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Sample records for resin regeneration waste

  1. System for processing ion exchange resin regeneration waste liquid in atomic power plant

    International Nuclear Information System (INIS)

    Onaka, Noriyuki; Tanno, Kazuo; Shoji, Saburo.

    1976-01-01

    Object: To reduce the quantity of radioactive waste to be solidified by recovering and repeatedly using sulfuric acid and sodium hydroxide which constitute the ion exchange resin regeneration waste liquid. Structure: Cation exchange resin regeneration waste liquid is supplied to an anion exchange film electrolytic dialyzer for recovering sulfuric acid through separation from impurity cations, while at the same time anion exchange resin regeneration waste liquid is supplied to a cation exchange film electrolytic dialyzer for recovering sodium hydroxide through separation from impurity anions. The sulfuric acid and sodium hydroxide thus recovered are condensed by a thermal condenser and then, after density adjustment, repeatedly used for the regeneration of the ion exchange resin. (Aizawa, K.)

  2. Device for processing regenerative wastes of ion exchange resin

    International Nuclear Information System (INIS)

    Kuroda, Osamu; Ebara, Katsuya; Shindo, Toshikazu; Takahashi, Sankichi

    1986-01-01

    Purpose: To facilitate the operation and maintenance of a processing device by dividing radioactive wastes produced in the regenerative process of ion exchange resin into a regenerated usable recovery liquid and wastes. Constitution: Sulfuric acid is recovered by a diffusion dialysis method from wastes containing sulfuric acid that are generated in the regenerative process of cation-exchange resin and also caustic soda is recovered by the diffusion dialysis method from wastes containing caustic soda that are generated in the regenerative process of anion-exchange resin. The sulfuric acid and caustic soda thus recovered are used for the regeneration of ion-exchange resin. A concentrator is provided for concentrating the sulfuric acid and caustic soda water solution to concentration suitable for the regeneration of these ion-exchange resins. Also provided is a recovery device for recovering water generated from the concentrator. This device is of so simple a constitution that its operation and maintenance can be performed very easily, thereby greatly reducing the quantity of waste liquid required to be stored in drums. (Takahashi, M.)

  3. Evaluation of Resin Regeneration Using HCl and H2SO4 for the Ion Exchanger of Copper

    International Nuclear Information System (INIS)

    Prayitno; Djoko Sardjono

    2002-01-01

    The experimental investigation on the regeneration of resin using HCl and H 2 SO 4 with its varian concentration of 1; 2.5; 2; 2.5 and 3 N and the stirring time was 5; 10; 15; 20; and 25 minutes. For evaluating their effectiveness on the separation of ion copper in the waste with concentration 500 ppm. Experimentally this investigation is the first step of resin results of regeneration process usage as an alternative resin for the treatment of liquid waste containing especially copper. The experimental resulted by mixing the feed copper waste with resin after regeneration. Therefore it could be concluded that the most effective regeneration was obtained with HCl as the regeneration of concentration 2 N and the stirring time 15 minutes with the percentage of separation used of 85.1 %. (author)

  4. Separation of boron isotopes by ion exchange chromatography: studies on regeneration of strong base anion exchange resins

    International Nuclear Information System (INIS)

    Sharma, B.K.; Subramanian, R.; Mathur, P.K.

    1994-01-01

    The optimum conditions for the regeneration of strong base anion exchange resins of type-I and type-II were determined for cost-effective separation of isotopes of boron by ion exchange chromatography where the hydroxyl form of an anion exchange resin is equilibrated with boric acid solution containing mannitol as a complexing reagent. The possibility of using unspent alkali content of the effluent was also exploited. Removal of carbonate impurity from Rayon grade caustic lye (used as regenerant after dilution) and recycling of Ba(OH) 2 was studied to avoid waste disposal problems. (author)

  5. Method for regenerating magnetic polyamine-epichlorohydrin resin

    Science.gov (United States)

    Kochen, Robert L.; Navratil, James D.

    1997-07-29

    Magnetic polymer resins capable of efficient removal of actinides and heavy metals from contaminated water are disclosed together with methods for making, using, and regenerating them. The resins comprise polyamine-epichlorohydrin resin beads with ferrites attached to the surfaces of the beads. Markedly improved water decontamination is demonstrated using these magnetic polymer resins of the invention in the presence of a magnetic field, as compared with water decontamination methods employing ordinary ion exchange resins or ferrites taken separately.

  6. Saturated Resin Ectopic Regeneration by Non-Thermal Dielectric Barrier Discharge Plasma

    Directory of Open Access Journals (Sweden)

    Chunjing Hao

    2017-11-01

    Full Text Available Textile dyes are some of the most refractory organic compounds in the environment due to their complex and various structure. An integrated resin adsorption/Dielectric Barrier Discharge (DBD plasma regeneration was proposed to treat the indigo carmine solution. It is the first time to report ectopic regeneration of the saturated resins by non-thermal Dielectric Barrier Discharge. The adsorption/desorption efficiency, surface functional groups, structural properties, regeneration efficiency, and the intermediate products between gas and liquid phase before and after treatment were investigated. The results showed that DBD plasma could maintain the efficient adsorption performance of resins while degrading the indigo carmine adsorbed by resins. The degradation rate of indigo carmine reached 88% and the regeneration efficiency (RE can be maintained above 85% after multi-successive regeneration cycles. The indigo carmine contaminants were decomposed by a variety of reactive radicals leading to fracture of exocyclic C=C bond, which could cause decoloration of dye solution. Based on above results, a possible degradation pathway for the indigo carmine by resin adsorption/DBD plasma treatment was proposed.

  7. Electro regeneration of iodide loaded resin. Contributed Paper RD-18

    International Nuclear Information System (INIS)

    Kumar, Ratnesh; Kumar, T.; Sree Kumar, B.; Seshadri, K.S.; Paul, Biplob

    2014-01-01

    Spent resins generated in the nuclear reactor contain essentially cationic activities due to Cesium, Strontium, Cobalt, and anionic activities due to Iodide, Iodate etc with activity loading to the extent of 0.1 Cim -3 and a surface dose of the order of 5 R. It is necessary to convert the spent resin into innocuous, reusable forms. An attempt has been made to regenerate Iodide containing spent resin into OH - electrolytically by using the OH - produced at the cathode compartment of an electrolytic cell. Results show that the regeneration of the spent resin containing Iodide could be completely accomplished electrolytically more efficiently than by addition of alkali. (author)

  8. The regeneration test of the secondary loop condensate polishing mixed bed resin in Qinshan NPP

    International Nuclear Information System (INIS)

    Xu Meijing; Dong Liming

    1995-12-01

    There are four condensate polishing mixed beds in the water chemical treatment plant of Qinshan NPP. 2125 kg of D001-TR type cation exchange resin, 2000 kg of D201-TR type anion exchange resin, and 375 kg of S-TR type inert resin are filled into each mixed bed. The bed height of resin is 1.2 m and the volume is about 2.7 m 3 . In order to regenerate the exhausted resin out of the bed, the pre-designed condensate polishing mixed bed regeneration process was used to regenerate the first exhausted resin. After the resin was scrubbed and separated, cation resin and anion resin were respectively regenerated, rinsed to resume the exchange capability of the resin. The regenerated mixed bed is able to keep higher efficiency for condensate polishing. The outlet water quality and the resin service-life are able to meet the design requirements or more favorable than that. During the test, some main cations and anions in the blow-off water at each procedure were analyzed. The analyzed results were used to make pre-designed regeneration process better. The test results proved that pre-designed process is reasonable and effective. (6 refs., 6 figs., 7 tabs.)

  9. Solidification of ion exchange resin wastes

    International Nuclear Information System (INIS)

    1982-08-01

    Solidification media investigated included portland type I, portland type III and high alumina cements, a proprietary gypsum-based polymer modified cement, and a vinyl ester-styrene thermosetting plastic. Samples formulated with hydraulic cement were analyzed to investigate the effects of resin type, resin loading, waste-to-cement ratio, and water-to-cement ratio. The solidification of cation resin wastes with portland cement was characterized by excessive swelling and cracking of waste forms, both after curing and during immersion testing. Mixed bed resin waste formulations were limited by their cation component. Additives to improve the mechanical properties of portland cement-ion exchange resin waste forms were evaluated. High alumina cement formulations dislayed a resistance to deterioration of mechanical integrity during immersion testing, thus providing a significant advantage over portland cements for the solidification of resin wastes. Properties of cement-ion exchange resin waste forms were examined. An experiment was conducted to study the leachability of 137 Cs, 85 Sr, and 60 Co from resins modified in portland type III and high alumina cements. The cumulative 137 Cs fraction release was at least an order of magnitude greater than that of either 85 Sr or 60 Co. Release rates of 137 Cs in high alumina cement were greater than those in portland III cement by a factor of two.Compressive strength and leach testing were conducted for resin wastes solidified with polymer-modified gypsum based cement. 137 Cs, 85 Sr, and 60 Co fraction releases were about one, two and three orders of magnitude higher, respectively, than in equivalent portland type III cement formulations. As much as 28.6 wt % dry ion exchange resin was successfully solidified using vinyl ester-styrene compared with a maximum of 25 wt % in both portland and gypsum-based cement

  10. Disposal of bead ion exchange resin wastes

    International Nuclear Information System (INIS)

    Gay, R.L.; Granthan, L.F.

    1985-01-01

    Bead ion exchange resin wastes are disposed of by a process which involves spray-drying a bead ion exchange resin waste in order to remove substantially all of the water present in such waste, including the water on the surface of the ion exchange resin beads and the water inside the ion exchange resin beads. The resulting dried ion exchange resin beads can then be solidified in a suitable solid matrix-forming material, such as a polymer, which solidifies to contain the dried ion exchange resin beads in a solid monolith suitable for disposal by burial or other conventional means

  11. Solidification of ion exchange resin wastes in hydraulic cement

    International Nuclear Information System (INIS)

    Neilson, R.M. Jr.; Kalb, P.; Fuhrmann, M.; Colombo, P.

