to manage the radioactive liquid waste under the legal restrictions. PMID:23402205

  1. Low-level liquid radioactive waste treatment at Murmansk, Russia: Technical design and review of facility upgrade and expansion

    SciTech Connect

    Dyer, R.S.; Diamante, J.M.; Duffey, R.B.

    1996-07-01

    The governments of Norway and the US have committed their mutual cooperation and support the Murmansk Shipping Company (MSCo) to expand and upgrade the Low-Level Liquid Radioactive Waste (LLRW) treatment system located at the facilities of the Russian company RTP Atomflot, in Murmansk, Russia. RTP Atomflot provides support services to the Russian icebreaker fleet operated by the MSCo. The objective is to enable Russia to permanently cease disposing of this waste in Arctic waters. The proposed modifications will increase the facility`s capacity from 1,200 m{sup 3} per year to 5,000 m{sup 3} per year, will permit the facility to process high-salt wastes from the Russian Navy`s Northern fleet, and will improve the stabilization and interim storage of the processed wastes. The three countries set up a cooperative review of the evolving design information, conducted by a joint US and Norwegian technical team from April through December, 1995. To ensure that US and Norwegian funds produce a final facility which will meet the objectives, this report documents the design as described by Atomflot and the Russian business organization, ASPECT, both in design documents and orally. During the detailed review process, many questions were generated, and many design details developed which are outlined here. The design is based on the adsorption of radionuclides on selected inorganic resins, and desalination and concentration using electromembranes. The US/Norwegian technical team reviewed the available information and recommended that the construction commence; they also recommended that a monitoring program for facility performance be instituted.

  2. The alkaline and alkaline-carbonatite magmatism from Southern Brazil

    NASA Astrophysics Data System (ADS)

    Ruberti, E.; Gomes, C. D. B.; Comin-Chiaramonti, P.

    2015-12-01

    Early to Late Cretaceous lasting to Paleocene alkaline magmatism from southern Brazil is found associated with major extensional structural features in and around the Paraná Basin and grouped into various provinces on the basis of several data. Magmatism is variable in size, mode of occurrence and composition. The alkaline rocks are dominantly potassic, a few occurrences showing sodic affinity. The more abundant silicate rocks are evolved undersaturated to saturated in silica syenites, displaying large variation in igneous forms. Less evolved types are restricted to subvolcanic environments and outcrops of effusive suites occur rarely. Cumulatic mafic and ultramafic rock types are very common, particularly in the alkali-carbonatitic complexes. Carbonatite bodies are represented by Ca-carbonatites and Mg-carbonatites and more scarcely by Fe-carbonatites. Available radiometric ages for the alkaline rocks fit on three main chronological groups: around 130 Ma, subcoveal with the Early Cretaceous flood tholeiites of the Paraná Basin, 100-110 Ma and 80-90 Ma (Late Cretaceous). The alkaline magmatism also extends into Paleocene times, as indicated by ages from some volcanic lavas. Geochemically, alkaline potassic and sodic rock types are distinguished by their negative and positive Nb-Ta anomalies, respectively. Negative spikes in Nb-Ta are also a feature common to the associated tholeiitic rocks. Sr-Nd-Pb systematics confirm the contribution of both HIMU and EMI mantle components in the formation of the alkaline rocks. Notably, Early and Late Cretaceous carbonatites have the same isotopic Sr-Nd initial ratios of the associated alkaline rocks. C-O isotopic Sr-Nd isotopic ratios indicate typical mantle signature for some carbonatites and the influence of post-magmatic processes in others. Immiscibility of liquids of phonolitic composition, derived from mafic alkaline parental magmas, has been responsible for the origin of the carbonatites. Close association of alkaline

  3. Long-term management of liquid high-level radioactive wastes stored at the Western New York Nuclear Service Center, West Valley

    NASA Astrophysics Data System (ADS)

    1981-07-01

    Environmental implications of possible alternatives for long-term management of the liquid high-level radioactive wastes stored in underground tanks in West Valley, New York were assessed and compared. Four basic alternatives, as well as options within these alternatives, considered in the EIS: (1) onsite processing to a terminal waste form for shipment and disposal in a federa repository; (2) onsite conversion to a solid interim form for shipment to a federal waste facility for later processing to a terminal form and shipment and subsequent disposal in a federal repository; (3) mixing the liquid wastes with cement and other additives, pouring it back into the existing tanks, and leaving onsite; and (4) no action (continued storage of the wastes in liquid form in the underground tanks at West Valley). Mitigative measures for environmental impacts were be required.

  4. Process for extracting technetium from alkaline solutions

    SciTech Connect

    Moyer, B.A.; Sachleben, R.A.; Bonnesen, P.V.

    1994-12-31

    This invention relates generally to a process for extracting technetium from nuclear wastes and more particularly to a process for extracting technetium from alkaline waste solutions containing technetium and high concentrations of alkali metal nitrates. A process for extracting technetium values from an aqueous alkaline solution containing at least one alkali metal hydroxide and at least one alkali metal nitrate comprises the steps of: contacting the aqueous alkaline solution with a solvent consisting of a crown ether in a diluent, the diluent being a water-immiscible organic liquid in which the crown ether is soluble, for a period of time sufficient to selectively extract the technetium values from the aqueous alkaline solution into the solvent; separating the solvent containing the technetium values from the aqueous alkaline solution; and stripping the technetium values from the solvent by contacting the solvent with water.

  5. Technetium recovery from high alkaline solution

    DOEpatents

    Nash, Charles A.

    2016-07-12

    Disclosed are methods for recovering technetium from a highly alkaline solution. The highly alkaline solution can be a liquid waste solution from a nuclear waste processing system. Methods can include combining the solution with a reductant capable of reducing technetium at the high pH of the solution and adding to or forming in the solution an adsorbent capable of adsorbing the precipitated technetium at the high pH of the solution.

  6. {sup 1}H and {sup 23}Na MAS NMR spectroscopy of cationic species in CO{sub 2} selective alkaline earth metal porous silicoaluminophosphates prepared via liquid and solid state ion exchange

    SciTech Connect

    Arevalo-Hidalgo, Ana G.; Dugar, Sneha; Fu, Riqiang; Hernandez-Maldonado, Arturo J.

    2012-07-15

    The location of extraframework cations in Sr{sup 2+} and Ba{sup 2+} ion-exchanged SAPO-34 was estimated by means of {sup 1}H and {sup 23}Na MAS NMR spectroscopy and spectral deconvolution. Incorporation of the alkaline earth metal cations onto the SAPO framework was achieved via liquid state ion exchange, coupled partial detemplation/solid-state ion exchange, and combination of both techniques. MAS NMR revealed that the level of ion exchange was limited by the presence of protons and sodium cations near hexagonal prisms (site SI), which are relatively difficult to exchange with the alkaline earth metal due to steric and charge repulsion criteria. In addition, the presence of ammonium cations in the supercages facilitated the exchange of otherwise tenacious hydrogen as corroborated by unit cell compositional data as well as enhanced CO{sub 2} adsorption at low partial pressures. The extraframework ammonium species were produced from partial detemplation of the structure-directing agent employed for the SAPO-34 synthesis, tetraethylammonium. - Graphical abstract: MAS NMR was used to elucidate the position the cationic species in alkaline earth metal exchanged silicoaluminophosphates. These species played a significant role during the ion exchange process and, therefore, the materials ultimate CO{sub 2} adsorption performance. Highlights: Black-Right-Pointing-Pointer Location of extraframework Sr{sup 2+} or Ba{sup 2+} cations was estimated by means of {sup 1}H and {sup 23}Na MAS NMR. Black-Right-Pointing-Pointer Level of Sr{sup 2+} or Ba{sup 2+} ion exchange was limited by the presence of protons and sodium cations. Black-Right-Pointing-Pointer Presence of ammonium cations in the supercages facilitated the exchange. Black-Right-Pointing-Pointer Sr{sup 2+} and Ba{sup 2+} ion exchanged SAPOs are outstanding CO{sub 2} adsorbents.

  7. ALP (Alkaline Phosphatase) Test

    MedlinePlus

    ... known as: ALK PHOS; Alkp Formal name: Alkaline Phosphatase Related tests: AST ; ALT ; GGT ; Bilirubin ; Liver Panel ; Bone Markers ; Alkaline Phosphatase Isoenzymes; Bone Specific ALP All content on Lab ...

  8. DEVELOPMENT OF A CROSSFLOW FILTER TO REMOVE SOLIDS FROM RADIOACTIVE LIQUID WASTE: COMPARISON OF TEST DATA WITH OPERATING EXPERIENCE - 9119

    SciTech Connect

    Poirier, M; David Herman, D; Samuel Fink, S; Julius Lacerna, J

    2009-03-01

    In 2008, the Savannah River Site (SRS) began treatment of liquid radioactive waste from its Tank Farms. To treat waste streams containing {sup 137}Cs, {sup 90}Sr, and actinides, SRS developed the Actinide Removal Process (ARP) and the Modular Caustic Side Solvent Extraction Unit (MCU). The Actinide Removal Process contacts the waste with monosodium titanate (MST) to sorb strontium and select actinides. After MST contact, the process filters the resulting slurry to remove the MST (with sorbed strontium and actinides) and any entrained sludge. The filtrate is transported to the MCU to remove cesium. The solid particle removed by the filter are concentrated to {approx} 5 wt %, washed to reduce the concentration of dissolved sodium, and transported to the Defense Waste Processing Facility (DWPF) for vitrification. The authors conducted tests with 0.5 {micro} and 0.1 {micro} Mott sintered stainless steel crossflow filter at bench-scale (0.19 ft{sup 2} surface area) and pilot-scale (11.2 ft{sup 2}). The collected data supported design of the filter for the process and identified preferred operating conditions for the full-scale process (230 ft{sup 2}). The testing investigated the influence of operating parameters, such as filter pore size, axial velocity, transmembrane pressure, and solids loading, on filter flux, and validated the simulant used for pilot-scale testing. The conclusions from this work follow: (1) The 0.1 {micro} Mott sintered stainless steel filter produced higher flux than the 0.5 {micro} filter. (2) The filtrate samples collected showed no visible solids. (3) The filter flux with actual waste is comparable to the filter flux with simulated waste, with the simulated waste being conservative. This result shows the simulated sludge is representative of the actual sludge. (4) When the data is adjusted for differences in transmembrane pressure, the filter flux in the Actinide Removal Process is comparable to the filter flux in the bench-scale and pilot

  9. Structural Integrity Program for the 300,000-Gallon Radioactive Liquid Waste Storage Tanks at the Idaho Nuclear Technology and Engineering Center

    SciTech Connect

    Bryant, J.W.; Nenni, J.A.; Yoder, T.S.

    2003-04-22

    This report provides a record of the Structural Integrity Program for the 300,000-gal liquid waste storage tanks and associated equipment at the Idaho Nuclear Technology and Engineering Center, as required by U.S. Department of Energy M 435.1-1, ''Radioactive Waste Management Manual.'' This equipment is known collectively as the Tank Farm Facility. The conclusion of this report is that the Tank Farm Facility tanks, vaults, and transfer systems that remain in service for storage are structurally adequate, and are expected to remain structurally adequate over the remainder of their planned service life through 2012. Recommendations are provided for continued monitoring of the Tank Farm Facility.

  10. Separation and determination of americium in low-level alkaline waste of NPP origin

    NASA Astrophysics Data System (ADS)

    Todorov, B.; Djingova, R.; Nikiforova, A.