    1982-01-01

    Work has been conducted to investigate the solidification of ion exchange resin wastes with portland cements. These efforts have been directed toward the development of acceptable formulations for the solidification of ion exchange resin wastes and the characterization of the resultant waste forms. This paper describes formulation development work and defines acceptable formulations in terms of ternary phase compositional diagrams. The effects of cement type, resin type, resin loading, waste/cement ratio and water/cement ratio are described. The leachability of unsolidified and solidified resin waste forms and its relationship to full-scale waste form behavior is discussed. Gamma irradiation was found to improve waste form integrity, apparently as a result of increased resin crosslinking. Modifications to improve waste form integrity are described. 3 tables

  12. HIGH ASPECT RATIO ION EXCHANGE RESIN BED - HYDRAULIC RESULTS FOR SPERICAL RESIN BEADS

    International Nuclear Information System (INIS)

    Duignan, M; Charles Nash, C; Timothy Punch, T

    2007-01-01

    A principal role of the DOE Savannah River Site is to safely dispose of a large volume of liquid nuclear waste held in many storage tanks. An in-tank ion exchange unit is being considered for cesium removal to accelerate waste processing. This unit is planned to have a relatively high bed height to diameter ratio (10:1). Complicating the design is the need to cool the ion exchange media; therefore, the ion exchange column will have a central cooling core making the flow path annular. To separate cesium from waste the media being considered is made of resorcinol formaldehyde resin deposited on spherical plastic beads and is a substitute for a previously tested resin made of crystalline silicotitanate. This spherical media not only has an advantage of being mechanically robust, but, unlike its predecessor, it is also reusable, that is, loaded cesium can be removed through elution and regeneration. Resin regeneration leads to more efficient operation and less spent resin waste, but its hydraulic performance in the planned ion exchange column was unknown. Moreover, the recycling process of this spherical resorcinol formaldehyde causes its volume to significantly shrink and swell. To determine the spherical media's hydraulic demand a linearly scaled column was designed and tested. The waste simulant used was prototypic of the wastes' viscosity and density. This paper discusses the hydraulic performance of the media that will be used to assist in the design of a full-scale unit

  13. Embedding of reactor wastes in plastic resins

    International Nuclear Information System (INIS)

    1979-01-01

    STEAG Kernenergie GmbH is so far the only firm commercially to condition radioactive bead ion exchange resins by embedding in polystyrene resins. The objective of the work reported here was to study and develop methods for immobilization of other reactor wastes in plastic resins. Comparison studies on high quality cement however showed favourable results for cement with respect to process safety and economy. For this reason STEAG interrupted its work in the field of resin embedding after about one year. The work carried out during this period is surveyed in this report, which includes a comprehensive literature study on reactor wastes and their solidification in plastic resins as well as on regulations with regard to radioactive waste disposal in the member states of the European Communities

  14. Method of heat decomposition for chemical decontaminating resin waste

    International Nuclear Information System (INIS)

    Kikuchi, Akira.

    1988-01-01

    Purpose: To make resin wastes into non-deleterious state, discharge them into a resin waste storage tank of existent radioactive waste processing facility and store and dispose them. Constitution: In the processing of chemical decontaminating resin wastes, iron exchange resins adsorbing chemical decontaminating agents comprising a solution of citric acid, oxalic acid, formic acid and EDTA alone or as a mixture of them are heated to dry, thermally decomposed and then separated from the ion exchange resins. That is, the main ingredients of the chemical decontaminating agents are heat-decomposed when heated and dried at about 250 deg C in air and converted into non-toxic gases such as CO, CO 2 , NO, NO 2 or H 2 O. Further, since combustion or carbonization of the basic materials for the resin is not caused at such a level of temperature, the resin wastes removed with organic acid and chelating agents are transferred to an existent resin waste storage tank and stored therein. In this way, facility cost and radiation exposure can remarkably be decreased. (Kamimura, M.)

  15. Regeneration of pilot-scale ion exchange columns for hexavalent chromium removal.

    Science.gov (United States)

    Korak, Julie A; Huggins, Richard; Arias-Paic, Miguel

    2017-07-01

    Due to stricter regulations, some drinking water utilities must implement additional treatment processes to meet potable water standards for hexavalent chromium (Cr(VI)), such as the California limit of 10 μg/L. Strong base anion exchange is effective for Cr(VI) removal, but efficient resin regeneration and waste minimization are important for operational, economic and environmental considerations. This study compared multiple regeneration methods on pilot-scale columns on the basis of regeneration efficiency, waste production and salt usage. A conventional 1-Stage regeneration using 2 N sodium chloride (NaCl) was compared to 1) a 2-Stage process with 0.2 N NaCl followed by 2 N NaCl and 2) a mixed regenerant solution with 2 N NaCl and 0.2 N sodium bicarbonate. All methods eluted similar cumulative amounts of chromium with 2 N NaCl. The 2-Stage process eluted an additional 20-30% of chromium in the 0.2 N fraction, but total resin capacity is unaffected if this fraction is recycled to the ion exchange headworks. The 2-Stage approach selectively eluted bicarbonate and sulfate with 0.2 N NaCl before regeneration using 2 N NaCl. Regeneration approach impacted the elution efficiency of both uranium and vanadium. Regeneration without co-eluting sulfate and bicarbonate led to incomplete uranium elution and potential formation of insoluble uranium hydroxides that could lead to long-term resin fouling, decreased capacity and render the resin a low-level radioactive solid waste. Partial vanadium elution occurred during regeneration due to co-eluting sulfate suppressing vanadium release. Waste production and salt usage were comparable for the 1- and 2-Stage regeneration processes with similar operational setpoints with respect to chromium or nitrate elution. Published by Elsevier Ltd.

  16. Chemical derivatization to enhance chemical/oxidative stability of resorcinol-formaldehyde resin

    Energy Technology Data Exchange (ETDEWEB)

    Hubler, T.L. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-10-01

    The goal of this task is to develop modified resorcinol-formaldehyde (R-F) resin to improve the chemical/oxidative stability of the resin. R-F resin is a regenerable organic ion-exchange resin that is selective for cesium ion in highly alkaline, high ionic-strength solutions. R-F resin tends to undergo chemical degradation, reducing its ability to remove cesium ion from waste solutions; the mechanistic details of these decomposition reactions are currently unknown. The approach used for this task is chemical modification of the resin structure, particularly the resorcinol ring unit of the polymer resin. This approach is based on prior characterization studies conducted at Pacific Northwest National Laboratory (PNNL) that indicated the facile chemical degradation of the resin is oxidation of the resorcinol ring to the para-quinone structure, with subsequent loss of ion-exchange sites for cesium ion. R-F resin represents an important alternative to current radiocesium remediation technology for tank wastes at both the Hanford and Savannah River sites, particularly if regenerable resins are needed.

  17. Column adsorption of perchlorate by amine-crosslinked biopolymer based resin and its biological, chemical regeneration properties.

    Science.gov (United States)

    Song, Wen; Xu, Xing; Tan, Xin; Wang, Yan; Ling, Jianya; Gao, Baoyu; Yue, Qinyan

    2015-01-22

    Column adsorption of perchlorate by amine-crosslinked biopolymer based resin was investigated by considering the bed depth, stream flow rate and influent pH. The empty bed contact time (EBCT) increased with the growth of bed depths, meanwhile rising flow rate at constant bed depth (3.4 cm) decreased the breakthrough time. It was observed that perchlorate adsorption capacity was optimum at neutral condition (pH: 6.0, 170.4 mg/g), and decreased at acidic (pH: 3.0, 96.4 mg/g) or alkalic (pH: 12.0, 72.8 mg/g) influents. The predominant strains of the acclimated sludge for resin biological regeneration were the β-subclass of Proteobacteria. Biological regeneration of the saturated amine-crosslinked biopolymer based resin with mixed bacteria have shown its merit with regeneration and biological perchlorate destruction simultaneously, although its regeneration efficiency was only 61.2-84.1% by contrast to chemical regeneration with efficiency more than 95%. Copyright © 2014 Elsevier Ltd. All rights reserved.

  18. Bio nitrate Project: a new technology for water nitrate elimination by means of ionic exchange resins

    International Nuclear Information System (INIS)

    Arellano Ortiz, J.

    2009-01-01

    The use of ion exchange resins for nitrate elimination from water generates a waste containing a sodium chloride mixture plus the retained nitrates. this waste must be correctly disposed. In this project, the resin ionic form is modified to be regenerated with other compounds, different from the common salt, which are interesting because of the presence of mineral nutrition. So, with Bio nitrate Project, nitrates are recovered and the regeneration waste is apt to be use as fertilizer, for agricultural uses, or as complementary contribution of nutrients in biological water treatment. (Author) 27 refs.

  19. Dilute chemical decontamination resins and the mixed waste issue

    International Nuclear Information System (INIS)

    Denault, R.P.; Hallman, J.T.

    1988-01-01

    The decontamination of reactor primary systems, sub-systems and components is an important method used to reduce the occupational radiation exposure of nuclear plant personnel. The waste produced by the application of this technology is mainly solid in the form of ion exchange resins. As a result of a recent agreement between the Environmental Protection Agency (EPA) and the Nuclear Regulatory Commission (NRC), all radioactive waste must meet EPA burial criteria. The chemicals used in a decontamination and certain metals dissolved during the process, primarily chromium, could render the waste hazardous as well as radioactive or more commonly called a mixed waste. This paper defines mixed waste as described in the EPA directive 9432.00-2, and examine the criteria by which waste is categorized as hazardous. The decontamination waste resin generated by two processes, the CAN-DEREM and the LOMI process, is described in detail. Waste data obtained from decontaminations performed by LN Technologies Corporation including chemical, metal and radionuclide loadings on resins from both PWR and BWR applications are presented

  20. Solidification of radioactive waste resins using cement mixed with organic material

    Energy Technology Data Exchange (ETDEWEB)

    Laili, Zalina, E-mail: liena@nm.gov.my [Nuclear Science Programme, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia (UKM), Bangi, 43600, Selangor Malaysia (Malaysia); Waste and Environmental Technology Division, Malaysian Nuclear Agency (Nuclear Malaysia), Bangi, 43000 Kajang, Selangor (Malaysia); Yasir, Muhamad Samudi [Nuclear Science Programme, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia (UKM), Bangi, 43600, Selangor Malaysia (Malaysia); Wahab, Mohd Abdul [Waste and Environmental Technology Division, Malaysian Nuclear Agency (Nuclear Malaysia), Bangi, 43000 Kajang, Selangor (Malaysia)

    2015-04-29

    Solidification of radioactive waste resins using cement mixed with organic material i.e. biochar is described in this paper. Different percentage of biochar (0%, 5%, 8%, 11%, 14% and 18%) was investigated in this study. The characteristics such as compressive strength and leaching behavior were examined in order to evaluate the performance of solidified radioactive waste resins. The results showed that the amount of biochar affect the compressive strength of the solidified resins. Based on the data obtained for the leaching experiments performed, only one formulation showed the leached of Cs-134 from the solidified radioactive waste resins.