    2006-01-01

    The aim of this work is to develop a short and cost-saving procedure for the determination of 241Am in sludge sample of the alkaline low-level radioactive waste (LL LRAW) collected from Nuclear Power Plant “Kozloduy”. The determination of americium was a part of a complex analytical approach, where group actinide separation was achieved. An anion exchange was used for separation of americium from uranium, plutonium and iron. For the separation of americium extraction with diethylhexyl phosphoric acid (DEHPA) was studied. The final radioactive samples were prepared by micro co-precipitation with NdF3, counted by alpha and gamma spectrometry. The procedure takes 2 hours. The recovery yield of the procedure amounts to (95 ± 1.5)% and the detection limit is 53 mBq/kg 241Am (t=150 000 s). The analytical procedure was applied for actual liquid wastes and results were compared to standard procedure.

  11. Separation of alkali, alkaline earth and rare earth cations by liquid membranes containing macrocyclic carriers. Fourth progress report, 1 November 1981-31 July 1982

    SciTech Connect

    Christensen, J J; Izatt, R M

    1982-07-31

    The H/sub 2/O-CHCl/sub 3/-H/sub 2/O liquid membrane system was characterized with respect to the effect on cation (K/sup +/) transport rate of salt concentration and anion type. A bulk liquid membrane cell was used. A mathematical model for cation flux is being developed for several cations, several macrocycles, and mixtures of two or three cations. Eu/sup 3 +/ was not transported by 18-crown-6, but its reduced from Eu/sup 2 +/ was. Cation transport properties of calixarenes are also being investigated. Emulsion membrane systems were studied as a way of increasing the cation transport. Pb/sup 2 +/ was found to be transported by dicyclohexano-18-crown-6 through the liquid membrane. Transport rates of metal cation nitrates were measured in a water-toluene-water emulsion membrane system. 14 figures, 7 tables. (DLC)

  12. Implementation plan for liquid low-level radioactive waste systems under the FFA for fiscal years 1996 and 1997 at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    SciTech Connect

    1996-06-01

    This document is the fourth annual revision of the plans and schedules for implementing the Federal Facility Agreement (FFA) compliance program, originally submitted in 1992 as ES/ER-17&D1, Federal Facility Agreement Plans and Schedules for Liquid Low-Level Radioactive Waste Tank Systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee. This document summarizes the progress that has been made to date implementing the plans and schedules for meeting the FFA commitments for the Liquid Low-Level Waste (LLLW) System at Oak Ridge National Laboratory (ORNL). In addition, this document lists FFA activities planned for FY 1997. Information presented in this document provides a comprehensive summary to facilitate understanding of the FFA compliance program for LLLW tank systems and to present plans and schedules associated with remediation, through the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) process, of LLLW tank systems that have been removed from service.

  13. 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

    SciTech Connect

    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 {sup 90}Sr, {sup 137}Cs, and {sup 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.

  14. 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

    SciTech Connect

    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 [sup 90]Sr, [sup 137]Cs, and [sup 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.

  15. Vitrification of high-level radioactive waste in a pilot-scale liquid-fed ceramic melter

    SciTech Connect

    Bjorklund, W.J.; Holton, L.K.; Knowlton, D.E.

    1985-03-01

    Under the sponsorship of the Nuclear Waste Treatment Program (NWTP), a high-level radioactive waste vitrification system has been installed in a Radiochemical Engineering Facility at Hanford, Washington. The facility is designed for remote operation of equipment using master-slave manipulators and overhead cranes. The pilot-scale radioactive system consists of a melter, canister handling turntable, glass-level detection system and supporting waste preparation, offgas treatment and condensate treatment systems. The vitrification system will be operated with radioactive wastes to test remote equipment features and process performance. Destructive and nondestructive methods will be used in the characterization of the waste glass product after canisters are filled with specific waste compositions. 4 refs., 9 figs.

  16. Effect of temperature on the durability of class C fly ash belite cement in simulated radioactive liquid waste: synergy of chloride and sulphate ions.

    PubMed

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

    2009-06-15

    The durability of class C fly ash belite cement (FABC-2-W) in simulated radioactive liquid waste (SRLW) rich in a mixed sodium chloride and sulphate solution is presented here. The effect of the temperature and potential synergic effect of chloride and sulfate ions are discussed. This study has been carried out according to the Koch-Steinegger test, at the temperature of 20 degrees C and 40 degrees C during a period of 180 days. The durability has been evaluated by the changes of the flexural strength of mortar, fabricated with this cement, immersed in a simulated radioactive liquid waste rich in sulfate (0.5M), chloride (0.5M) and sodium (1.5M) ions--catalogued like severely aggressive for the traditional Portland cement--and demineralised water, which was used as reference. The reaction mechanism of sulphate, chloride and sodium ions with the mortar was evaluated by scanning electron microscopy (SEM), porosity and pore-size distribution, and X-ray diffraction (XRD). The results showed that the chloride binding and formation of Friedel's salt was inhibited by the presence of sulphate. Sulphate ion reacts preferentially with the calcium aluminate hydrates forming non-expansive ettringite which precipitated inside the pores; the microstructure was refined and the mechanical properties enhanced. This process was faster and more marked at 40 degrees C. PMID:19056176

  17. Vapor-liquid partitioning of alkaline earth and transition metals in NaCl-dominated hydrothermal fluids: An experimental study from 360 to 465 °C, near-critical to halite saturated conditions

    NASA Astrophysics Data System (ADS)

    Pester, Nicholas J.; Ding, Kang; Seyfried, William E.

    2015-11-01

    Multi-phase fluid flow is a common occurrence in magmatic hydrothermal systems; and extensive modeling efforts using currently established P-V-T-x properties of the NaCl-H2O system are impending. We have therefore performed hydrothermal flow experiments (360-465 °C) to observe vapor-liquid partitioning of alkaline earth and first row transition metals in NaCl-dominated source solutions. The data allow extraction of partition coefficients related to the intrinsic changes in both chlorinity and density along the two-phase solvus. The coefficients yield an overall decrease in vapor affinity in the order Cu(I) > Na > Fe(II) > Zn > Ni(II) ⩾ Mg ⩾ Mn(II) > Co(II) > Ca > Sr > Ba, distinguished with 95% confidence for vapor densities greater than ∼0.2 g/cm3. The alkaline earth metals are limited to purely electrostatic interactions with Cl ligands, resulting in an excellent linear correlation (R2 > 0.99) between their partition coefficients and respective ionic radii. Though broadly consistent with this relationship, relative behavior of the transition metals is not well resolved, being likely obscured by complex bonding processes and the potential participation of Na in the formation of tetra-chloro species. At lower densities (at/near halite saturation) partitioning behavior of all metals becomes highly non-linear, where M/Cl ratios in the vapor begin to increase despite continued decreases in chlorinity and density. We refer to this phenomenon as "volatility", which is broadly associated with substantial increases in the HCl/NaCl ratio (eventually to >1) due to hydrolysis of NaCl. Some transition metals (e.g., Fe, Zn) exhibit volatility prior to halite stability, suggesting a potential shift in vapor speciation relative to nearer critical regions of the vapor-liquid solvus. The chemistry of deep-sea hydrothermal fluids appears affected by this process during magmatic events, however, our results do not support suggestions of subseafloor halite precipitation

  18. Determination of the specific radioactivity of fatty acids separated as their methyl esters by gas-liquid chromatography.

    PubMed

    Bishop, C; Glascock, R F; Newell, E M; Welch, V A

    1971-11-01

    Free or combined (3)H-labeled fatty acids are converted to their methyl-(14)C esters or, if labeled with (14)C, to their methyl-(3)H esters. For a given specific radioactivity of the methyl group, the nuclide ratio in the esters separated by GLC is a direct measure of the specific radioactivity of the fatty acids, and quantitative collection is unnecessary. Methods of methylation with minimum quantities of labeled methanol, and of deriving nuclide ratios from channel ratios in a scintillation spectrometer, are given. PMID:5124543

  19. The nitrate to ammonia and ceramic (NAC) process for the denitration and immobilization of low-level radioactive liquid waste (LLW)

    NASA Astrophysics Data System (ADS)

    Muguercia, Ivan

    Hazardous radioactive liquid waste is the legacy of more than 50 years of plutonium production associated with the United States' nuclear weapons program. It is estimated that more than 245,000 tons of nitrate wastes are stored at facilities such as the single-shell tanks (SST) at the Hanford Site in the state of Washington, and the Melton Valley storage tanks at Oak Ridge National Laboratory (ORNL) in Tennessee. In order to develop an innovative, new technology for the destruction and immobilization of nitrate-based radioactive liquid waste, the United State Department of Energy (DOE) initiated the research project which resulted in the technology known as the Nitrate to Ammonia and Ceramic (NAC) process. However, inasmuch as the nitrate anion is highly mobile and difficult to immobilize, especially in relatively porous cement-based grout which has been used to date as a method for the immobilization of liquid waste, it presents a major obstacle to environmental clean-up initiatives. Thus, in an effort to contribute to the existing body of knowledge and enhance the efficacy of the NAC process, this research involved the experimental measurement of the rheological and heat transfer behaviors of the NAC product slurry and the determination of the optimal operating parameters for the continuous NAC chemical reaction process. Test results indicate that the NAC product slurry exhibits a typical non-Newtonian flow behavior. Correlation equations for the slurry's rheological properties and heat transfer rate in a pipe flow have been developed; these should prove valuable in the design of a full-scale NAC processing plant. The 20-percent slurry exhibited a typical dilatant (shear thickening) behavior and was in the turbulent flow regime due to its lower viscosity. The 40-percent slurry exhibited a typical pseudoplastic (shear thinning) behavior and remained in the laminar flow regime throughout its experimental range. The reactions were found to be more efficient in the

  20. Charge Compensation in RE3+ (RE = Eu, Gd) and M+ (M = Li, Na, K) Co-Doped Alkaline Earth Nanofluorides Obtained by Microwave Reaction with Reactive Ionic Liquids Leading to Improved Optical Properties

    SciTech Connect

    Lorbeer, C; Behrends, F; Cybinska, J; Eckert, H; Mudring, Anja -V

    2014-01-01

    Alkaline earth fluorides are extraordinarily promising host matrices for phosphor materials with regard to rare earth doping. In particular, quantum cutting materials, which might considerably enhance the efficiency of mercury-free fluorescent lamps or SC solar cells, are often based on rare earth containing crystalline fluorides such as NaGdF4, GdF3 or LaF3. Substituting most of the precious rare earth ions and simultaneously retaining the efficiency of the phosphor is a major goal. Alkaline earth fluoride nanoparticles doped with trivalent lanthanide ions (which are required for the quantum cutting phenomenon) were prepared via a microwave assisted method in ionic liquids. As doping trivalent ions into a host with divalent cations requires charge compensation, this effect was thoroughly studied by powder X-ray and electron diffraction, luminescence spectroscopy and 23Na, 139La and 19F solid state NMR spectroscopy. Monovalent alkali ions were codoped with the trivalent lanthanide ions to relieve stress and achieve a better crystallinity and higher quantum cutting abilities of the prepared material. 19F-magic angle spinning (MAS)-NMR-spectra, assisted by 19F{23Na} rotational echo double resonance (REDOR) studies, reveal distinct local fluoride environments, the populations of which are discussed in relation to spatial distribution and clustering models. In the co-doped samples, fluoride species having both Na+ and La3+ ions within their coordination sphere can be identified and quantified. This interplay of mono- and trivalent ions in the CaF2 lattice appears to be an efficient charge compensation mechanism that allows for improved performance characteristics of such co-doped phosphor materials.

  1. Phosphatidylinositol anchor of HeLa cell alkaline phosphatase

    SciTech Connect

    Jemmerson, R.; Low, M.G.

    1987-09-08

    Alkaline phosphatase from cancer cells, HeLa TCRC-1, was biosynthetically labeled with either /sup 3/H-fatty acids or (/sup 3/H)ethanolamine as analyzed by sodium dodecyl sulfate-polyacrylamide gel electrophoresis and fluorography of immunoprecipitated material. Phosphatidylinositol-specific phospholipase C (PI-PLC) released a substantial proportion of the /sup 3/H-fatty acid label from immunoaffinity-purified alkaline phosphatase but had no effect on the radioactivity of (/sup 3/H)ethanolamine-labeled material. PI-PLC also liberated catalytically active alkaline phosphatase from viable cells, and this could be selectively blocked by monoclonal antibodies to alkaline phosphatase. However, the alkaline phosphatase released from /sup 3/H-fatty acid labeled cells by PI-PLC was not radioactive. By contrast, treatment with bromelain removed both the /sup 3/H-fatty acid and the (/sup 3/H)ethanolamine label from purified alkaline phosphatase. Subtilisin was also able to remove the (/sup 3/H)ethanolamine label from the purified alkaline phosphatase. The /sup 3/H radioactivity in alkaline phosphatase purified from (/sup 3/H)ethanolamine-labeled cells comigrated with authentic (/sup 3/H)ethanolamine by anion-exchange chromatography after acid hydrolysis. The data suggest that the /sup 3/H-fatty acid and (/sup 3/H)ethanolamine are covalently attached to the carboxyl-terminal segment since bromelain and subtilisin both release alkaline phosphatase from the membrane by cleavage at that end of the polypeptide chain. The data are consistent with findings for other proteins recently shown to be anchored in the membrane through a glycosylphosphatidylinositol structure and indicate that a similar structure contributes to the membrane anchoring of alkaline phosphatase.

  2. Radioactive waste disposal package

    DOEpatents

    Lampe, Robert F.

    1986-11-04

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

  3. Radioactive waste disposal package

    DOEpatents

    Lampe, Robert F.

    1986-01-01

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

  4. Design and synthesis of redox-switched lariat ethers and their application for transport of alkali and alkaline-Earth metal cations across supported liquid membrane.

    PubMed

    Awasthy, Anubhuti; Bhatnagar, Mamta; Tomar, Jyoti; Sharma, Uma

    2006-01-01

    A new class of redox-switched anthraquinone derived lariat ethers 1-(1-anthraquinonyloxy) 3, 6, 9 trioxaundecane 11-ol (M(1)), 1-(1-anthraquinonyloxy) 3, 6 dioxaoctane 9-ol (M(2)), 1-(1-anthraquinonyloxy) 3 oxapentane 5-ol (M(3)), 1-(1-anthraquinonyloxy) 3 oxapentane 5-butane (M(4)), 1-(1-anthraquinonyloxy) 3, 6 dioxaoctane 9-methane (M(5)) and 1-(1-anthraquinonyloxy) 3 oxapentane 5-methane (M(6)) have been synthesized and characterized by spectral analysis. These ionophores were used in liquid membrane carrier facilitated transport of main group metal cations across supported liquid membrane (SLM). Cellulose nitrate membrane was used as membrane support. Effect of various parameters such as variation in concentration of metal as well as ionophore, effect of chain length and end group of ionophore have been studied. The sequence of metal ions transported by ionophore M(1) is Na(+) > Li(+) > K(+) > Ca(2+) > Mg(2+) and the order of metal ions transported by ionophores (M(2)-M(6)) is Li(+) > Na(+) > K(+) > Ca(2+) > Mg(2+). Ionophore M(1) is selective for Na(+), Li(+), and K(+) and ionophores (M(2)-M(6)) are selective for Li(+) and Na(+). PMID:17497021

  5. Technology for Treatment of Liquid Radioactive Waste Generated during Uranium and Plutonium Chemical and Metallurgical Manufacturing in FSUE PO Mayak - 13616

    SciTech Connect

    Adamovich, D.

    2013-07-01

    Created technological scheme for treatment of liquid radioactive waste generated while uranium and plutonium chemical and metallurgical manufacturing consists of: - Liquid radioactive waste (LRW) purification from radionuclides and its transfer into category of manufacturing waste; - Concentration of suspensions containing alpha-nuclides and their further conversion to safe dry state (calcinate) and moving to long controlled storage. The following technologies are implemented in LRW treatment complex: - Settling and filtering technology for treatment of liquid intermediate-level waste (ILW) with volume about 1500m{sup 3}/year and alpha-activity from 10{sup 6} to 10{sup 8} Bq/dm{sup 3} - Membrane and sorption technology for processing of low-level waste (LLW) of radioactive drain waters with volume about 150 000 m{sup 3}/year and alpha-activity from 10{sup 3} to 10{sup 4} Bq/dm{sup 3}. Settling and filtering technology includes two stages of ILW immobilization accompanied with primary settling of radionuclides on transition metal hydroxides with the following flushing and drying of the pulp generated; secondary deep after settling of radionuclides on transition metal hydroxides with the following solid phase concentration by the method of tangential flow ultrafiltration. Besides, the installation capacity on permeate is not less than 3 m{sup 3}/h. Concentrates generated are sent to calcination on microwave drying (MW drying) unit. Membrane and sorption technology includes processing of averaged sewage flux by the method of tangential flow ultrafiltration with total capacity of installations on permeate not less than 18 m{sup 3}/h and sorption extraction of uranium from permeate on anionite. According to radionuclide contamination level purified solution refers to general industrial waste. Concentrates generated during suspension filtering are evaporated in rotary film evaporator (RFE) in order to remove excess water, thereafter they are dried on infrared heating

  6. Anodes for alkaline electrolysis

    DOEpatents

    Soloveichik, Grigorii Lev

    2011-02-01

    A method of making an anode for alkaline electrolysis cells includes adsorption of precursor material on a carbonaceous material, conversion of the precursor material to hydroxide form and conversion of precursor material from hydroxide form to oxy-hydroxide form within the alkaline electrolysis cell.

  7. Alkaline "Permanent" Paper.

    ERIC Educational Resources Information Center

    Pacey, Antony

    1991-01-01

    Discussion of paper manufacturing processes and their effects on library materials focuses on the promotion of alkaline "permanent" paper, with less acid, by Canadian library preservation specialists. Standards for paper acidity are explained; advantages of alkaline paper are described, including decreased manufacturing costs; and recyclability is…

  8. Alkaline igneous rocks

    SciTech Connect

    Fitton, J.G.; Upton, B.G.J.

    1987-01-01

    In this volume, an international team of scientists provides an up-to-date overview of the nature, origin, and evolution of alkaline magmas. Particular attention is paid to carbonatites, lamprophyres, and lamproites which are rock suites of current interest not recently reviewed elsewhere. Recent work on the classical alkaline provinces of East Africa, South Greenland, and the Kola Peninsula is included together with reviews of other areas of alkaline magmatism in North and South America, East Greenland, Europe, West Africa, and the ocean basins. Other papers discuss the impact of experimental isotopic and geochemical studies of the petrogenesis of alkaline rocks. This book will be of interest to petrologists and geochemists studying alkaline igneous rocks, and to other earth scientists as a reference on the rapidly expanding field of igneous petrology.

  9. Structural Integrity Program for the 300,000-Gallon Radioactive Liquid Waste Tanks at the Idaho Nuclear Technology and Engineering Center

    SciTech Connect

    Bryant, Jeffrey Whealdon; Nenni, Joseph A; Timothy S. Yoder

    2003-04-01

    This report provides a record of the Structural Integrity Program for the 300,000-gal liquid waste storage tanks and associated equipment at the Idaho Nuclear Technology and Engineering Center, as required by U.S. Department of Energy M 435.1-1, “Radioactive Waste Management Manual.” This equipment is known collectively as the Tank Farm Facility. The conclusion of this report is that the Tank Farm Facility tanks, vaults, and transfer systems that remain in service for storage are structurally adequate, and are expected to remain structurally adequate over the remainder of their planned service life through 2012. Recommendations are provided for continued monitoring of the Tank Farm Facility.

  10. Structural Integrity Program for the 300,000-Gallon Radioactive Liquid Waste Storage Tanks at the Idaho Nuclear Technology and Engineering Center

    SciTech Connect

    Bryant, Jeffrey W.

    2010-08-12

    This report provides a record of the Structural Integrity Program for the 300,000-gal liquid waste storage tanks and associated equipment at the Idaho Nuclear Technology and Engineering Center, as required by U.S. Department of Energy M 435.1-1, “Radioactive Waste Management Manual.” This equipment is known collectively as the Tank Farm Facility. This report is an update, and replaces the previous report by the same title issued April 2003. The conclusion of this report is that the Tank Farm Facility tanks, vaults, and transfer systems that remain in service for storage are structurally adequate, and are expected to remain structurally adequate over the remainder of their planned service life through 2012. Recommendations are provided for continued monitoring of the Tank Farm Facility.

  11. 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

    SciTech Connect

    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.

  12. Feasibility study on the solidification of liquid low-level radioactive mixed waste in the inactive tank system at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    SciTech Connect

    Trussell, S. . Dept. of Civil Engineering); Spence, R.D. )

    1993-01-01

    A literature survey was conducted to help determine the feasibility of solidifying a liquid low-level radioactive mixed waste in the inactive tank system at Oak Ridge National Laboratory (ORNL). The goal of this report is to facilitate a decision on the disposition of these wastes by identifying any waste constituents that might (1) compromise the strength or stability of the waste form or (2) be highly leachable. Furthermore, its goal is to identify ways to circumvent interferences and to decrease the leachability of the waste constituents. This study has sought to provide an understanding of inhibition of cement set by identifying the fundamental chemical mechanisms by which this inhibition takes place. From this fundamental information, it is possible to draw some conclusions about the potential effects of waste constituents, even in the absence of particular studies on specific compounds.

  13. Implementation plan for liquid low-level radioactive waste systems under the FFA for Fiscal years 1996 and 1997 at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    SciTech Connect

    1996-10-01

    The Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) requires a Federal Facility Agreement (FFA) for federal facilities placed on the National Priorities List. The Oak Ridge Reservation was placed on that list on December 21, 1989, and the agreement was signed in November 1991 by the Department of Energy Oak Ridge Operations Office (DOE-ORO), the U.S. Environmental Protection Agency (EPA)-Region IV, and the Tennessee Department of Environment and Conservation (TDEC). The effective date of the FFA was January 1, 1992. Section IX and Appendix F of the agreement impose design and operating requirements on the Oak Ridge National Laboratory (ORNL) liquid low-level radioactive waste (LLLW) tank systems and identify several plans, schedules, and assessments that must be submitted to EPA/TDEC for review of approval. The issue of ES/ER-17&D1 Federal Facility Agreement Plans and Schedules for Liquid Low-Level Radioactive Waste Tank Systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee in March 1992 transmitted to EPA/TDEC those plans and schedules that were required within 60 to 90 days of the FFA effective date. This document updates the plans, schedules, and strategy for achieving compliance with the FFA as presented in ES/ER-17&D I and summarizes the progress that has been made to date. This document supersedes all updates of ES/ER- 17&D 1. Chapter 1 describes the history and operation of the ORNL LLLW System and the objectives of the FFA. Chapters 2 through 5 contain the updated plans and schedules for meeting FFA requirements. This document will continue to be periodically reassessed and refined to reflect newly developed information and progress.