  1. Solidification of radioactive waste resins using cement mixed with organic material

    International Nuclear Information System (INIS)

    Laili, Zalina; Yasir, Muhamad Samudi; Wahab, Mohd Abdul

    2015-01-01

    Solidification of radioactive waste resins using cement mixed with organic material i.e. biochar is described in this paper. Different percentage of biochar (0%, 5%, 8%, 11%, 14% and 18%) was investigated in this study. The characteristics such as compressive strength and leaching behavior were examined in order to evaluate the performance of solidified radioactive waste resins. The results showed that the amount of biochar affect the compressive strength of the solidified resins. Based on the data obtained for the leaching experiments performed, only one formulation showed the leached of Cs-134 from the solidified radioactive waste resins

  2. Evaluation of ferrocyanide anion exchange resins regarding the uptake of Cs+ ions and their regeneration

    International Nuclear Information System (INIS)

    Won, Hui Jun; Mooon, Jei Kwon; Jung, Chong Hun; Chung, Won Yang

    2008-01-01

    Ferrocyanide-anion exchange resin was prepared and the prepared ion exchange resins were tested on the ability to uptake Cs + ion. The prepared ion exchange resins were resin-KCoFC, resin-KNiFC, and resin-KCuFC. The three tested ion exchange resins showed ion exchange selectivity on the Cs + ion of the surrogate soil decontamination solution, and resin- KCoFC showed the best Cs + ion uptake capability among the tested ion exchange resins. The ion exchange behaviors were explained well by the modified Dubinin-Polanyi equation. A regeneration feasibility study of the spent ion exchange resins was also performed by the successive application of hydrogen peroxide and hydrazine. The desorption of the Cs + ion from the ion exchange resin satisfied the electroneutrality condition in the oxidation step; the desorption of the Fe 2+ ion in the reduction step could also be reduced by adding the K + ion

  3. Fixing of various simulated radioactive wastes in urea-formaldehyde resin

    International Nuclear Information System (INIS)

    Du Dahai; Wei Peng

    1986-01-01

    This paper outlines the results of the fixing of a variety of simulated radioactive wastes in the urea-formaldehyde resin. The radioactive waste materials fixed include spent ion exchange resin, concentrates of NaNO 3 -NaBO 2 as well as NaBO 2 and sludge. The performance of the fixed products has been improved by means of selecting the synthetic conditions of resin, a suitable hardener and an inorganic additive

  4. Embedding of radioactive wastes by thermosetting resins

    International Nuclear Information System (INIS)

    Baer, A.; Traxler, A.; Limongi, A.; Thiery, D.

    The process for embedding radioactive wastes in thermosetting resins perfected and applied at the Grenoble Nuclear Research Center and its application to the treatment of radioactive wastes from Light-Water Nuclear Power Plants (PWR and BWR) are presented. The various types of wastes are enumerated and their activities and quantities are estimated: evaporator concentrates, ion exchange resins, filtration sludges, filters, various solid wastes, etc. The authors review the orientations of the research performed and indicate, for each type of waste considered, the cycle of treatment operations from rendering the radioelements insoluble to drying the concentrates to final embedding. The operational safety of the process and the safety of transport and storage of the embedded wastes are investigated. The essential technical features concerning the safety of the installation and of the final product obtained are presented. In particular, results are presented from tests of resistance to fire, irradiation, leaching, etc., these being characteristics which represent safety criteria. The economic aspects of the process are considered by presenting the influences of the reduction of volume and weight of wastes to be stored, simplicity of installations and cost of primary materials

  5. Destruction of organic wastes by ammonium peroxydisulfate with electrolytic regeneration of the oxidant

    International Nuclear Information System (INIS)

    Cooper, J.F.; Wang, J.F.; Krueger, R.; King, K.

    1997-01-01

    Research is reported concerning a new aqueous process for oxidative destruction of solid- and liquid organic wastes. This process uses acidified ammonium peroxydisulfate and operates at ambient pressure and at 80- to 100 degrees C. The oxidant may be efficiently regenerated by electrolysis of the sulfate by-product at Pt anodes, even in the presence of organic and inorganic contaminants expected to be entrained in the cycle. Integral rate constants were determined for the oxidation of 25 diverse organic compounds at low (50 ppm) concentrations through fixed-time experiments with excess oxidant and a Pt wire catalyst. For high initial concentrations, uncatalyzed mineralization rates were measured for waste surrogates including kerosene, triethylamine, ion exchange resin, oxalic acid, trinitrotoluene, and cellulose. A packed bed reactor was tested with ethylene glycol, with offgas analysis by mass spectroscopy. Rate data extrapolate to throughputs of approximately 200 kg/m 3 -day. The process may benefit the destruction of highly toxic or specialized industrial wastes as well as the organic fraction of mixed wastes

  6. Evaluation of ferrocyanide anion exchange resins regarding the uptake of Cs{sup +} ions and their regeneration

    Energy Technology Data Exchange (ETDEWEB)

    Won, Hui Jun; Mooon, Jei Kwon; Jung, Chong Hun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Chung, Won Yang [Kangwon University, Chuncheon (Korea, Republic of)

    2008-10-15

    Ferrocyanide-anion exchange resin was prepared and the prepared ion exchange resins were tested on the ability to uptake Cs{sup +} ion. The prepared ion exchange resins were resin-KCoFC, resin-KNiFC, and resin-KCuFC. The three tested ion exchange resins showed ion exchange selectivity on the Cs{sup +} ion of the surrogate soil decontamination solution, and resin- KCoFC showed the best Cs{sup +} ion uptake capability among the tested ion exchange resins. The ion exchange behaviors were explained well by the modified Dubinin-Polanyi equation. A regeneration feasibility study of the spent ion exchange resins was also performed by the successive application of hydrogen peroxide and hydrazine. The desorption of the Cs{sup +} ion from the ion exchange resin satisfied the electroneutrality condition in the oxidation step; the desorption of the Fe{sup 2+} ion in the reduction step could also be reduced by adding the K{sup +} ion.

  7. Plastic solidification method for radioactive waste

    International Nuclear Information System (INIS)

    Tomita, Toshihide; Inakuma, Masahiko.

    1992-01-01

    Condensed liquid wastes in radioactive wastes are formed by mixing and condensing several kinds of liquid wastes such as liquid wastes upon regeneration of ion exchange resins, floor draining liquid wastes and equipment draining liquid wastes. Accordingly, various materials are contained, and it is found that polymerization reaction of plastics is inhibited especially when reductive material, such as sodium nitrite is present. Then, in the present invention, upon mixing thermosetting resins to radioactive wastes containing reducing materials, an alkaline material is admixed to an unstaturated polyester resin. This can inactivate the terminal groups of unsaturated polyester chain, to prevent the dissociation of the reducing agent such as sodium nitrite. Further, if an unsaturated polyester resin of low acid value and a polymerization initiator for high temperature are used in addition to the alkaline material, the effect is further enhanced, thereby enabling to obtain a strong plastic solidification products. (T.M.)

  8. Evaluating Residence Time for Cesium Removal from Simulated Hanford Tank Wastes Using SuperLig(R) 644 Resin

    International Nuclear Information System (INIS)

    Hassan, N.M.

    2003-01-01

    Batch contact and column experiments were performed to evaluate the effect of residence time on cesium (Cs) removal from two simulated Hanford tank wastes using SuperLig(R) 644 resin. The two waste simulants mimic the compositions of tanks 241-AZ-102 and 241-AN-107 at the U.S. Department of Energy (DOE) Hanford site. A single column made of glass tube (2.7-cm i.d.), which contained approximately 100 mL of H-form SuperLig(R) 644 resin was used in the column experiments. The experiments each consisted of loading, elution, and regeneration steps were performed at flow rates ranging from 0.64 to 8.2 BV/h for AZ-102 and from 1.5 to 18 BV/h for AN-107 simulant. The lowest flow rates of 0.64 and 1.5 BV/h were selected to evaluate less than optimal flow conditions in the plant. The range of the flow rates is consistent with the River Protection Project design for the waste treatment plant (WTP) columns, which will operate at a flow rate between 1.5 to 3 BV/h. Batch contact experiments were also performed for two batches of SuperLig(R) 644 to determine the equilibrium distribution coefficients (Kds) as a function of Cs concentration

  9. Leaching test of bituminized waste and waste solidified by epoxy resin

    International Nuclear Information System (INIS)

    Yoshinaka, Kazuyuki; Sugaya, Atsushi; Onizawa, Toshikazu; Takano, Yugo; Kimura, Yukihiko

    2008-10-01

    About 30,000 bituminized waste drums and about 1800 drums of waste solidified by epoxy resin, generated from Tokai Reprocessing Plant, were stored in storage facilities. And study for disposal of these waste is performed. It was considered that radioactive nuclides and chemical components were released from these waste by contact of underground water, when disposed there waste. This paper is reported that result of leaching tests for these waste, done from 2003 to 2006. We've get precious knowledge and data, as follows. (1) In leaching tests for bituminized waste, it has detected iodine-129 peak, considered difficult too low energy gamma to detect. We've get data and knowledge of iodine-129 behavior first. Leached radioactivity for 50 days calculated by peak area was equal for about 40% and 100% of including radioactivity in bituminized waste sample. And we've get data of behavior of nitric acid ion and so on, important to study for disposal, in various condition of sample shape or leaching liquid temperature. (2) In leaching test for waste solidified by epoxy resin, we've get data of behavior of TBP, radionuclides and so on, important to study for disposal. Leached TBP was equal about 1% of including of sample. And we've get data of iodine-129 behavior, too. It was confirmed that leached iodine-129 was equal for about 60% and 100% of including sample, for 90 days. (author)

  10. Ontario Hydro Research Division's program for treatment of spent ion-exchange resins

    International Nuclear Information System (INIS)

    Nott, B.R.; Dodd, D.J.R.

    1981-09-01

    A brief review of the evolution of work programmes for chemical treatment of spent ion-exchange resins in Ontario Hydro's Research Division is presented. Attention has been focussed on pre-treatment processes for the treatment of the spent resins prior to encapsulation of the products in solid matrices. Spent Resin Regeneration and Acid Stripping processes were considered in some detail. Particular attention was paid to carbon-14 on spent resins, its determination in and removal from the spent resins (with the acid stripping technique). The use of separate cation and anion resin beds instead of mixed bed resins was examined with a view to reducing the volume of resin usage and consequently the volume of waste radioactive ion-exchange resin generated. (author)

  11. Receiving Basin for Offsite Fuels and the Resin Regeneration Facility Safety Analysis Report, Executive Summary

    International Nuclear Information System (INIS)

    Shedrow, C.B.

    1999-01-01

    The Safety Analysis Report documents the safety authorization basis for the Receiving Basin for Offsite Fuels (RBOF) and the Resin Regeneration Facility (RRF) at the Savannah River Site (SRS). The present mission of the RBOF and RRF is to continue in providing a facility for the safe receipt, storage, handling, and shipping of spent nuclear fuel assemblies from power and research reactors in the United States, fuel from SRS and other Department of Energy (DOE) reactors, and foreign research reactors fuel, in support of the nonproliferation policy. The RBOF and RRF provide the capability to handle, separate, and transfer wastes generated from nuclear fuel element storage. The DOE and Westinghouse Savannah River Company, the prime operating contractor, are committed to managing these activities in such a manner that the health and safety of the offsite general public, the site worker, the facility worker, and the environment are protected

  12. Receiving Basin for Offsite Fuels and the Resin Regeneration Facility Safety Analysis Report, Executive Summary

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, C.B.

    1999-11-29

    The Safety Analysis Report documents the safety authorization basis for the Receiving Basin for Offsite Fuels (RBOF) and the Resin Regeneration Facility (RRF) at the Savannah River Site (SRS). The present mission of the RBOF and RRF is to continue in providing a facility for the safe receipt, storage, handling, and shipping of spent nuclear fuel assemblies from power and research reactors in the United States, fuel from SRS and other Department of Energy (DOE) reactors, and foreign research reactors fuel, in support of the nonproliferation policy. The RBOF and RRF provide the capability to handle, separate, and transfer wastes generated from nuclear fuel element storage. The DOE and Westinghouse Savannah River Company, the prime operating contractor, are committed to managing these activities in such a manner that the health and safety of the offsite general public, the site worker, the facility worker, and the environment are protected.