  14. Chemical Corrosion of Liquid-Phase Sintered SiC in Acidic/Alkaline Solutions Part 1. Corrosion in HNO3 Solution

    NASA Astrophysics Data System (ADS)

    Zhang, Lei; Zhang, Ming; He, Xinnong; Tang, Wenming

    2016-03-01

    The corrosion behavior of the liquid-phase sintered SiC (LPS-SiC) was studied by dipping in 3.53 mol/L HNO3 aqueous solution at room temperature and 70 °C, respectively. The weight loss, strength reduction and morphology evolution of the SiC specimens during corroding were revealed and also the chemical corrosion process and mechanism of the SiC specimens in the acidic solution were clarified. The results show that the corrosion of the LPS-SiC specimens in the HNO3 solution is selective. The SiC particles are almost free from corrosion, but the secondary phases of BaAl2Si2O8 (BAS) and Y2Si2O7 are corroded via an acid-alkali neutralization reaction. BAS has a higher corrosion rate than Y2Si2O7, resulting in the formation of the bamboo-leaf-like corrosion pits. As the SiC specimens etched in the HNO3 solution at room temperature for 75 days, about 80 μm thickness corrosion layer forms. The weight loss and bending strength reduction of the etched SiC specimens are 2.6 mg/cm2 and 52%, respectively. The corrosion of the SiC specimens is accelerated in the 70 °C HNO3 solution with a rate about five times bigger than that in the same corrosion medium at room temperature.

  15. Analysis of Cocoa Proanthocyanidins Using Reversed Phase High-Performance Liquid Chromatography and Electrochemical Detection: Application to Studies on the Effect of Alkaline Processing.

    PubMed

    Stanley, Todd H; Smithson, Andrew T; Neilson, Andrew P; Anantheswaran, Ramaswamy C; Lambert, Joshua D

    2015-07-01

    Flavan-3-ols and proanthocyanidins play a key role in the health beneficial effects of cocoa. Here, we developed a new reversed phased high-performance liquid chromatography-electrochemical detection (HPLC-ECD) method for the analysis of flavan-3-ols and proanthocyanidins of degree of polymerization (DP) 2-7. We used this method to examine the effect of alkalization on polyphenol composition of cocoa powder. Treatment of cocoa powder with NaOH (final pH 8.0) at 92 °C for up to 1 h increased catechin content by 40%, but reduced epicatechin and proanthocyanidins by 23-66%. Proanthocyanidin loss could be modeled using a two-phase exponential decay model (R(2) > 0.7 for epicatchin and proanthocyanidins of odd DP). Alkalization resulted in a significant color change and 20% loss of total polyphenols. The present work demonstrates the first use of HPLC-ECD for the detection of proanthocyanidins up to DP 7 and provides an initial predictive model for the effect of alkali treatment on cocoa polyphenols. PMID:26042917

  16. Radioactive Air Emission Notice of Construction (NOC) for Construction of Liquid Effluent Transfer System Project W-519

    SciTech Connect

    HOMAN, N.A.

    2000-05-01

    The proposed action is to install a new liquid effluent transfer system (three underground waste transfer pipelines). As such, a potential new source will be created as a result of the construction activities. The anticipated emissions associated with this activity are insignificant.

  17. Alkaline battery operational methodology

    DOEpatents

    Sholklapper, Tal; Gallaway, Joshua; Steingart, Daniel; Ingale, Nilesh; Nyce, Michael

    2016-08-16

    Methods of using specific operational charge and discharge parameters to extend the life of alkaline batteries are disclosed. The methods can be used with any commercial primary or secondary alkaline battery, as well as with newer alkaline battery designs, including batteries with flowing electrolyte. The methods include cycling batteries within a narrow operating voltage window, with minimum and maximum cut-off voltages that are set based on battery characteristics and environmental conditions. The narrow voltage window decreases available capacity but allows the batteries to be cycled for hundreds or thousands of times.

  18. Evidence of the generation of isosaccharinic acids and their subsequent degradation by local microbial consortia within hyper-alkaline contaminated soils, with relevance to intermediate level radioactive waste disposal.

    PubMed

    Rout, Simon P; Charles, Christopher J; Garratt, Eva J; Laws, Andrew P; Gunn, John; Humphreys, Paul N

    2015-01-01

    The contamination of surface environments with hydroxide rich wastes leads to the formation of high pH (>11.0) soil profiles. One such site is a legacy lime works at Harpur Hill, Derbyshire where soil profile indicated in-situ pH values up to pH 12. Soil and porewater profiles around the site indicated clear evidence of the presence of the α and β stereoisomers of isosaccharinic acid (ISA) resulting from the anoxic, alkaline degradation of cellulosic material. ISAs are of particular interest with regards to the disposal of cellulosic materials contained within the intermediate level waste (ILW) inventory of the United Kingdom, where they may influence radionuclide mobility via complexation events occurring within a geological disposal facility (GDF) concept. The mixing of uncontaminated soils with the alkaline leachate of the site resulted in ISA generation, where the rate of generation in-situ is likely to be dependent upon the prevailing temperature of the soil. Microbial consortia present in the uncontaminated soil were capable of surviving conditions imposed by the alkaline leachate and demonstrated the ability to utilise ISAs as a carbon source. Leachate-contaminated soil was sub-cultured in a cellulose degradation product driven microcosm operating at pH 11, the consortia present were capable of the degradation of ISAs and the generation of methane from the resultant H2/CO2 produced from fermentation processes. Following microbial community analysis, fermentation processes appear to be predominated by Clostridia from the genus Alkaliphilus sp, with methanogenesis being attributed to Methanobacterium and Methanomassiliicoccus sp. The study is the first to identify the generation of ISA within an anthropogenic environment and advocates the notion that microbial activity within an ILW-GDF is likely to influence the impact of ISAs upon radionuclide migration. PMID:25748643

  19. Evidence of the Generation of Isosaccharinic Acids and Their Subsequent Degradation by Local Microbial Consortia within Hyper-Alkaline Contaminated Soils, with Relevance to Intermediate Level Radioactive Waste Disposal

    PubMed Central

    Rout, Simon P.; Charles, Christopher J.; Garratt, Eva J.; Laws, Andrew P.; Gunn, John; Humphreys, Paul N.

    2015-01-01

    The contamination of surface environments with hydroxide rich wastes leads to the formation of high pH (>11.0) soil profiles. One such site is a legacy lime works at Harpur Hill, Derbyshire where soil profile indicated in-situ pH values up to pH 12. Soil and porewater profiles around the site indicated clear evidence of the presence of the α and β stereoisomers of isosaccharinic acid (ISA) resulting from the anoxic, alkaline degradation of cellulosic material. ISAs are of particular interest with regards to the disposal of cellulosic materials contained within the intermediate level waste (ILW) inventory of the United Kingdom, where they may influence radionuclide mobility via complexation events occurring within a geological disposal facility (GDF) concept. The mixing of uncontaminated soils with the alkaline leachate of the site resulted in ISA generation, where the rate of generation in-situ is likely to be dependent upon the prevailing temperature of the soil. Microbial consortia present in the uncontaminated soil were capable of surviving conditions imposed by the alkaline leachate and demonstrated the ability to utilise ISAs as a carbon source. Leachate-contaminated soil was sub-cultured in a cellulose degradation product driven microcosm operating at pH 11, the consortia present were capable of the degradation of ISAs and the generation of methane from the resultant H2/CO2 produced from fermentation processes. Following microbial community analysis, fermentation processes appear to be predominated by Clostridia from the genus Alkaliphilus sp, with methanogenesis being attributed to Methanobacterium and Methanomassiliicoccus sp. The study is the first to identify the generation of ISA within an anthropogenic environment and advocates the notion that microbial activity within an ILW-GDF is likely to influence the impact of ISAs upon radionuclide migration. PMID:25748643

  20. Implementation plan for liquid low-level radioactive waste tank systems for fiscal year 1995 at Oak Ridge National Laboratory under the Federal Facility Agreement, Oak Ridge, Tennessee

    SciTech Connect

    1995-06-01

    This document is the third annual revision of the plans and schedules for implementing the Federal Facility Agreement (FFA) compliance program, originally submitted in 1992 as ES/ER-17&D1, Federal Facility Agreement Plans and Schedules for Liquid Low-Level Radioactive Waste Tank Systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee. This document summarizes the progress that has been made to date in implementing the plans and schedules for meeting the FFA commitments for the Liquid Low-Level Waste (LLLW) System at Oak Ridge National Laboratory (ORNL). Information presented in this document provides a comprehensive summary to facilitate understanding of the FFA compliance program for LLLW tank systems and to present plans and schedules associated with remediation, through the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) process, of LLLW tank systems that have been removed from service. ORNL has a comprehensive program underway to upgrade the LLLW System as necessary to meet the FFA requirements. The tank systems that are removed from service are being investigated and remediated through the CERCLA process. Waste and risk characterizations have been submitted. Additional data will be prepared and submitted to EPA/TDEC as tanks are taken out of service and as required by the remedial investigation/feasibility study (RI/FS) process. Chapter 1 provides general background information and philosophies that led to the plans and schedules that appear in Chaps. 2 through 5.

  1. Extraction processes and solvents for recovery of cesium, strontium, rare earth elements, technetium and actinides from liquid radioactive waste

    DOEpatents

    Zaitsev, Boris N.; Esimantovskiy, Vyacheslav M.; Lazarev, Leonard N.; Dzekun, Evgeniy G.; Romanovskiy, Valeriy N.; Todd, Terry A.; Brewer, Ken N.; Herbst, Ronald S.; Law, Jack D.

    2001-01-01

    Cesium and strontium are extracted from aqueous acidic radioactive waste containing rare earth elements, technetium and actinides, by contacting the waste with a composition of a complex organoboron compound and polyethylene glycol in an organofluorine diluent mixture. In a preferred embodiment the complex organoboron compound is chlorinated cobalt dicarbollide, the polyethylene glycol has the formula RC.sub.6 H.sub.4 (OCH.sub.2 CH.sub.2).sub.n OH, and the organofluorine diluent is a mixture of bis-tetrafluoropropyl ether of diethylene glycol with at least one of bis-tetrafluoropropyl ether of ethylene glycol and bis-tetrafluoropropyl formal. The rare earths, technetium and the actinides (especially uranium, plutonium and americium), are extracted from the aqueous phase using a phosphine oxide in a hydrocarbon diluent, and reextracted from the resulting organic phase into an aqueous phase by using a suitable strip reagent.

  2. EVALUATION AND SELECTION OF 99TC GETTERS FOR SEQUESTRATION OF LIQUID SECONDARY WASTE RESULTING FROM VITRIFICATION OF RADIOACTIVE WASTE FROM HANFORD

    SciTech Connect

    Mattigod, Shas V.; Westsik, Joseph H.

    2011-03-31

    Getters are most commonly inorganic materials that selectively adsorb radionuclide and metallic contaminants. Typically, these materials have been deployed in two different modes to immobilize and retard contaminant release from monolithic waste forms. One mode is to first use getters to selectively scavenge the radionuclide of interest from a liquid waste stream, and then incorporate the radionuclide-loaded getters in cementitious or other monolithic waste forms. The other mode consists of mixing getters and liquid waste together during formulation of monolithic waste forms. Desirable characteristics for a getter material include, (1) specific adsorption of radionuclide of interest and very high selectivity toward radionuclides of concern in concentrations that would be several orders of magnitude less than the concentrations of competing anions and cations, (2) adsorption capacity that should be sufficient for the mass and volume of the material that will be deployed to be within practicable limits, (3) long-term adsorption and retention of radionuclide, (4) sufficient physical and chemical stability that its radionuclide retention performance will not degrade significantly during the designed life span of the waste form, (5) chemical stability under the range of Eh, pH, and solution conditions that exist in the waste form environment, and (6) should not adversely affect chemical and physical integrity of waste forms. We conducted a literature review to identify getters that are suitable for effectively sequestering 99Tc in monolithic waste forms that are being evaluated for stabilizing secondary liquid waste streams resulting from treatment and vitrification of radioactive tank wastes at Hanford. As a result of this review, we identified a set of getters that warrant further evaluation for this specific application.

  3. First results of in-can microwave processing experiments for radioactive liquid wastes at the Oak Ridge National Laboratory

    SciTech Connect

    White, T.L.; Youngblood, E.L.; Berry, J.B.; Mattus, A.J.

    1990-01-01

    The Oak Ridge National Laboratory (ORNL) Waste Handling and Packaging Plant is developing a microwave process to reduce and solidify remote-handled transuranic (RH-TRU) liquids and sludges presently stored in large tanks at ORNL. Testing has recently begun on an in drum microwave process using nonradioactive RH-TRU surrogates. The microwave process development effort has focused on an in-drum process to dry the RH-TRU liquids and sludges in the final storage container and then melt the salt residues to form a solid monolith. A 1/3-scale proprietary microwave applicator was designed, fabricated, and tested to demonstrate the essential features of the microwave design and to provide input into the design of the full-scale applicator. Conductivity cell measurements suggest that the microwave energy heats near the surface of the surrogate over a wide range of temperatures. The final wasteform meets the waste acceptance criteria for the Waste Isolation Pilot Plant, a federal repository for defense transuranic wastes near Carlsbad, New Mexico. 7 refs., 3 figs., 1 tab.

  4. Radioactive by-products of a self-shielded cyclotron and the liquid target system for F-18 routine production.

    PubMed

    Kambali, I; Suryanto, H; Parwanto

    2016-06-01

    Routine production of F-18 radionuclide using proton beams accelerated in a cyclotron could potentially generate residual radioisotopes in the cyclotron vicinity which eventually become major safety concerns over radiation exposure to the workers. In this investigation, a typical 11-MeV proton, self-shielded cyclotron has been assessed for its residual radiation sources in the cyclotron's shielding, tank/chamber, cave wall as well as target system. Using a portable gamma ray spectroscopy system, the radiation measurement in the cyclotron environment has been carried out. Experimental results indicate that relatively long-lived radioisotopes such as Mn-54, Zn-65 and Eu-152 are detected in the inner and outer surface of the cyclotron shielding respectively while Mn-54 spectrum is observed around the cyclotron chamber. Weak intensity of Eu-152 radioisotope is again spotted in the inner and outer surface of the cyclotron cave wall. Angular distribution measurement of the Eu-152 shows that the intensity slightly drops with increasing observation angle relative to the proton beam incoming angle. In the target system, gamma rays from Co-56, Mn-52, Co-60, Mn-54, Ag-110 m are identified. TALYS-calculated nuclear cross-section data are used to study the origins of the radioactive by-products. PMID:26867652

  5. Copper Ferrocyanide Functionalized Core-Shell Magnetic Silica Composites for the Selective Removal of Cesium Ions from Radioactive Liquid Waste.

    PubMed

    Lee, Hyun Kyu; Yang, Da Som; Oh, Wonzin; Choi, Sang-June

    2016-06-01

    The copper ferrocyanide functionalized core-shell magnetic silica composite (mag@silica-CuFC) was prepared and was found to be easily separated from aqueous solutions by using magnetic field. The synthesized mag@silica-CuFC composite has a high sorption ability of Cs owing to its strong affinity for Cs as well as the high surface area of the supports. Cs sorption on the mag@silica-CuFC composite quickly reached the sorption equilibrium after 2 h of contact time. The effect of the presence of salts with a high concentration of up to 3.5 wt% on the efficiency of Cs sorption onto the composites was also studied. The maximum sorption ability was found to be maintained in the presence of up to 3.5 wt% of NaCl in the solution. Considering these results, the mag@silica-CuFC composite has great potential for use as an effective sorbent for the selective removal of radioactive Cs ions. PMID:27427694

  6. Efficiency of inductively torch plasma operating at atmospheric pressure on destruction of chlorinated liquid wastes- A path to the treatment of radioactive organic halogen liquid wastes

    NASA Astrophysics Data System (ADS)

    Kamgang-Youbi, G.; Poizot, K.; Lemont, F.

    2012-12-01

    The performance of a plasma reactor for the degradation of chlorinated hydrocarbon waste is reported. Chloroform was used as a target for a recently patented destruction process based using an inductive plasma torch. Liquid waste was directly injected axially into the argon plasma with a supplied power of ~4 kW in the presence of oxygen as oxidant and carrier gas. Decomposition was performed at CHCl3 feed rates up to 400 g·h-1 with different oxygen/waste molar ratios, chloroform destruction was obtained with at least 99% efficiency and the energy efficiency reached 100 g·kWh-1. The conversion end products were identified and assayed by online FTIR spectroscopy (CO2, HCl and H2O) and redox titration (Cl2). Considering phosgene as representative of toxic compounds, only very small quantities of toxics were released (< 1 g·h-1) even with high waste feed rates. The experimental results were very close to the equilibrium composition predicted by thermodynamic calculations. At the bottom of the reactor, the chlorinated acids were successfully trapped in a scrubber and transformed into mineral salts, hence, only CO2 and H2O have been found in the final off-gases composition.

  7. The alkaline earth intercalates of molybdenum disulfide

    NASA Technical Reports Server (NTRS)

    Somoano, R. B.; Hadek, V.; Rembaum, A.; Samson, S.; Woollam, J. A.

    1975-01-01

    Molybdenum disulfide has been intercalated with calcium and strontium by means of the liquid ammonia technique. Chemical, X-ray, and superconductivity data are presented. The X-ray data reveal a lowering of crystal symmetry and increase of complexity of the structure upon intercalation with the alkaline earth metals. The Ca and Sr intercalates start to superconduct at 4 and 5.6 K, respectively, and show considerable anisotropy regarding the critical magnetic field.

  8. Federal Facility Agreement plans and schedules for liquid low-level radioactive waste tank systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee. Environmental Restoration Program

    SciTech Connect

    Not Available

    1992-03-01

    Although the Federal Facility Agreement (FFA) addresses the entire Oak Ridge Reservation, specific requirements are set forth for the liquid low-level radioactive waste (LLLW) storage tanks and their associated piping and equipment, tank systems, at ORNL. The stated objected of the FFA as it relates to these tank systems is to ensure that structural integrity, containment and detection of releases, and source control are maintained pending final remedial action at the site. The FFA requires that leaking LLLW tank systems be immediately removed from service. It also requires the LLLW tank systems that do not meet the design and performance requirements established for secondary containment and leak detection be either upgraded or replaced. The FFA establishes a procedural framework for implementing the environmental laws. For the LLLW tank systems, this framework requires the specified plans and schedules be submitted to EPA and TDEC for approval within 60 days, or in some cases, within 90 days, of the effective date of the agreement.

  9. Federal Facility Agreement plans and schedules for liquid low-level radioactive waste tank systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    SciTech Connect

    Not Available

    1992-03-01

    Although the Federal Facility Agreement (FFA) addresses the entire Oak Ridge Reservation, specific requirements are set forth for the liquid low-level radioactive waste (LLLW) storage tanks and their associated piping and equipment, tank systems, at ORNL. The stated objected of the FFA as it relates to these tank systems is to ensure that structural integrity, containment and detection of releases, and source control are maintained pending final remedial action at the site. The FFA requires that leaking LLLW tank systems be immediately removed from service. It also requires the LLLW tank systems that do not meet the design and performance requirements established for secondary containment and leak detection be either upgraded or replaced. The FFA establishes a procedural framework for implementing the environmental laws. For the LLLW tank systems, this framework requires the specified plans and schedules be submitted to EPA and TDEC for approval within 60 days, or in some cases, within 90 days, of the effective date of the agreement.

  10. Evaluation of plasma melter technology for verification of high-sodium content low-level radioactive liquid wastes: Demonstration test No. 4 preliminary test report

    SciTech Connect

    McLaughlin, D.F.; Gass, W.R.; Dighe, S.V.; D`Amico, N.; Swensrud, R.L.; Darr, M.F.

    1995-01-10

    This document provides a preliminary report of plasma arc vitrification testing by a vendor in support of the Hanford Tank Waste Remediation System Low-Level Waste (LLW) Vitrification Program. Phase I test conduct included 26 hours (24 hours steady state) of melting of simulated high-sodium low-level radioactive liquid waste. Average processing rate was 4.9 kg/min (peak rate 6.2 kg/min), producing 7330 kg glass product. Free-flowing glass pour point was 1250 C, and power input averaged 1530 kW(e), for a total energy consumption of 19,800 kJ/kg glass. Restart capability was demonstrated following a 40-min outage involving the scrubber liquor heat exchanger, and glass production was continued for another 2 hours. Some volatility losses were apparent, probably in the form of sodium borates. Roughly 275 samples were collected and forwarded for analysis. Sufficient process data were collected for heat/material balances. Recommendations for future work include lower boron contents and improved tuyere design/operation.

  11. HUMAN MACHINE INTERFACE (HMI) EVALUATION OF ROOMS TA-50-1-60/60A AT THE RADIOACTIVE LIQUID WASTE TREATMENT FACILITY (RLWTF)

    SciTech Connect

    Gilmore, Walter E.; Stender, Kerith K.

    2012-08-29

    This effort addressed an evaluation of human machine interfaces (HMIs) in Room TA-50-1-60/60A of the Radioactive Liquid Waste Treatment Facility (RLWTF). The evaluation was performed in accordance with guidance outlined in DOE-STD-3009, DOE Standard Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses, 2006 [DOE 2006]. Specifically, Chapter 13 of DOE 2006 highlights the 10 CFR 830, Nuclear Safety Management, 2012, [CFR 2012] and DOE G 421.1-2 [DOE 2001a] requirements as they relate to the human factors process and, in this case, the safety of the RLWTF. The RLWTF is a Hazard Category 3 facility and, consequently, does not have safety-class (SSCs). However, safety-significant SSCs are identified. The transuranic (TRU) wastewater tanks and associated piping are the only safety-significant SSCs in Rooms TA-50-1-60/60A [LANL 2010]. Hence, the human factors evaluation described herein is only applicable to this particular assemblage of tanks and piping.

  12. Designing a New Radioactive Liquid Waste Treatment Facility for Los Alamos National Laboratory to Update Treatment Technologies and Meet Current Regulatory Requirements

    SciTech Connect

    Lucas, M.E.; Evans, K.M.