  13. Decontamination of spent ion-exchangers contaminated with cesium radionuclides using resorcinol-formaldehyde resins.

    Science.gov (United States)

    Palamarchuk, Marina; Egorin, Andrey; Tokar, Eduard; Tutov, Mikhail; Marinin, Dmitry; Avramenko, Valentin

    2017-01-05

    The origin of the emergence of radioactive contamination not removable in the process of acid-base regeneration of ion-exchange resins used in treatment of technological media and liquid radioactive waste streams has been determined. It has been shown that a majority of cesium radionuclides not removable by regeneration are bound to inorganic deposits on the surface and inside the ion-exchange resin beads. The nature of the above inorganic inclusions has been investigated by means of the methods of electron microscopy, IR spectrometry and X-ray diffraction. The method of decontamination of spent ion-exchange resins and zeolites contaminated with cesium radionuclides employing selective resorcinol-formaldehyde resins has been suggested. Good prospects of such an approach in deep decontamination of spent ion exchangers have been demonstrated. Copyright © 2016 Elsevier B.V. All rights reserved.

  14. Experience with the incorporation of low and medium-level wastes in thermosetting resins

    International Nuclear Information System (INIS)

    Aubouin, G.; Hallier, P.; Bruand, J.P.

    1980-01-01

    This paper deals with the experience gained in the packaging of low and medium-level radioactive wastes in thermosetting resins. A prototype workshop has been functioning in the Nuclear Research Centre at Grenoble since 1975. The wastes processed are evaporator concentrates and ion exchange resins. A pilot plant which has been built at the PWR power station in Chooz enables evaporator concentrates, ion-exchange resins and filter cartridges to be processed. In each case, the solidifying agent is based on a polyester or epoxy resin. The properties of the cured product (leaching rate, irradiation and fire resistance, and mechanical strength) are given. In order to widen the application of thermosetting resins, the containment of soluble radioactive salts has been studied. The use of this process for wastes arising from the decommissioning of nuclear power stations seems feasible. The coefficients of diffusion of radioactive elements through the thermosetting resins have been measured. Using them, the amounts of radioactivity released as a function of time have been calculated

  15. Processing ix spent resin waste for C-14 isotope recovery

    International Nuclear Information System (INIS)

    Chang, F. H.; Woodall, K. B.; Sood, S. K.; Vogt, H. K.; Krochmainek, L. S.

    1991-01-01

    A process developed at Ontario Hydro for recovering carbon-14 (C-14) from spent ion exchange resin wastes is described. Carbon-14 is an undesirable by-product of CANDU 1 nuclear reactor operation. It has an extremely long (5730 years) half-life and can cause dosage to inhabitants by contact, inhalation, or through the food cycle via photosynthesis. Release of carbon-14 to the environment must be minimized. Presently, all the C-14 produced in the Moderator and Primary Heat Transport (PHT) systems of the reactor is effectively removed by the respective ion exchange columns, and the spent ion exchange resins are stored in suitably engineered concrete structures. Because of the large volumes of spent resin waste generated each year this method of disposal by long term storage tends to be uneconomical; and may also be unsatisfactory considering the long half-life of the C-14. However, purified C-14 is a valuable commercial product for medical, pharmaceutical, agricultural, and organic chemistry research. Currently, commercial C-14 is made artificially in research reactors by irradiating aluminum nitride targets for 4.5 years. If the C-14 containing resin waste can be used to reduce this unnecessary production of C-14, the total global build-up of this radioactive chemical can be reduced. There is much incentive in removing the C-14 from the resin waste to reduce the volume of C-14 waste, and also in purifying the recovered C-14 to supply the commercial market. The process developed by Ontario Hydro consists of three main steps: C-14 removal from spent resins, enrichment of recovered C-14, and preparation of final product. Components of the process have been successfully tested at Ontario Hydro's Research Division, but the integration of the process is yet to be demonstrated. A pilot scale plant capable of processing 4 m 3 of spent resins annually is being planned for demonstrating the technology. The measured C-14 activity levels on the spent resins ranged from 47

  16. Bio nitrate Project: a new technology for water nitrate elimination by means of ionic exchange resins; Proyecto Bionitrate: una nueva tecnologia para la eliminacion de nitratos en aguas mediante resinas de intercambio ionico

    Energy Technology Data Exchange (ETDEWEB)

    Arellano Ortiz, J.

    2009-07-01

    The use of ion exchange resins for nitrate elimination from water generates a waste containing a sodium chloride mixture plus the retained nitrates. this waste must be correctly disposed. In this project, the resin ionic form is modified to be regenerated with other compounds, different from the common salt, which are interesting because of the presence of mineral nutrition. So, with Bio nitrate Project, nitrates are recovered and the regeneration waste is apt to be use as fertilizer, for agricultural uses, or as complementary contribution of nutrients in biological water treatment. (Author) 27 refs.

  17. Hot dewatering and resin encapsulation of intermediate level radioactive waste

    International Nuclear Information System (INIS)

    Rickman, J.; Birch, D.

    1985-01-01

    The chemistry of the processes involved in the hot dewatering and encapsulation of alumino-ferric hydroxide floc in epoxide resin have been studied. Pretreatment of the floc to reduce resin attack and hydrolysis and to increase the dimensional stability of the solidified wasteform has been evaluated. It has been demonstrated that removal of ammonium nitrate from the floc and control of the residual water in the resin are important factors in ensuring dimensional stability of the solidified resin. Resin systems have been identified which, together with the appropriate waste pretreatment have successfully encapsulated a simulated magnox sludge producing a stable wasteform having mechanical and physical properties comparable with the basic resin. (author)

  18. Investigations on cement/polymer Waste packages containing intermediate level waste and organic exchange resins

    Energy Technology Data Exchange (ETDEWEB)

    ELsourougy, M R; Zaki, A A; Aly, H F [Atomic energy authority, hot laboratory center, Cairo, (Egypt); Khalil, M Y [Nuclear engineering department, Alexandria university. Alexandria, (Egypt)

    1995-10-01

    Polymers can be added to cements to improve its nuclear waste immobilization properties. This trend in cementation processes is attracting attention and requiring through investigations. In this work, polymers of different kinds were added to ordinary portland cement for the purpose of solidifying intermediate level liquid wastes and organic ion exchange resins. Epoxy polymer such as Kemapoxy-150 reduced the leaching rate of cesium compared to cement alone. Latex to cement ratio less than 4% caused an increase in leaching rate of cesium. When cesium was absorbed to an organic resin its leachability was improved. 5 figs., 4 tabs.

  19. Investigations on cement/polymer Waste packages containing intermediate level waste and organic exchange resins

    International Nuclear Information System (INIS)

    ELsourougy, M.R.; Zaki, A.A.; Aly, H.F.; Khalil, M.Y.

    1995-01-01

    Polymers can be added to cements to improve its nuclear waste immobilization properties. This trend in cementation processes is attracting attention and requiring through investigations. In this work, polymers of different kinds were added to ordinary portland cement for the purpose of solidifying intermediate level liquid wastes and organic ion exchange resins. Epoxy polymer such as Kemapoxy-150 reduced the leaching rate of cesium compared to cement alone. Latex to cement ratio less than 4% caused an increase in leaching rate of cesium. When cesium was absorbed to an organic resin its leachability was improved. 5 figs., 4 tabs

  20. Research on Recycling Mixed Wastes Based on Fiberglass and Organic Resins

    Science.gov (United States)

    Platon, M. A.; Ştef, M.; Popa, C.; Tiuc, A. E.; Nemeş, O.

    2018-06-01

    In recycling, according to principles of Directive 2008/98/EC of the European Parliament and of the Council on waste, research is upheld for achieving innovative technologies for reuse and keep as long it is possible, in economic chain, a waste. The aim of this research is to study and test a new composite material based on fiberglass waste mixed with organic resins with large application in the industry but not limited to this. Fiberglass is a material widely used for reinforcement of composite materials. As waste, fiberglass was less studied for ways to be reused. Filling fiberglass mixed with organic resins as PMMA and epoxy resins possess proper physical features for thermoforming. Three mixes are studied: fiberglass with PMMA, fiberglass with PMMA and rubber granules or sawdust. Samples will be tested for to define the mechanical and chemical behavior to have a complete description of the material. Analyzing the results can be concluded that mixes are suitable for board production, with improved features, compared with equivalent products on the market.

  1. Effect of ingredients in waste water on property of ion exchange resin for uranium-contained waste water treatment

    International Nuclear Information System (INIS)

    Ren Junshu; Mu Tao; Zhang Wei; Yang Shengya

    2008-01-01

    The effect of ingredients in waste water on the property of ion exchange resin for uranium-contained waste water treatment was studied by the method of static ad- sorption combined with dynamic experiment. The experimental result shows that the efficiency or breackthrough volume of resin is reduced if there are other general anions, triethanolamine and oil in the solution. When the concentrations of CO 3 2- , HCO 3 - , SO 3 2- , Cl - in the solution are more than 0.24, 0.28, 0.23 and 0.09 mol/L, respectively, the concentrations of uranium in the outlet waste water will exceed 20 μg/L. The maximal allowable concentration of triethanolamine through the resin is no more than 250 mg/L. When the content of oil in the resin exceeds 1%(by quality), the breackthrough volume reduces by 16%, and when it exceeds 11%, the breackthrough volume almost loses at all. (authors)

  2. Immobilization of spent resin with epoxy resin

    International Nuclear Information System (INIS)

    Gultom, O.; Suryanto; Sayogo; Ramdan

    1997-01-01

    immobilization of spent resin using epoxy resin has been conducted. The spent resin was mixtured with epoxy resin in variation of concentration, i.e., 30, 40, 50, 60, 70 weight percent of spent resin. The mixture were pour into the plastic tube, with a diameter of 40 mm and height of 40 mm. The density, compressive strength and leaching rate were respectively measured by quanta chrome, paul weber apparatus and gamma spectrometer. The results showed that the increasing of waste concentration would be decreased the compressive strength, and increased density by immobilized waste. The leaching rate of 137 Cs from waste product was not detected in experiment (author)

  3. Design and manufacturing concrete cells for shielding and storing radioactive semi liquid waste (resin) from MPR-GAS

    International Nuclear Information System (INIS)

    Pudjijanto-MS; Bahdir-Johan

    2003-01-01

    Semi liquid or quasi solid waste on Multipurpose Reactor G.A. Siwabessy (MPR-GAS) produced from operating resin rinsing systems and resin disposal systems during changes insert trap resin. Volume of the disposal resin waste in the filter mixed-bed per operation rinsing period are approx. 1.00 m 3 (in the Primary Cooling Water Treatment System) with activity ∼ 18.6 Ci/m 3 (0.688 TBq/m 3 ), 0.50 m 3 (in the Radioisotope Storage Pool Water Treatment System) with activity approx ∼ 0.162 Ci/m 3 (5.99 x 10 3 MBq/m 3 ) and 0.50 m 3 (in the Interim Spent Fuel Storage Pool Water Treatment System) with activity ∼ 0.162 Ci/m 3 (5.99 x 10 3 MBq/m 3 ) respectively. On the discharging and unloading, the gross radioactivity concentration of the resin waste loaded in the disposal resin waste tank are approx. 10 Ci/m 3 (0.37 TBq/m 3 ). After 6 months delayed, this activity is still 0.32 Ci/m 3 (11.84 GBq/m 3 ). Based on this data, some concrete cells to storage resin waste as semi liquid or quasi solid waste produced continuously by MPR-GAS installation has been designed and manufactured eternally

  4. Resin regenerating device in condensate desalting system

    International Nuclear Information System (INIS)

    Sato, Yoshiaki; Igarashi, Hiroo; Oosumi, Katsumi; Nishimura, Yusaku; Ebara, Katsuya; Shindo, Norikazu.