    2008-07-01

    This paper discusses the new Radioactive Liquid Waste Treatment Facility (RLWTF) being designed for Los Alamos National Laboratory (LANL) and also provides information regarding the updates to the facility resulting from improvements in technologies and changes in regulations and codes that have occurred since the existing facility was built. It explains how the new facility will be a suitable replacement for the existing facility and how the new facility will allow LANL to continue its current mission as well as have the capability to treat any new waste stream resulting from a future expansion of the laboratory. Specific topics relating to the improvements in technology that are discussed include replacing existing settling technologies with improved chemical precipitation technologies using hydrochloric acid, sodium hydroxide, magnesium chloride, and ferric chloride. Also discussed are improved filtration technologies and dewatering technologies. Topics discussed regarding regulatory and code requirements include the current state and federal regulatory criteria and the new seismic criteria requirements that the design conforms to. (authors)

  13. [Comparative characteristic of some parameters of the surfactant pulmonary system in children-residents of radioactive contaminated territories and children born to participants of liquidation of Chernobyl accident consequences].

    PubMed

    Parkhomenko, V M; Kolpakov, I Ie; Briuzhina, T S; Shumeĭko, V M

    2008-01-01

    The study of lipid fat acid content in condensate of expired air with the help of the gas liquid chromatography revealed the increased sum of polyunsaturated fatty acids (PUFA) in a large measure of increased content of arachidonic and linolic PUFA in children-residents of radioactive contaminated territories in comparison with children born by participants of liquidation of Chernobyl accident consequences. It is evidence of more active processes of lipid peroxidation and shift of the balance of fatty acids to the side of omega-6 family. PMID:19253733

  14. Alkaline flooding injection strategy

    SciTech Connect

    French, T.R.; Josephson, C.B.

    1992-03-01

    The objective of this project is to improved alkali-surfactant flooding methods, and this includes determining the proper design of injection strategy. Several different injection strategies have been used or suggested for recovering heavy oils with surfactant-enhanced alkaline flooding methods. Oil recovery was compared for four different injection strategies: (1) surfactant followed by polymer, (2) surfactant followed by alkaline polymer, (3) alkaline surfactant followed by polymer, and (4) alkali, surfactant, and polymer mixed in a single formulation. The effect of alkaline preflush was also studied under two different conditions. All of the oil recovery experiments were conducted under optimal conditions with a viscous, non-acidic oil from Hepler (KS) oil field. The coreflood experiments were conducted with Berea sandstone cores since field core was not available in sufficient quantity for coreflood tests. The Tucker sand of Hepler field is a Class I fluvial dominated deltaic reservoir, as classified by the Department of Energy, which has been selected as the site of a DOE-sponsored field pilot test.

  15. Evaluation and ranking of the tank focus area solid liquid separation needs

    SciTech Connect

    McCabe, D.J.

    1995-08-17

    The Tank Focus Area (TFA) of the Department of Energy (DOE) Office of Environmental Restoration and Waste Management (EM) addresses remediation of liquid waste currently stored in underground tanks. Several baseline technologies for treatment of tank waste can be categorized into three types of solid liquid separation: (a) removal of radioactive species that have been absorbed or precipitated, (b) pretreatment for ion exchange, and (c) volume reduction of sludge and wash water. The solids formed from precipitation or absorption of radioactive ions require separation from the liquid phase to permit treatment of the liquid as Low Level Waste. Prior to ion exchange of radioactive ions, removal of insoluble solids is needed to prevent bed fouling and downstream contamination. Volume reduction of washed sludge solids would reduce the tank space required for interim storage. The scope of this document is to evaluate the solid/liquid separations needed to permit treatment of tank wastes to accomplish these goals. The document summarizes previous alkaline waste testing, with an emphasis on crossflow filtration, to-obtain a general understanding of the behavior of radioactive wastes on available equipment. The document also provides general information about filtration and a path forward for testing.

  16. Federal Facility Agreement plans and schedules for liquid low-level radioactive waste tank systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    SciTech Connect

    Not Available

    1993-06-01

    The Superfund Amendments and Reauthorization Act of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) requires a Federal Facility Agreement (FFA) for federal facilities placed on the National Priorities List. The Oak Ridge Reservation was placed on that list on December 21, 1989, and the agreement was signed in November 1991 by the Department of Energy Oak Ridge Field Office (DOE-OR), the US Environmental Protection Agency (EPA)-Region IV, and the Tennessee Department of Environment and Conservation (TDEC). The effective date of the FFA was January 1, 1992. Section 9 and Appendix F of the agreement impose design and operating requirements on the Oak Ridge National Laboratory (ORNL) liquid low-level radioactive waste (LLLW) tank systems and identify several plans, schedules, and assessments that must be submitted to EPA/TDEC for review or approval. The initial issue of this document in March 1992 transmitted to EPA/TDEC those plans and schedules that were required within 60 to 90 days of the FFA effective date. The current revision of this document updates the plans, schedules, and strategy for achieving compliance with the FFA, and it summarizes the progress that has been made over the past year. Chapter 1 describes the history and operation of the ORNL LLLW System, the objectives of the FFA, the organization that has been established to bring the system into compliance, and the plans for achieving compliance. Chapters 2 through 7 of this report contain the updated plans and schedules for meeting FFA requirements. This document will continue to be periodically reassessed and refined to reflect newly developed information and progress.

  17. Status of the Oak Ridge National Laboratory new hydrofracture facility: Implications for the disposal of liquid low-level radioactive wastes by underground injection

    SciTech Connect

    Haase, C.S.; Stow, S.H.

    1987-01-01

    From 1982 to 1984, Oak Ridge National Laboratory (ORNL) disposed of approximately 2.8 x 10/sup 16/ Bq (7.5 x 10/sup 5/ Ci) of liquid low-level radioactive wastes by underground injection at its new hydrofracture facility. This paper summarizes the regulatory and operational status of that ORNL facility and discusses its future outlook. Operational developments and regulatory changes that have raised major questions about the continued operation of the new hydrofracture facility include: (1) significant /sup 90/Sr contamination of some groundwater in the injection formation; (2) questions about the design of the injection well, completed prior to the application of the underground injection control (UIC) regulations to the ORNL facility; (3) questions about the integrity of the reconfigured injection well put into service following the loss of the initial injection well; and (4) implementation of UIC regulations. Ultimately, consideration of the regulatory and operational factors led to the decision in early 1986 not to proceed with a UIC permit application for the ORNL facility. Subsequent to the decision not to proceed with a UIC permit application, closure activities were initiated for the ORNL hydrofracture facility. Closure of the facility will occur under both state of Tennessee and federal UIC regulations. The facility also falls under the provisions of part 3004(u) of the Resource Conservation and Recovery Act pertaining to corrective actions. Nationally, there is an uncertain outlook for the disposal of wastes by underground injection. All wells used for the injection of hazardous wastes (Class I wells) are being reviewed. 8 refs., 4 figs., 2 tabs.

  18. RADIOACTIVE CONCENTRATOR AND RADIATION SOURCE

    DOEpatents

    Hatch, L.P.

    1959-12-29

    A method is presented for forming a permeable ion exchange bed using Montmorillonite clay to absorb and adsorb radioactive ions from liquid radioactive wastes. A paste is formed of clay, water, and a material that fomns with clay a stable aggregate in the presence of water. The mixture is extruded into a volume of water to form clay rods. The rods may then be used to remove radioactive cations from liquid waste solutions. After use, the rods are removed from the solution and heated to a temperature of 750 to 1000 deg C to fix the ratioactive cations in the clay.

  19. SELF SINTERING OF RADIOACTIVE WASTES

    DOEpatents

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

    1959-12-29

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

  20. Alkaline Phosphatase in Stem Cells

    PubMed Central

    Štefková, Kateřina; Procházková, Jiřina; Pacherník, Jiří

    2015-01-01

    Alkaline phosphatase is an enzyme commonly expressed in almost all living organisms. In humans and other mammals, determinations of the expression and activity of alkaline phosphatase have frequently been used for cell determination in developmental studies and/or within clinical trials. Alkaline phosphatase also seems to be one of the key markers in the identification of pluripotent embryonic stem as well as related cells. However, alkaline phosphatases exist in some isoenzymes and isoforms, which have tissue specific expressions and functions. Here, the role of alkaline phosphatase as a stem cell marker is discussed in detail. First, we briefly summarize contemporary knowledge of mammalian alkaline phosphatases in general. Second, we focus on the known facts of its role in and potential significance for the identification of stem cells. PMID:25767512

  1. Medium-Sized Mammals around a Radioactive Liquid Waste Lagoon at Los Alamos National Laboratory: Uptake of Contaminants and Evaluation of Radio-Frequency Identification Technology

    SciTech Connect

    Leslie A. Hansen; Phil R. Fresquez; Rhonda J. Robinson; John D. Huchton; Teralene S. Foxx

    1999-11-01

    Use of a radioactive liquid waste lagoon by medium-sized mammals and levels of tritium, other selected radionuclides, and metals in biological tissues of the animals were documented at Technical Area 53 (TA-53) of Los Alamos National Laboratory during 1997 and 1998. Rock squirrel (Spermophilus variegates), raccoon (Procyon lotor), striped skunk (Mephitis mephitis), and bobcat (Lynx rufus) were captured at TA-53 and at a control site on the Santa Fe National Forest. Captured animals were anesthetized and marked with radio-frequency identification (RFD) tags and/or ear tags. We collected urine and hair samples for tritium and metals (aluminum, antimony, arsenic, barium, beryllium, cadmium, chromium, copper, lead, mercury, nickel, selenium, silver, and thallium) analyses, respectively. In addition, muscle and bone samples from two rock squirrels collected from each of TA-53, perimeter, and regional background sites were tested for tritium, {sup 137}Cs, {sup 90}Sr, {sup 238}Pu, {sup 239,240}Pu, {sup 241}Am, and total uranium. Animals at TA-53 were monitored entering and leaving the lagoon area using a RFID monitor to read identification numbers from the RFID tags of marked animals and a separate camera system to photograph all animals passing through the monitor. Cottontail rabbit (Sylvilagus spp.), rock squirrel, and raccoon were the species most frequently photographed going through the RFID monitor. Less than half of all marked animals in the lagoon area were detected using the lagoon. Male and female rock squirrels from the lagoon area had significantly higher tritium concentrations compared to rock squirrels from the control area. Metals tested were not significantly higher in rock squirrels from TA-53, although there was a trend toward increased levels of lead in some individuals at TA-53. Muscle and bone samples from squirrels in the lagoon area appeared to have higher levels of tritium, total uranium, and {sup 137}Cs than samples collected from perimeter and

  2. Alkaline fuel cells applications

    NASA Astrophysics Data System (ADS)

    Kordesch, Karl; Hacker, Viktor; Gsellmann, Josef; Cifrain, Martin; Faleschini, Gottfried; Enzinger, Peter; Fankhauser, Robert; Ortner, Markus; Muhr, Michael; Aronson, Robert R.

    On the world-wide automobile market technical developments are increasingly determined by the dramatic restriction on emissions as well as the regimentation of fuel consumption by legislation. Therefore there is an increasing chance of a completely new technology breakthrough if it offers new opportunities, meeting the requirements of resource preservation and emission restrictions. Fuel cell technology offers the possibility to excel in today's motive power techniques in terms of environmental compatibility, consumer's profit, costs of maintenance and efficiency. The key question is economy. This will be decided by the costs of fuel cell systems if they are to be used as power generators for future electric vehicles. The alkaline hydrogen-air fuel cell system with circulating KOH electrolyte and low-cost catalysed carbon electrodes could be a promising alternative. Based on the experiences of Kordesch [K. Kordesch, Brennstoffbatterien, Springer, Wien, 1984, ISBN 3-387-81819-7; K. Kordesch, City car with H 2-air fuel cell and lead-battery, SAE Paper No. 719015, 6th IECEC, 1971], who operated a city car hybrid vehicle on public roads for 3 years in the early 1970s, improved air electrodes plus new variations of the bipolar stack assembly developed in Graz are investigated. Primary fuel choice will be a major issue until such time as cost-effective, on-board hydrogen storage is developed. Ammonia is an interesting option. The whole system, ammonia dissociator plus alkaline fuel cell (AFC), is characterised by a simple design and high efficiency.

  3. Silica in alkaline brines

    USGS Publications Warehouse

    Jones, B.F.; Rettig, S.L.; Eugster, H.P.

    1967-01-01

    Analysis of sodium carbonate-bicarbonate brines from closed basins in volcanic terranes of Oregon and Kenya reveals silica contents of up to 2700 parts per million at pH's higher than 10. These high concentrations of SiO 2 can be attributed to reaction of waters with silicates, and subsequent evaporative concentration accompanied by a rise in pH. Supersaturation with respect to amorphous silica may occur and persist for brines that are out of contact with silicate muds and undersaturated with respect to trona; correlation of SiO2 with concentration of Na and total CO2 support this interpretation. Addition of moredilute waters to alkaline brines may lower the pH and cause inorganic precipitation of substantial amounts of silica.

  4. Bifunctional alkaline oxygen electrodes

    NASA Technical Reports Server (NTRS)

    Swette, L.; Kackley, N.; Mccatty, S. A.

    1991-01-01

    The authors describe the identification and testing of electrocatalysts and supports for the positive electrode of moderate-temperature, single-unit, rechargeable alkaline fuel cells. Recent work on Na(x)Pt3O4, a potential bifunctional catalyst, is described, as well as the application of novel approaches to the development of more efficient bifunctional electrode structures. The three dual-character electrodes considered here showed similar superior performance; the Pt/RhO2 and Rh/RhO2 electrodes showed slightly better performance than the Pt/IrO2 electrode. It is concluded that Na(x)Pt3O4 continues to be a promising bifunctional oxygen electrode catalyst but requires further investigation and development.

  5. DNA DAMAGE QUANTITATION BY ALKALINE GEL ELECTROPHORESIS.

    SciTech Connect

    SUTHERLAND,B.M.; BENNETT,P.V.; SUTHERLAND, J.C.

    2004-03-24

    Physical and chemical agents in the environment, those used in clinical applications, or encountered during recreational exposures to sunlight, induce damages in DNA. Understanding the biological impact of these agents requires quantitation of the levels of such damages in laboratory test systems as well as in field or clinical samples. Alkaline gel electrophoresis provides a sensitive (down to {approx} a few lesions/5Mb), rapid method of direct quantitation of a wide variety of DNA damages in nanogram quantities of non-radioactive DNAs from laboratory, field, or clinical specimens, including higher plants and animals. This method stems from velocity sedimentation studies of DNA populations, and from the simple methods of agarose gel electrophoresis. Our laboratories have developed quantitative agarose gel methods, analytical descriptions of DNA migration during electrophoresis on agarose gels (1-6), and electronic imaging for accurate determinations of DNA mass (7-9). Although all these components improve sensitivity and throughput of large numbers of samples (7,8,10), a simple version using only standard molecular biology equipment allows routine analysis of DNA damages at moderate frequencies. We present here a description of the methods, as well as a brief description of the underlying principles, required for a simplified approach to quantitation of DNA damages by alkaline gel electrophoresis.

  6. Modulators of intestinal alkaline phosphatase.

    PubMed

    Bobkova, Ekaterina V; Kiffer-Moreira, Tina; Sergienko, Eduard A

    2013-01-01

    Small molecule modulators of phosphatases can lead to clinically useful drugs and serve as invaluable tools to study functional roles of various phosphatases in vivo. Here, we describe lead discovery strategies for identification of inhibitors and activators of intestinal alkaline phosphatases. To identify isozyme-selective inhibitors and activators of the human and mouse intestinal alkaline phosphatases, ultrahigh throughput chemiluminescent assays, utilizing CDP-Star as a substrate, were developed for murine intestinal alkaline phosphatase (mIAP), human intestinal alkaline phosphatase (hIAP), human placental alkaline phosphatase (PLAP), and human tissue-nonspecific alkaline phosphatase (TNAP) isozymes. Using these 1,536-well assays, concurrent HTS screens of the MLSMR library of 323,000 compounds were conducted for human and mouse IAP isozymes monitoring both inhibition and activation. This parallel screening approach led to identification of a novel inhibitory scaffold selective for murine intestinal alkaline phosphatase. SAR efforts based on parallel testing of analogs against different AP isozymes generated a potent inhibitor of the murine IAP with IC50 of 540 nM, at least 65-fold selectivity against human TNAP, and >185 selectivity against human PLAP. PMID:23860652

  7. Alkaline battery, separator therefore

    NASA Technical Reports Server (NTRS)

    Schmidt, George F. (Inventor)

    1980-01-01

    An improved battery separator for alkaline battery cells has low resistance to electrolyte ion transfer and high resistance to electrode ion transfer. The separator is formed by applying an improved coating to an electrolyte absorber. The absorber, preferably, is a flexible, fibrous, and porous substrate that is resistant to strong alkali and oxidation. The coating composition includes an admixture of a polymeric binder, a hydrolyzable polymeric ester and inert fillers. The coating composition is substantially free of reactive fillers and plasticizers commonly employed as porosity promoting agents in separator coatings. When the separator is immersed in electrolyte, the polymeric ester of the film coating reacts with the electrolyte forming a salt and an alcohol. The alcohol goes into solution with the electrolyte while the salt imbibes electrolyte into the coating composition. When the salt is formed, it expands the polymeric chains of the binder to provide a film coating substantially permeable to electrolyte ion transfer but relatively impermeable to electrode ion transfer during use.

  8. Phase 2 report on the evaluation of polyacrylonitrile (PAN) as a binding polymer for absorbers used to treat liquid radioactive wastes

    SciTech Connect

    Sebesta, F.; John, J.; Motl, A.

    1996-05-01

    The performance of PAN-based composite absorbers was evaluated in dynamic experiments at flow rates ranging from 25--100 bed volumes (BV) per hour. Composite absorbers with active components of ammonium molybdophosphate (AMP) PAN and K-Co ferrocyanide (KCoFC) PAN were used for separating Cs from a 1 M HNO{sub 3} + 1 M NaNO{sub 3} + 2 {times} 10{sup {minus}5} M CsCl acidic simulant solution. KCoFC-PAN and two other FC-based composite absorbers were tested for separating Cs from alkaline simulant solutions containing 0.01 M to 1 M NaOH and 1 M NaNO{sub 3} + x {times} 10{sup {minus}4} M CsCl. The efficiency of the Cs sorption on the AMP-PAN absorber from acidic simulant solutions was negatively influenced by the dissolution of the AMP active component. At flow rates of 50 BV/hr, the decontamination factor of about 10{sup 3} could be maintained for treatment of 380 BV of the feed. With the KCoFC-PAN absorber, the decontamination factor of about 10{sup 3} could be maintained for a feed volume as great as 1,800 BV. In alkaline simulant solutions, significant decomposition of the active components was observed, and the best performance was exhibited by the KCoFC-PAN absorber. Introductory experiments confirmed that Cs may be washed out of the composite absorbers. Regeneration of both absorbers for repetitive use was also found to be possible. The main result of the study is that PAN was proven to be a versatile polymer capable of forming porous composite absorbers with a large number of primary absorbers. The composite absorbers proved to be capable of withstanding the harsh acidic and alkaline conditions and significant radiation doses that may be expected in the treatment of US DOE wastes. A field demonstration is proposed as a follow-on activity.

  9. Evaluation of Alkaline Cleaner Materials

    NASA Technical Reports Server (NTRS)

    Partz, Earl

    1998-01-01

    Alkaline cleaners used to process aluminum substrates have contained chromium as the corrosion inhibitor. Chromium is a hazardous substance whose use and control are described by environmental laws. Replacement materials that have the characteristics of chromated alkaline cleaners need to be found that address both the cleaning requirements and environmental impacts. This report will review environmentally friendly candidates evaluated as non-chromium alkaline cleaner replacements and methods used to compare those candidates one versus another. The report will also list characteristics used to select candidates based on their declared contents. It will also describe and evaluate methods used to discriminate among the large number of prospective candidates.

  10. Novel Fission-Product Separation based on Room-Temperature Ionic Liquids

    SciTech Connect

    Rogers, Robin D.

    2004-12-31

    U.S. DOE's underground storage tanks at Hanford, SRS, and INEEL contain liquid wastes with high concentrations of radioactive cesium-137 and strontium-90. Because the primary chemical components of alkaline supernatants are sodium nitrate and sodium hydroxide, the majority of this could be disposed of as low level waste if radioactive cesium-137 and strontium- 90 could be selectively removed. The underlying goal of this project was to investigate the application of ionic liquids as novel solvents for new solvent extraction processes for separation of cesium-137 and strontium-90 from tank wastes. Ionic liquids are a distinct sub-set of liquids, comprising only of cations and anions they are proving to be increasingly interesting fluids for application in systems from electrochemistry to energetic materials, and are also rapidly establishing their promise as viable media for synthesis and separations operations. Properties including low melting points, electrochemical conductivity, wide liquid ranges, lack of vapor-pressure, and chemical tunability have encouraged researchers to explore the uses of ILs in place of volatile organic solvents. The most promising current developments arise from control of the unique combinations of chemical and physical properties characteristic of ionic liquids.

  11. Radioactive Wastes.

    PubMed

    Choudri, B S; Baawain, Mahad

    2015-10-01

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

  12. Radioactive Wastes.

    PubMed

    Choudri, B S; Baawain, Mahad

    2016-10-01

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

  13. Preliminary report on the geology and hydrology of Mortandad Canyon near Los Alamos, New Mexico, with reference to disposal of liquid low-level radioactive waste

    USGS Publications Warehouse

    Baltz, E.H.; Abrahams, J.H., Jr.; Purtyman, W.D.

    1963-01-01

    The U.S. Geological Survey, in cooperation with the U.S. Atomic Energy Commission and the Los Alamos Scientific Laboratory, selected the upper part of Mortandad Canyon near Los Alamos, New Mexico for a site for disposal of treated liquid low-level radioactive waste. This report summarizes the part of a study of the geology and hydrology that was done from October 1960 through June 1961. Additional work is being continued. Mortandad Canyon is a narrow east-southeast-trending canyon about 9? miles long that heads on the central part of the Pajarito Plateau at an altitude of about 7,340 feet. The canyon is tributary to the Rio Grande. The drainage area of the part of Mortandad Canyon that was investigated is about 2 square miles, and the total drainage area is about 4.9 square miles. The Pajarito Plateau is capped by the Bandelier Tuff of Pleistocene age. Mortandad Canyon is cut in the Bandelier, and alluvium covers the floor of the canyon to depths ranging from less than 1 foot to as much as 100 feet. The Bandelier is underlain by silt, sand, conglomerate, and interbedded basalt of the Santa Fe Group of Miocene, Pliocene, and Pleistocene(?) age. Some ground water is perched in the alluvium in the canyon; however, the top of the main aquifer is in the Santa Fe Group at a depth of about 990 feet below the canyon floor. Joints in the Bandelier Tuff probably were caused by shrinkage of the tuff during cooling. The joints range in width from hairline cracks to fissures several inches wide. Water can infiltrate along the open joints where the Bandelier is at the surface; however, soil, alluvial fill, and autochthonous clay inhibit infiltration on the tops of mesas and probably in the alluvium-floored canyons also. Thirty-three test holes, each less than 100 feet deep, were drilled in 10 lies across Mortandad Canyon from the western margin of the study area to just west of the Los Alamos-Santa Fe County line. Ten of the holes were cased for observation wells to measure

  14. Radioactivity Calculations

    ERIC Educational Resources Information Center

    Onega, Ronald J.