    1984-01-01

    Purpose: To improve the accuracy in the separation of anionic and cationic exchange resins. Constitution: Resins transferred from a condensate desalting column are charged in a cationic exchange resin column. The temperature of water for separating and transferring the resins is measured by a temperature detector disposed in a purified water injection line, and water is adjusted to a suitable flow rate for the separation and transfer of the resins by an automatic flow rate control valve, and then is injected. The resins are separated into cationic exchange resins and anionic exchange resins, in which only the anionic exchange resins are transferred, through an anionic exchange transfer line, into an anionic exchange resin column. By controlling the flow rate depending on the temperature of the injected water, the developing rate of the resin layer is made constant to enable separation and transfer of the resins at high accuracy. (Seki, T.)

  5. Bituminization of simulated waste, spent resins, evaporator concentrates and animal ashes by extrusion process

    International Nuclear Information System (INIS)

    Grosche Filho, C.E.; Chandra, U.

    1986-01-01

    The results of the study of simulated radwaste, spent ion-exchange resins, borates/evaporator-concentrates and animal ashes, in bituminized form, are presented and discussed. Distilled and oxidized bitumen were used for characterizing the crude material and simulated wastes-bitumen mixtures of varying weight composition 30, 40, 50, 60% by weight the dry waste material. The asphaltine and parafin contents in the bitumens were determined. Some additives and clays were used aiming best characteristics of solidified wastes. For leaching studies, granular ion-exchange resins were loaded with Cs 134 and mixtures of resins-bitumens were prepared. The leaching studies were executed using the IAEA recommendation and the ISO method. It was used a conventional screw-extruder, used in plastic industry, to determine operational conditions and process difficulties. Mixtures resins-bitumen and concentrate-bitumen in differents operational condition were prepared and analysed. (Author) [pt

  6. Literature search on the use of resins for treatment of radioactive wastes

    International Nuclear Information System (INIS)

    AlMahamid, I.; Smith, B.M.

    1997-10-01

    Over 100 commercial providers with mixed-waste treatability capabilities exist in the US. The maturity level of these technologies varies from a bench scale to a pilot or a commercial scale. The techniques include deactivation, chemical oxidation, recovery of metals, stabilization, vitrification, incineration, biodegradation, and chemical extraction. This report focuses on the use of resins to remove actinides and heavy metals from aqueous waste streams. Only the literature that described resins with high removing efficiency are presented here. The majority of the literature reviewed are proceedings and national or international reports ordered through the Berkeley Lab Library. Some of the reports that the authors requested have not yet arrived. Only a few papers were found in the open literature (journals or magazines). Although this report does not include all existing references, it provides an accurate assessment of efficient resins to be considered for waste minimization procedures. 70 refs

  7. Bituminization of simulated waste, spent resins, evaporator concentrates and animal ashes by extrusion process

    International Nuclear Information System (INIS)

    Grosche Filho, C.E.; Chandra, U.

    1987-01-01

    The results of the study of bituminization of simulated radwaste - spennt ion-exchange resins, borate evaporator/concentrates and animal ashes, are presented and discussed. Distilled and oxidizer bitumen were used. Characterization of the crude material and simulated wastes-bitumen mixtures of varying weigt composition (30, 40, 50, 60% by weight of dry waste material) was carried out. The asphaltene and parafin contents in the bitumens were also determined. Some additives and were used with an aim to improve the characteristcs of solidified wastes. For leaching studies, granular ion-exchange resins were with Cs - 134 and mixtures of resin-bitumen were prepared. The leaching studies were executed using the IAEA recommendation and the ISO method. A conventional screw-extruder, common in plastic industry, was used determine operational parameters and process difficulties. Mixtures of resin-bitumen and evaporator concentrate-bitumen obtained from differents operational conditions were characterized. (Author) [pt

  8. Ion exchange removal of cesium from Hanford tank waste supernates with SuperLigR 644 resin

    International Nuclear Information System (INIS)

    Hassan, N.M.; McCabe, D.J.; King, W.D.; Hamm, L.L.

    2002-01-01

    SuperLig R 644 ion exchange resin is currently being evaluated for cesium ( 137 Cs) removal from radioactive Hanford tank waste supernates as part of the River Protection Project. Testing was performed with actual Hanford tank wastes of widely different compositions using two identical ion exchange columns connected in series each containing approximately 5.5-6.5 ml of SuperLig R 644 resin. The ion exchange columns utilized the same resin material that was eluted between the column tests. This was done to demonstrate the performance of the SuperLig R 644 resin for cesium removal from waste samples of different compositions, determine the loading and elution profiles, and to validate design assumptions for full-scale column performances. Decontaminated product solutions generated at the same operating temperature and constant residence times (bed volumes per hour) exhibited the same chemical compositions as their feed samples. The compositions of eluate solutions were generally as expected with the exception of uranium and total organic carbon, which where concentrated by the resin. Development of a pretreatment method for the SuperLig R 644 resin has been critical to successful column operation with different waste solutions. (author)

  9. Application of nanoemulsions in the regeneration of adsorbent polymeric resins; Emprego de nanoemulsoes na regeneracao de resinas polimericas adsorvedoras

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Josane A.; Queiros, Yure G.C.; Vieira, Helida V.P.; Lucas, Elizabete F.; Mansur, Claudia R.E. [Universidade Federal do Rio de Janeiro (IMA/UFRJ), RJ (Brazil). Inst. de Macromoleculas. Lab. de Macromoleculas e Coloides na Industria de Petroleo], e-mails: josaneacosta@yahoo.com.br, yuregomes@ima.ufrj.br, elucas@ima.ufrj.br, celias@ima.ufrj.br

    2011-07-01

    In this work, the solbrax/water/polyoxyethylene nanoemulsions were produced under high pressure homogenizer (HPH). These systems presented droplets diameter ranging between 7 to 30 nm. The nanoemulsions were used in regeneration of a polymeric resin, which has been used in treatment of oily water. This resin was contaminated with different kinds of oils. The nanoemulsions presented high cleaning efficiency, above of 90% and its performance was higher than equivalent micellar systems. (author)

  10. Pre design processing of waste of ex-resin without materials matrix from nuclear power plant type PWR 1000 MW

    International Nuclear Information System (INIS)

    Cerdas Tarigan

    2010-01-01

    Have been done pre design processing of waste ex-resin without capacities matrix materials from nuclear power plant type PWR 1000 MW During the time radioactive waste of ex-resin processed to use process of immobilization use matrix materials like mixture cement and epoxy resin and then conditioning. This process is not effective and efficient because end result volume of end product bigger than volume early operation system and maintenance of its installation more difficult. To overcome this created a design of technology processing of waste of ex- resin without matrix materials through process of strainer, drying and conditioning represent technological innovation newly processing of radioactive waste of ex-resin. Besides this process more effective and efficient, volume of end product waste much more small from volume early and operation system and maintenance of its easier installation. Pre design is expected to be used as a basis to make conceptual of pre design installation of strainer, drying and conditioning for the processing of waste of ex-resin from nuclear power plant type PWR 1000 MW. (author)

  11. Study on dehydrochlorination of waste poly (vinyl chloride) resins by microwave irradiation

    Science.gov (United States)

    Moriwaki, Saburo; Qian, Qingrong; Sunohara, Satoshi; Machida, Motoi; Tatsumoto, Hideki

    Waste poly (vinyl chloride: PVC) resins are experimentally dehydrochlorinated by microwave irradiation. The following unique results are obtained: (1) plasticizer in PVC resin absorbs microwave power more effectively than PVC polymer. The higher the plasticizer content in PVC resin, the higher is the dehydrochlorination reaction (2) low PVC polymer content materials such as cushion floor require high microwave irradiation power to secure a high dehydrochlorination yield, (3) calcium carbonate in PVC resin reacts with released hydrochloric acid gas and results calcium chloride during microwave irradiation, (4) additives in PVC resin strongly influence dehydrochlorination yield, (5) it is evidenced that the PVC copolymer is also dehydrochlorinated by microwave irradiation.

  12. Use of a phenolic-carboxylic acid cation resin in the treatment of low-level liquid waste at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Chilton, J.M.

    1981-09-01

    The loading capacity of CS-100 resin, using plant waste as feed, was found to be significantly reduced after 20 loading-elution cycles; one-fourth or less of the original capacity was retained after 30 cycles. No important differences were noted between an untreated column and a column that had been reconverted to the sodium form in the regeneration step. Omission of the sodium regeneration could not be adopted as a routine procedure because it produced a packing effect in the plant beds; however, reconversion to the sodium form is now achieved by using a stoichiometric amount of caustic rather than a 100% excess, as was previous practice. Since the distribution coefficients for calcium and strontium are about six times greater than that for cesium, no loss of 90 Sr would be expected while 137 Cs is loading. Laboratory results obtained by using plant conditions and feed indicate that a typical bed would remove 96% of 90 Sr in the feed. Cobalt-60 is generally the greatest contributor to the radioactivity of the plant effluent. Laboratory tests indicate that this 60 Co is present as a mixture of a soluble anionic complex and insoluble colloids. The anionic complex could be removed by placing an anion exchange column in the effluent from the CS-100 resin bed. In studies of the dynamics of loading on CS-100 resin, the contact time in plant operation (3 to 4 min per column volume) was found to be more than adequate to obtain the desired results. Effects of flow velocity were not investigated. Data from a series of laboratory experiments show that CS-100 resin can be eluted satisfactorily with 0.5 to 1.0 M formic or acetic acid, although a larger volume is required than for elution with 0.5 M nitric acid

  13. Organic resin anion exchangers for the treatment of radioactive wastes

    International Nuclear Information System (INIS)

    Dyer, A.; McGinnes, D.F.

    1988-07-01

    Organic anion exchange resins are evaluated for 99-TcO 4 - (pertechnate) removed from aqueous nuclear waste streams. Chemical, thermal and radiation stabilities were studied. Selected resins were examined in detail for their selectivities in the presence of I - , NO 3 - , SO 4 = , CO 3 = , Cl - and OH - . Ion exchange equilibria and kinetic mechanisms were determined. Preliminary investigations of cement encapsulation in polymer modified form were made and some leach studies carried out. (author)

  14. Solidification of radioactive wastes with thermosetting resin

    International Nuclear Information System (INIS)

    Hayashi, M.; Kobayashi, K.; Okamoto, O.; Kagawa, T.; Wakamatsu, K.; Irie, H.; Matsuura, H.; Yasumura, K.; Nakayama, Y.

    1982-01-01

    Dried simulated radioactive wastes were solidified with thermosetting resin and their properties were investigated with laboratory scale and real scale products through extensive testings, such as mechanical resistance, resistance to leaching and swelling in water, radiation resistance, fire resistance and resistance to temperature cycling. The typical results were as follows: over 600 kg/cm 2 of compressive strength, diffusion constant of approx. 10 - 5 cm 2 /day for 137 Cs leaching from solidified waste products, no significant change was found for up to 5 x 10 8 RAD irradiation, and damages were limited to the surface of the products after the thermal test and dropping impact test. 7 figures, 4 tables

  15. Waste water treatment of CO2+O2 in-situ leaching uranium

    International Nuclear Information System (INIS)

    Xu Lechang; Liu Naizhong; Du Zhiming; Wang Hongying

    2012-01-01

    An in-situ leaching uranium mine located in Northern China uses CO 2 +O 2 leaching process to leach uranium. The consumption of industrial reagent and water, and generation and discharge of waste water are minimized by comprehensive waste water treatment technology with process water recycle, reverse osmosis and natural evaporation. The process water of the mine that can be recycled and reused includes barren fluid, solution washing loaded resin, precipitating mother solution and filtered liquor of yellow cake. Solution regenerating barren resin is treated by reverse osmosis. Concentrated water from reverse osmosis and solution washing barren resin are naturally evaporated. (authors)

  16. Treatment of Simulated Soil Decontamination Waste Solution by Ferrocyanide-Anion Exchange Resin Beads

    Energy Technology Data Exchange (ETDEWEB)

    Won, Hui Jun; Kim, Min Gil; Kim, Gye Nam; Jung, Chung Hun; Park, Jin Ho; Oh, Won Zin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2005-03-15

    Preparation of ferrocyanide-anion exchange resin and adsorption test of the prepared resin on the Cs{sup -} ion were performed. Adsorption capability of the prepared resin on the Cs{sup -} ion in the simulated citric acid based soil decontamination waste solution was 4 times greater than that of the commercial cation exchange resin. Adsorption equilibrium of the prepared resin on the Cs{sup -} ion reached within 360 minutes. Adsorption capability on the Cs{sup -} ion became to decrease above the necessary Co{sup 2-} ion concentration in the experimental range. Recycling test of the spent ion exchange resin by the successive application of hydrogen peroxide and hydrazine was also performed. It was found that desorption of Cs{sup -} ion from the resin occurred to satisfy the electroneutrality condition without any degradation of the resin.

  17. A basic study for the boron thermal regeneration system using anion exchange resins

    International Nuclear Information System (INIS)

    Frantiesek, P.; Kotaka, Masahiro; Okamoto, Makoto; Kakihana, Hidetake.

    1979-01-01

    For the boron thermal regeneration system (BTRS), the basic characteristics of commercial anion exchange resin have been investigated on the swelling characteristics, absorption, desorption and temperature coefficient of exchange capacity for boric acid. The equilibrium capacity increases as decrease of temperature and depends strongly on the degrees of cross linking having a maximum point at about 7% of DVB. The temperature coefficient of equilibrium capacity of boric acid is also a function of the concentration of external solution and of the cross linking having a maximum point around 7% of DVB. (author)

  18. Conditioning radioactive wastes by means of thermosetting resins

    International Nuclear Information System (INIS)

    Auboing, G.; Limongi, A.; Thiery, D.

    1976-01-01

    The principle of the conditioning process of low and medium activity wastes by means of thermosetting resins is described. The two major phases of its application, viz: pre-treatment and coating are analysed. Finally as an example, the plant where this conditioning process is put into application (currently in operation at the Nuclear Study Center of Grenoble) is described [fr

  19. Cement waste form development for ion-exchange resins and fine particles ILW of AREVA La Hague Reprocessing Plant

    International Nuclear Information System (INIS)

    Chartier, D.; Sanchez-Canet, J.; Avril, D.; Roussel, C.; Pineau, J.N.

    2015-01-01

    Wastes have been temporarily stored in dedicated silos in La Hague reprocessing plant. These wastes are to be retrieved in the near future and to be conditioned for final disposal. Some of these wastes are supposed to be encapsulated in cement matrix and, depending on the chemical composition of the waste streams, several projects are presently ongoing. The present article aims to focus on one amongst these cement encapsulation relevant projects, namely the conditioning of a mix of spent ion-exchange resins (from filtration of pool) and fine particles (insoluble fission products from spent fuel dissolution and Zircaloy and stainless steel fines from cladding shearing). The project, aims to retrieve these wastes from a silo, separate the resins and fine particles from the other waste (hulls and end pieces), in order to encapsulate the intermediate-level fines and resins in a cement matrix. The waste forms will be produced in AREVA's La Hague reprocessing plant, prior to being sent as intermediate-level waste to a long-term repository. The cement formulation developments were initially carried out at a small scale at C.E.A. Marcoule on surrogate wastes. One of the main issues that were considered was the chemical compatibility between waste and cement matrix. Indeed, swelling phenomena are sometimes reported when ion exchange resins are embedded in cement matrixes such as Portland cement. This kind of destructive phenomenon has been prevented by the use of cement containing a high amount of ground granulated blast furnace slag. The impact of the variability of ionic charge of the resins on the waste form's properties has also been addressed in order to comfort the results obtained on the reference ionic charge of resins NaNO 3 . Once the results obtained were satisfactory, intermediate scale and full scale tests were performed by AREVA. These tests have focused on adjusting the mixing process and controlling the thermal properties of the mix during setting

  20. New ion exchange resin designs and regeneration procedures yield improved performance for various condensate polishing applications

    International Nuclear Information System (INIS)

    Najmy, S.W.

    2002-01-01

    Condensate polishing is an application with many different design and operational aspects. The past decade has brought new challenges for improved water quality with respect to both soluble and insoluble contaminants. Nonetheless, the endeavors to understand the compositional complexities of the ion exchange resin bead and the convoluted dynamics of ion exchange chemistry and chemical engineering mechanisms occurring within the mixed bed condensate polisher have brought new ideas and expectations for ion exchange resin in deep-bed condensate polishers than ever before. The new products and procedures presented here are a collaboration of a great deal of effort on the part of researchers, consultants, system engineers, station chemists, lab technicians and others. The studies discussed in this paper unequivocally demonstrate the merits of: 1. A specially designed cation resin to achieve greater than 95% insoluble iron removal efficiency, 2. A less-separable mixed resin for improved control of reactor water sulfate in BWR primary cycles, 3. Applying increased levels of regeneration chemicals and retrofitting the service vessels with re-mixing capability to improve the operation of deep-bed condensate polishers in PWR secondary cycles. (authors)

  1. Ion Exchange Modeling Of Cesium Removal From Hanford Waste Using Spherical Resorcinol-Formaldehyde Resin

    Energy Technology Data Exchange (ETDEWEB)

    Aleman, S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hamm, L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Smith, F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2007-06-27

    This report discusses the expected performance of spherical Resorcinol-Formaldehyde (RF) ion exchange resin for the removal of cesium from alkaline Hanford radioactive waste. Predictions of full scale column performance in a carousel mode are made for the Hot Commissioning, Envelope B, and Subsequent Operations waste compositions under nominal operating conditions and for perturbations from the nominal. Only the loading phase of the process cycle is addressed in this report. Pertinent bench-scale column tests, kinetic experiments, and batch equilibrium experiments are used to estimate model parameters and to benchmark the ion-exchange model. The methodology and application presented in this report reflect the expected behavior of spherical RF resin manufactured at the intermediate-scale (i.e., approximately 100 gallon batch size; batch 5E-370/641). It is generally believed that scale-up to production-scale in resin manufacturing will result in similarly behaving resin batches whose chemical selectivity is unaffected while total capacity per gram of resin may vary some. As such, the full-scale facility predictions provided within this report should provide reasonable estimates of production-scale column performance.

  2. Free Moisture in GT-73 Resin Waste Generated from the Effluent Treatment Facility

    International Nuclear Information System (INIS)

    Kaplan, D.I.

    2002-01-01

    Solid Waste Division is presently evaluating whether to transfer spent resin generated from the Effluent Treatment Facility (ETF) to the Nevada Test Site (NTS). One of the criteria for the waste to be accepted at the NTS is that the waste must not contain more than 1 vol-percent free liquid. This criterion reduces the amount of liquid, a primary vector for subsurface contaminant migration (along with colloids), introduced into the repository. This criterion also serves to reduce the chance of an accidental spill during transport of the waste to the NTS. On December 15, 1997, a shipment from Fernald to the NTS leaked some liquid waste onto a highway in Kingman, Arizona, resulting in a Type B Accident Investigation. The direct cause of the leak was attributed to broken welds related to the use of substandard containers. The overall objective of this study was to provide guidance as to whether the spent GT-73 resin would meet the free moisture WAC set by the NTS

  3. Analysis of Ion-Exchange Resin Capability of the RSG-GAS Demineralized Water System (GCA01)

    International Nuclear Information System (INIS)

    Diyah Erlina Lestari; Setyo Budi Utomo; Harsono

    2012-01-01

    The Demineralized water system (GCA01) is a system which is function to process raw water to be demineralized water using ion exchange resin unit consisting of a column of cation exchange resins, anion exchange resin column and the column resin mix bed. After certain time the ion exchange resins to be saturated so that is needed regeneration. The RSG-GAS demineralized water system (GCA01) not operated continuously and indication of when does an ion exchange resin regeneration on The RSG-GAS demineralized water system (GCA01) is the water conductivity from anion exchange resin column output indicates ≥ 5μS/cm. Analysis of capability of the ion exchange resin demineralized water system (GCA01) line I has been performed. The analysis was done by comparing the time required in the system operating cycle of regeneration to the next regeneration during the period 2011 and 2012. From the results of the analysis showed the cycle regeneration time is varies. This shows that ion exchange resin capability of the RSG-GAS demineralized water system (GCA01) is varies depending on the raw water quality and success of the regeneration ion exchange resin. (author)

  4. Encapsulation pilot plant of radioactive wastes in thermosetting resins

    International Nuclear Information System (INIS)

    1982-01-01

    The thermosetting resins (polyesters, epoxides) are used to encapsulate the low and intermediate - level radioactive wastes. The testing program concerning the drums produced by the pilot plant of the Chooz nuclear power plant is described. The installation operating is examined while thinking of the industrial application. The production costs are then evaluated

  5. Radioactive waste processing device

    International Nuclear Information System (INIS)

    Ikeda, Takashi; Funabashi, Kiyomi; Chino, Koichi.