    1969-01-01

    Three problems in radioactive buildup and decay are presented and solved. Matrix algebra is used to solve the second problem. The third problem deals with flux depression and is solved by the use of differential equations. (LC)

  15. Simulated Radioactivity

    ERIC Educational Resources Information Center

    Boettler, James L.

    1972-01-01

    Describes the errors in the sugar-cube experiment related to radioactivity as described in Project Physics course. The discussion considers some of the steps overlooked in the experiment and generalizes the theory beyond the sugar-cube stage. (PS)

  16. Concentrating Radioactivity

    ERIC Educational Resources Information Center

    Herrmann, Richard A.

    1974-01-01

    By concentrating radioactivity contained on luminous dials, a teacher can make a high reading source for classroom experiments on radiation. The preparation of the source and its uses are described. (DT)

  17. Feasibility study on the solidification of liquid low-level radioactive mixed waste in the inactive tank system at Oak Ridge National Laboratory, Oak Ridge, Tennessee. Environmental Restoration Program

    SciTech Connect

    Trussell, S.; Spence, R.D.

    1993-01-01

    A literature survey was conducted to help determine the feasibility of solidifying a liquid low-level radioactive mixed waste in the inactive tank system at Oak Ridge National Laboratory (ORNL). The goal of this report is to facilitate a decision on the disposition of these wastes by identifying any waste constituents that might (1) compromise the strength or stability of the waste form or (2) be highly leachable. Furthermore, its goal is to identify ways to circumvent interferences and to decrease the leachability of the waste constituents. This study has sought to provide an understanding of inhibition of cement set by identifying the fundamental chemical mechanisms by which this inhibition takes place. From this fundamental information, it is possible to draw some conclusions about the potential effects of waste constituents, even in the absence of particular studies on specific compounds.

  18. Distributions of 14 elements on 60 selected absorbers from two simulant solutions (acid-dissolved sludge and alkaline supernate) for Hanford HLW Tank 102-SY

    SciTech Connect

    Marsh, S.F.; Svitra, Z.V.; Bowen, S.M.

    1993-10-01

    Sixty commercially available or experimental absorber materials were evaluated for partitioning high-level radioactive waste. These absorbers included cation and anion exchange resins, inorganic exchangers, composite absorbers, and a series of liquid extractants sorbed on porous support-beads. The distributions of 14 elements onto each absorber were measured from simulated solutions that represent acid-dissolved sludge and alkaline supernate solutions from Hanford high-level waste (HLW) Tank 102-SY. The selected elements, which represent fission products (Ce, Cs, Sr, Tc, and Y); actinides (U, Pu, and Am); and matrix elements (Cr, Co, Fe, Mn, Zn, and Zr), were traced by radionuclides and assayed by gamma spectrometry. Distribution coefficients for each of the 1680 element/absorber/solution combinations were measured for dynamic contact periods of 30 min, 2 h, and 6 h to provide sorption kinetics information for the specified elements from these complex media. More than 5000 measured distribution coefficients are tabulated.

  19. Reactive transport modelling of the interaction of fission product ground contamination with alkaline and cementitious leachates

    SciTech Connect

    Kwong, S.; Small, J.

    2007-07-01

    The fission products Cs-137 and Sr-90 are amongst the most common radionuclides occurring in ground contamination at the UK civil nuclear sites. Such contamination is often associated with alkaline liquids and the mobility of these fission products may be affected by these chemical conditions. Similar geochemical effects may also result from cementitious leachate associated with building foundations and the use of grouts to remediate ground contamination. The behaviour of fission products in these scenarios is a complex interaction of hydrogeological and geochemical processes. A suite of modelling tools have been developed to investigate the behaviour of a radioactive plume containing Cs and Sr. Firstly the effects of sorption due to cementitious groundwater is modelled using PHREEQC. This chemical model is then incorporated into PHAST for the 3-D reactive solute transport modeling. Results are presented for a generic scenario including features and processes that are likely to be relevant to a number of civil UK nuclear sites. Initial results show that modelling can be a very cost-effective means to study the complex hydrogeological and geochemical processes involved. Modelling can help predict the mobility of contaminants in a range of site end point scenarios, and in assessing the consequences of decommissioning activities. (authors)

  20. RADIOACTIVE BATTERY

    DOEpatents

    Birden, J.H.; Jordan, K.C.

    1959-11-17

    A radioactive battery which includes a capsule containing the active material and a thermopile associated therewith is presented. The capsule is both a shield to stop the radiations and thereby make the battery safe to use, and an energy conventer. The intense radioactive decay taking place inside is converted to useful heat at the capsule surface. The heat is conducted to the hot thermojunctions of a thermopile. The cold junctions of the thermopile are thermally insulated from the heat source, so that a temperature difference occurs between the hot and cold junctions, causing an electrical current of a constant magnitude to flow.

  1. RECLAMATION OF ALKALINE ASH PILES

    EPA Science Inventory

    The objective of the study was to develop methods for reclaiming ash disposal piles for the ultimate use as agricultural or forest lands. The ashes studied were strongly alkaline and contained considerable amounts of salts and toxic boron. The ashes were produced from burning bit...

  2. Radioactive Waste.

    ERIC Educational Resources Information Center

    Blaylock, B. G.

    1978-01-01

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

  3. Radioactive wastes

    SciTech Connect

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

    1996-11-01

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

  4. Isolation of alkaline mutagens from complex mixtures

    SciTech Connect

    Ho, C.H.; Guerin, M.R.; Clark, B.R.; Rao, T.K.; Epler, J.L.

    1981-05-01

    A method for the preparative-scale enrichment of alkaline mutagens from complex natural and anthropogenic mixtures is described. Mutagenic alkaline fractions were isolated from cigarette smoke, crude petroleum, and petroleum substitutes derived from coal and shale.

  5. [Microbiological Aspects of Radioactive Waste Storage].

    PubMed

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

    2015-01-01

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

  6. Process for treating alkaline wastes for vitrification

    DOEpatents

    Hsu, Chia-lin W.

    1994-01-01

    According to its major aspects and broadly stated, the present invention is a process for treating alkaline waste materials, including high level radioactive wastes, for vitrification. The process involves adjusting the pH of the wastes with nitric acid, adding formic acid (or a process stream containing formic acid) to reduce mercury compounds to elemental mercury and MnO{sub 2} to the Mn(II) ion, and mixing with class formers to produce a melter feed. The process minimizes production of hydrogen due to noble metal-catalyzed formic acid decomposition during, treatment, while producing a redox-balanced feed for effective melter operation and a quality glass product. An important feature of the present invention is the use of different acidifying and reducing, agents to treat the wastes. The nitric acid acidifies the wastes to improve yield stress and supplies acid for various reactions; then the formic acid reduces mercury compounds to elemental mercury and MnO{sub 2}) to the Mn(II) ion. When the pH of the waste is lower, reduction of mercury compounds and MnO{sub 2}) is faster and less formic acid is needed, and the production of hydrogen caused by catalytically-active noble metals is decreased.

  7. Nucleotide sequences encoding a thermostable alkaline protease

    DOEpatents

    Wilson, David B.; Lao, Guifang

    1998-01-01

    Nucleotide sequences, derived from a thermophilic actinomycete microorganism, which encode a thermostable alkaline protease are disclosed. Also disclosed are variants of the nucleotide sequences which encode a polypeptide having thermostable alkaline proteolytic activity. Recombinant thermostable alkaline protease or recombinant polypeptide may be obtained by culturing in a medium a host cell genetically engineered to contain and express a nucleotide sequence according to the present invention, and recovering the recombinant thermostable alkaline protease or recombinant polypeptide from the culture medium.

  8. Nucleotide sequences encoding a thermostable alkaline protease

    DOEpatents

    Wilson, D.B.; Lao, G.

    1998-01-06

    Nucleotide sequences, derived from a thermophilic actinomycete microorganism, which encode a thermostable alkaline protease are disclosed. Also disclosed are variants of the nucleotide sequences which encode a polypeptide having thermostable alkaline proteolytic activity. Recombinant thermostable alkaline protease or recombinant polypeptide may be obtained by culturing in a medium a host cell genetically engineered to contain and express a nucleotide sequence according to the present invention, and recovering the recombinant thermostable alkaline protease or recombinant polypeptide from the culture medium. 3 figs.

  9. Overall strategy and program plan for management of radioactively contaminated liquid wastes and transuranic sludges at the Oak Ridge National Laboratory

    SciTech Connect

    McNeese, L.E.; Berry, J.B.; Butterworth, G.E. III; Collins, E.D.; Monk, T.H.; Patton, B.D.; Snider, J.W.

    1988-12-01

    The use of hydrofracture was terminated after 1984, and LW concentrate has been accumulated and stored since that time. Currently, the volume of stored LW concentrate is near the safe fill limit for the 11 storage tanks in the active LW system, and significant operational constraints are being experienced. The tanks that provide the storage capacity of the active LW system contain significant volumes of TRU sludges that have been designated remote-handled transuranic (RH-TRU) wastes because of associated quantities of other radioisotopes, including /sup 90/Sr and /sup 137/Cs. Thirty-three additional tanks, which are inactive, also contain significant volumes of TRU waste and radioactive LW. A lack of adequate storage volume for LW jeopardizes ORNL's ability to ensure continued conduct of research and development (RandD) activities that generate LW because an unexpected operational incident could quickly deplete the remaining storage volume. Accordingly, a planning team comprised of staff members from the ORNL Nuclear and Chemical Waste Programs (NCWP) was created for developing recommended actions to be taken for management of LW. A program plan is presented which outlines work required for the development of a disposal method for each of the likely future waste streams associated with LW management and the disposal of the bulk of the resulting solid waste on the ORR. 8 refs., 20 figs., 12 tabs.

  10. Development of alkaline fuel cells.

    SciTech Connect

    Hibbs, Michael R.; Jenkins, Janelle E.; Alam, Todd Michael; Janarthanan, Rajeswari; Horan, James L.; Caire, Benjamin R.; Ziegler, Zachary C.; Herring, Andrew M.; Yang, Yuan; Zuo, Xiaobing; Robson, Michael H.; Artyushkova, Kateryna; Patterson, Wendy; Atanassov, Plamen Borissov

    2013-09-01

    This project focuses on the development and demonstration of anion exchange membrane (AEM) fuel cells for portable power applications. Novel polymeric anion exchange membranes and ionomers with high chemical stabilities were prepared characterized by researchers at Sandia National Laboratories. Durable, non-precious metal catalysts were prepared by Dr. Plamen Atanassov's research group at the University of New Mexico by utilizing an aerosol-based process to prepare templated nano-structures. Dr. Andy Herring's group at the Colorado School of Mines combined all of these materials to fabricate and test membrane electrode assemblies for single cell testing in a methanol-fueled alkaline system. The highest power density achieved in this study was 54 mW/cm2 which was 90% of the project target and the highest reported power density for a direct methanol alkaline fuel cell.