    1992-01-01

    In a waste processing device for solidifying, pellets formed by condensing radioactive liquid wastes generated from a nuclear power plant, by using a solidification agent, sodium chloride, sodium hydroxide or sodium nitrate is mixed upon solidification. In particular, since sodium sulfate in a resin regenerating liquid wastes absorbs water in the cement upon cement solidification, and increases the volume by expansion, there is a worry of breaking the cement solidification products. This reaction can be prevented by the addition of sodium chloride and the like. Accordingly, integrity of the solidification products can be maintained for a long period of time. (T.M.)

  6. Novel silica-based ion exchange resin

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    Eichrom`s highly successful Diphonixo resin resembles a conventional ion exchange resin in its use of sulfonic acid ligands on a styrene- divinylbenzene matrix. Diphonix resin exhibits rapid exchange kinetics that allow economical operation of ion exchange systems. Unlike conventional resins, Diphonix resin contains chelating ligands that are diphosphonic acid groups that recognize and remove the targeted metals and reject the more common elements such as sodium, calcium and magnesium. This latter property makes Diphonix ideal for many industrial scale applications, including those involving waste treatment. For treatment of low-level, transuranic (TRU) and high- level radioactive wastes, Diphonix`s polystyrene backbone hinders its application due to radiolytic stability of the carbon-hydrogen bonds and lack of compatibility with expected vitrification schemes. Polystyrene-based Diphonix is approximately 60% carbon- hydrogen. In response to an identified need within the Department of Energy for a resin with the positive attributes of Diphonix that also exhibits greater radiolytic stability and final waste form compatibility, Eichrom has successfully developed a new, silica-based resin version of Diphonix. Target application for this new resin is for use in environmental restoration and waste management situations involving the processing of low-level, transuranic and high-level radioactive wastes. The resin can also be used for processing liquid mixed waste (waste that contains low level radioactivity and hazardous constituents) including mixed wastes contaminated with organic compounds. Silica-based Diphonix is only 10% carbon-hydrogen, with the bulk of the matrix silica.

  7. Ion-exchange-resin-catalyzed adamantylation of phenol derivatives with adamantanols: Developing a clean process for synthesis of 2-(1-adamantyl-4-bromophenol, a key intermediate of adapalene

    Directory of Open Access Journals (Sweden)

    Nan Wang

    2012-02-01

    Full Text Available A clean process has been developed for the synthesis of 2-adamantylphenol derivatives through adamantylation of substituted phenols with adamantanols catalyzed by commercially available and recyclable ion-exchange sulfonic acid resin in acetic acid. The sole byproduct of the adamantylation reaction, namely water, could be converted into the solvent acetic acid by addition of a slight excess of acetic anhydride during the work-up procedure, making the process waste-free except for regeneration of the ion-exchange resin, and facilitating the recycling of the resin catalyst. The ion-exchange sulfonic acid resin catalyst could be readily recycled by filtration and directly reused at least ten times without a significant loss of activity. The key intermediate of adapalene, 2-(1-adamantyl-4-bromophenol, could be produced by means of this waste-free process.

  8. A Novel Ion Exchange System to Purify Mixed ISS Waste Water Brines for Chemical Production and Enhanced Water Recovery

    Science.gov (United States)

    Lunn, Griffin; Spencer, LaShelle; Ruby, Anna-Maria; McCaskill, Andrew

    2014-01-01

    Current International Space Station water recovery regimes produce a sizable portion of waste water brine. This brine is highly toxic and water recovery is poor: a highly wasteful proposition. With new biological techniques that do not require waste water chemical pretreatment, the resulting brine would be chromium-free and nitrate rich which can allow possible fertilizer recovery for future plant systems. Using a system of ion exchange resins we can remove hardness, sulfate, phosphate and nitrate from these brines to leave only sodium and potassium chloride. At this point modern chlor-alkali cells can be utilized to produce a low salt stream as well as an acid and base stream. The first stream can be used to gain higher water recovery through recycle to the water separation stage while the last two streams can be used to regenerate the ion exchange beds used here, as well as other ion exchange beds in the ISS. Conveniently these waste products from ion exchange regeneration would be suitable as plant fertilizer. In this report we go over the performance of state of the art resins designed for high selectivity of target ions under brine conditions. Using ersatz ISS waste water we can evaluate the performance of specific resins and calculate mass balances to determine resin effectiveness and process viability. If this system is feasible then we will be one step closer to closed loop environmental control and life support systems (ECLSS) for current or future applications.

  9. Solidification of radioactive liquid wastes, Treatment options for spent resins and concentrates - 16405

    International Nuclear Information System (INIS)

    Roth, Andreas

    2009-01-01

    Ion exchange is one of the most common and effective treatment methods for liquid radioactive waste. However, spent ion exchange resins are considered to be problematic waste that in many cases require special approaches and pre-conditioning during its immobilization to meet the acceptance criteria for disposal. Because of the function that they fulfill, spent ion exchange resins often contain high concentrations of radioactivity and pose special handling and treatment problems. Another very common method of liquid radioactive waste treatment and water cleaning is the evaporation or diaphragm filtration. Both treatment options offer a high volume reduction of the total volume of liquids treated but generate concentrates which contain high concentrations of radioactivity. Both mentioned waste streams, spent resins as well as concentrates, resulting from first step liquid radioactive waste treatment systems have to be conditioned in a suitable manner to achieve stable waste products for final disposal. Spent resin and concentrate treatment often appear as a specific task in decommissioning projects, because in the past those waste streams typically had been stored in tanks for the lifetime of the plant and needs to be retrieved, conditioned and packed prior to dismantling activities. Additionally a large amount of contaminated liquids will be generated by utilizing decontamination processes and needs to be processed further on. Such treatment options need to achieve waste products acceptable for final disposal, because due to the closure of the site no interim storage can be envisaged. The most common method of treatment of such waste streams is the solidification in a solid matrix with additional inactive material like cement, polymer etc. In the past good results have been achieved and the high concentration of radioactivity can be reduced by adding the inactive material. On the other hand, under the environment of limited space for interim storage and the absence

  10. Sustainable nitrate-contaminated water treatment using multi cycle ion-exchange/bioregeneration of nitrate selective resin.

    Science.gov (United States)

    Ebrahimi, Shelir; Roberts, Deborah J

    2013-11-15

    The sustainability of ion-exchange treatment processes using high capacity single use resins to remove nitrate from contaminated drinking water can be achieved by regenerating the exhausted resin and reusing it multiple times. In this study, multi cycle loading and bioregeneration of tributylamine strong base anion (SBA) exchange resin was studied. After each cycle of exhaustion, biological regeneration of the resin was performed using a salt-tolerant, nitrate-perchlorate-reducing culture for 48 h. The resin was enclosed in a membrane to avoid direct contact of the resin with the culture. The results show that the culture was capable of regenerating the resin and allowing the resin to be used in multiple cycles. The concentrations of nitrate in the samples reached a peak in first 0.5-1h after placing the resin in medium because of desorption of nitrate from resin with desorption rate of 0.099 ± 0.003 hr(-1). After this time, since microorganisms began to degrade the nitrate in the aqueous phase, the nitrate concentration was generally non-detectable after 10h. The average of calculated specific degradation rate of nitrate was -0.015 mg NO3(-)/mg VSS h. Applying 6 cycles of resin exhaustion/regeneration shows resin can be used for 4 cycles without a loss of capacity, after 6 cycles only 6% of the capacity was lost. This is the first published research to examine the direct regeneration of a resin enclosed in a membrane, to allow reuse without any disinfection or cleaning procedures. Copyright © 2013 Elsevier B.V. All rights reserved.

  11. 'Like new': plastic wastes regeneration by radiation induced grafting

    International Nuclear Information System (INIS)

    Laizier, J.; Gaussens, G.; Lemaire, F.

    1978-01-01

    The reclaiming and the recycling of plastic wastes is made especially difficult when those wastes are a mixture of various plastics; this is due to the incompatibility of the polymers. The radiation induced grafting allows to overcome this incompatibility. Results are given which shows that, for various mixtures of reclaimed polyethylene, PVC and polystyrene, an improvement of the properties of the processed blends is obtained by grafting the mixtures of wastes by a suitable polymer; the obtained properties of those regenerated plastic blends are enough attractive from the technical point of view to open a market to those products with a reasonable economical value [fr

  12. Removal of cesium from aluminum decladding wastes generated in irradiated target processing using a fixed-bed column of resorcinol-formaldehyde resin

    International Nuclear Information System (INIS)

    Brunson, R.R.; Williams, D.F.; Bond, W.D.; Benker, D.E.; Chattin, F.R.; Collins, E.D.

    1994-09-01

    The removal of cesium (Cs) from a low-level liquid waste (LLLW) with a cation-exchange column was demonstrated using a resorcinol-formaldehyde (RF) resin. The RF resin was developed at the Westinghouse Savannah River Laboratory (SRL) and is highly specific for the removal of Cs from an alkaline waste of high sodium content. It was determined that the RF resin would be suitable for removing Cs, the largest gamma radiation contributor, from the LLLW generated at the Radiochemical Engineering Development Center located at the Oak Ridge National Laboratory. Presently, the disposal of the LLLW is limited due to the amount of Cs contained in the waste. Cesium removal from the waste solution offers immediate benefits by conserving valuable tank space and would allow cask shipments of the treated waste should the present Laboratory pipelines become unavailable in the future. Preliminary laboratory tests of the RF resins, supplied from two different sources, were used to design a full-scale cation-exchange column for the removal of Cs from a Mark 42 SRL fuel element dejacketing waste solution. The in-cell tests reproduced the preliminary bench-scale test results. The initial Cs breakthrough range was 85--92 column volumes (CV). The resin capacity for Cs was found to be ∼0.35 meq per gram of resin. A 1.5-liter resin bed loaded a combined ∼1,300 Ci of 134 Cs and 137 Cs. A distribution coefficient of ∼110 CV was determined, based on a 50% Cs breakthrough point. The kinetics of the system was studied by examining the rate parameters; however, it was decided that several more tests would be necessary to define the mass transfer characteristics of the system

  13. Method of solidifying radioactive ion exchange resin

    International Nuclear Information System (INIS)

    Minami, Yuji; Tomita, Toshihide

    1989-01-01

    Spent anion exchange resin formed in nuclear power plants, etc. generally catch only a portion of anions in view of the ion exchange resins capacity and most of the anions are sent while possessing activities to radioactive waste processing systems. Then, the anion exchange resins increase the specific gravity by the capture of the anions. Accordingly, anions are caused to be captured on the anion exchange resin wastes such that the specific gravity of the anion exchange resin wastes is greater than that of the thermosetting resins to be mixed. This enables satisfactory mixing with the thermosetting resins and, in addition, enables to form integral solidification products in which anion exchange resins and cation exchange resins are not locallized separately and which are homogenous and free from cracks. (T.M.)

  14. Casting granular ion exchange resins with medium-active waste in cement

    International Nuclear Information System (INIS)

    Beijer, O.

    1980-01-01

    Medium active waste from nuclear power stations in Sweden is trapped in granular ion exchange resins. The resin is mixed with cement paste and cast in a concrete shell which is cubic and has an edge dimension of 1.2 m. In some cases the ion exchange cement mortar has cracked. The report presents laboratory sutdies of the properties of the ion exchange resin and the mortar. Also the leaching of the moulds has been investigated. It was shown that a mixture with a water cement ratio higher than about 0.5 swells considerably during the first weeks after casting. The diffusion constant for cesium 137 has been determined at 3.10 -4 cm 2 /24-hour period in conjunction with exposure of the mould and mortar to sea water. The Swedish language report has 400 pages with 90 figures and 30 tables. (author)

  15. Development of radioactive waste treatment system for nuclear power stations by Toshiba (III)

    International Nuclear Information System (INIS)

    Irie, H.; Takahara, T.; Matsuda, T.; Matsuura, H.; Yasumura, K.; Nakayama, Y.

    1989-01-01

    This paper describes a solidification process with thermosetting resin to satisfy both requirements of volume reduction and quality of solidified products. Volumes of solidified products in drums generated from spent resins and concentrated wastes were reduced respectively to 1/4 and less than 1/6 of those in the conventional cement solidification process. In plants using a simple demineralizing system for condensate polishing, a large amount of waste water with regenerant chemicals is generated from the condensate demineralizer. In general, radioactivity concentration of wastes from this type of nuclear power plant is comparatively high, so the dose rate at the surface of drums containing solidified wastes exceeds 200mR/h. A pelletizing system for radioactive wastes was developed to reduce their volumes and allow their interim storage until the radioactivity decays down to a level at which they can be handled easily

  16. Cementation of secondary wastes generated from carbonisation of spent organic ion exchange resins from nuclear power plants

    International Nuclear Information System (INIS)

    Sathi Sasidharan, N.; Deshingkar, D.S.; Wattal, P.K.

    2004-07-01

    The spent IX resins containing radioactive fission and activation products from power reactors are highly active solid wastes generated during operations of nuclear reactors. Process for carbonization of IX resins to achieve weight and volume reduction has been optimized on 50 dm 3 /batch pilot test rig. The process generates carbonaceous residue, organic liquid condensates (predominantly styrene) and aqueous alkaline scrubber solutions as secondary wastes. The report discusses laboratory tests on leaching of 137 Cs from cement matrix incorporating carbonaceous residues and extrapolation of results to 200 liter matrix block. The cumulative fraction of 137 Cs leached from 200 liter cement matrix was estimated to be 0.0021 in 200 days and 0.0418 over a period of 30 years. Incorporation of organic liquid condensates into cement matrix has been tried out successfully. Thus two types of secondary wastes generated during carbonization of spent IX resins can be immobilized in cement matrix. (author)

  17. Solidification of radioactive liquid wastes. A comparison of treatment options for spent resins and concentrates

    International Nuclear Information System (INIS)

    Roth, A.; Willmann, F.; Ebata, M.; Wendt, S.

    2008-01-01

    Ion exchange is one of the most common and effective treatment methods for liquid radioactive waste. However, spent ion exchange resins are considered to be problematic waste that in many cases require special approaches and pre-conditioning during its immobilization to meet the acceptance criteria for disposal. Because of the function that they fulfill, spent ion exchange resins often contain high concentrations of radioactivity and pose special handling and treatment problems. Another very common method of liquid radioactive waste treatment and water cleaning is the evaporation or diaphragm filtration. Both treatment options offer a high volume reduction of the total volume of liquids treated but generate concentrates which contain high concentrations of radioactivity. Both mentioned waste streams, spent resins as well as concentrates, resulting from first step liquid radioactive waste treatment systems have to be conditioned in a suitable manner to achieve stable waste products for final disposal. The most common method of treatment of such waste streams is the solidification in a solid matrix with additional inactive material like cement, polymer etc. In the past good results have been achieved and the high concentration of radioactivity can be reduced by adding the inactive material. On the other hand, under the environment of limited space for interim storage and the absence of a final repository site, the built-up of additional volume has to be considered as very critical. Moreover, corrosive effects on cemented drums during long-term interim storage at the surface have raised doubts about the long-term stability of such waste products. In order to avoid such disadvantages solidification methods have been improved in order to get a well-defined product with a better load factor of wastes in the matrix. In a complete different approach, other technologies solidify the liquid radioactive wastes without adding of any inactive material by means of drying

  18. Enhanced DOC removal using anion and cation ion exchange resins.

    Science.gov (United States)

    Arias-Paic, Miguel; Cawley, Kaelin M; Byg, Steve; Rosario-Ortiz, Fernando L

    2016-01-01

    Hardness and DOC removal in a single ion exchange unit operation allows for less infrastructure, is advantageous for process operation and depending on the water source, could enhance anion exchange resin removal of dissolved organic carbon (DOC). Simultaneous application of cationic (Plus) and anionic (MIEX) ion exchange resin in a single contact vessel was tested at pilot and bench scales, under multiple regeneration cycles. Hardness removal correlated with theoretical predictions; where measured hardness was between 88 and 98% of the predicted value. Comparing bench scale DOC removal of solely treating water with MIEX compared to Plus and MIEX treated water showed an enhanced DOC removal, where removal was increased from 0.5 to 1.25 mg/L for the simultaneous resin application compared to solely applying MIEX resin. A full scale MIEX treatment plant (14.5 MGD) reduced raw water DOC from 13.7 mg/L to 4.90 mg/L in the treated effluent at a bed volume (BV) treatment rate of 800, where a parallel operation of a simultaneous MIEX and Plus resin pilot (10 gpm) measured effluent DOC concentrations of no greater than 3.4 mg/L, even at bed volumes of treatment 37.5% greater than the full scale plant. MIEX effluent compared to simultaneous Plus and MIEX effluent resulted in differences in fluorescence intensity that correlated to decreases in DOC concentration. The simultaneous treatment of Plus and MIEX resin produced water with predominantly microbial character, indicating the enhanced DOC removal was principally due to increased removal of terrestrially derived organic matter. The addition of Plus resin to a process train with MIEX resin allows for one treatment process to remove both DOC and hardness, where a single brine waste stream can be sent to sewer at a full-scale plant, completely removing lime chemical addition and sludge waste disposal for precipitative softening processes. Published by Elsevier Ltd.

  19. Commercial Ion Exchange Resin Vitrification Studies

    International Nuclear Information System (INIS)

    Cicero-Herman, C.A

    2002-01-01

    In the nuclear industry, ion exchange resins are used for purification of aqueous streams. The major contaminants of the resins are usually the radioactive materials that are removed from the aqueous streams. The use of the ion exchange resins creates a waste stream that can be very high in both organic and radioactive constituents. Therefore, disposal of the spent resin often becomes an economic problem because of the large volumes of resin produced and the relatively few technologies that are capable of economically stabilizing this waste. Vitrification of this waste stream presents a reasonable disposal alternative because of its inherent destruction capabilities, the volume reductions obtainable, and the durable product that it produces

  20. Efficiencies and Optimization of Weak Base Anion Ion-Exchange Resin for Groundwater Hexavalent Chromium Removal at Hanford

    International Nuclear Information System (INIS)

    Nesham, Dean O.; Ivarson, Kristine A.; Hanson, James P.; Miller, Charles W.; Meyers, P.; Jaschke, Naomi M.

    2014-01-01

    The U.S. Department of Energy's (DOE's) contractor, CH2M HILL Plateau Remediation Company, has successfully converted a series of groundwater treatment facilities to use a new treatment resin that is delivering more than $3 million in annual cost savings and efficiency in treating groundwater contamination at the DOE Hanford Site in southeastern Washington State. During the production era, the nuclear reactors at the Hanford Site required a continuous supply of high-quality cooling water during operations. Cooling water consumption ranged from about 151,417 to 378,541 L/min (40,000 to 100,000 gal/min) per reactor, depending on specific operating conditions. Water from the Columbia River was filtered and treated chemically prior to use as cooling water, including the addition of sodium dichromate as a corrosion inhibitor. Hexavalent chromium was the primary component of the sodium dichromate and was introduced into the groundwater at the Hanford Site as a result of planned and unplanned discharges from the reactors starting in 1944. Groundwater contamination by hexavalent chromium and other contaminants related to nuclear reactor operations resulted in the need for groundwater remedial actions within the Hanford Site reactor areas. Beginning in 1995, groundwater treatment methods were evaluated, leading to the use of pump-and-treat facilities with ion exchange using Dowex 21K, a regenerable, strong-base anion exchange resin. This required regeneration of the resin, which was performed offsite. In 2008, DOE recognized that regulatory agreements would require significant expansion for the groundwater chromium treatment capacity. As a result, CH2M HILL performed testing at the Hanford Site in 2009 and 2010 to demonstrate resin performance in the specific groundwater chemistry at different waste sites. The testing demonstrated that a weak-base anion, single-use resin, specifically ResinTech SIR-700 ®, was effective at removing chromium, had a significantly higher

  1. Physical and Mechanical Characterization of Artificial Stone with Marble Calcite Waste and Epoxy Resin

    OpenAIRE

    Silva, Fernanda Souza; Ribeiro, Carlos Eduardo Gomes; Rodriguez, Rubén Jesus Sánchez

    2017-01-01

    The incorporation of calcite marble waste in epoxy resin for the production of artificial stone can represent a technical-economical method and environmentally viable, reducing the amount of discarded residue in the environment, and adding economic value to marble waste and enabling the generation of jobs. The production of natural stone in Brazil recorded an exorbitant amount of waste generated in marble processing. Only 75% of marble taken from the deposits it becomes the finished product t...

  2. Towards reinforcement solutions for urban fibre/fabric waste using bio-based biodegradable resins.

    Science.gov (United States)

    Agrawal, Pramod; Hermes, Alina; Bapeer, Solaf; Luiken, Anton; Bouwhuis, Gerrit; Brinks, Ger

    2017-10-01

    The main research question is how to systematically define and characterize urban textile waste and how to effectively utilise it to produce reinforcement(s) with selected bio-based biodegradable resin(s). Several composite samples have been produced utilising predominantly natural and predominantly synthetic fibres by combining loose fibres with PLA, nonwoven fabric with PLA, woven fabric with PLA, two-layer composite & four-layer composite samples. Physio-chemical characterisations according to the established standards have been conducted. The present work is a step toward the circular economy and closing the loop in textile value chain.

  3. Resin for processing radioactive waste water

    International Nuclear Information System (INIS)

    Onozuka, Teruo; Shindo, Manabu; Kiba, Hideaki; Kubota, Hirohisa; Sawada, Shintaro.

    1995-01-01

    The present invention concerns an anionic exchange resin having a long service life with less radiation degradation. The resin is an anionic exchange resin in which a trimethyl ammonium group is introduced to a copolymer of 4-bromo-butoxymethyl styrene and divinyl benzene. The resin is excellent in economic performance, and can reduce the frequency for the exchange of cross-linked anionic exchangers. (T.M.)

  4. Resin for processing radioactive waste water

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, Teruo; Shindo, Manabu [Tohoku Electric Power Co., Inc., Sendai (Japan); Kiba, Hideaki; Kubota, Hirohisa; Sawada, Shintaro

    1995-11-07

    The present invention concerns an anionic exchange resin having a long service life with less radiation degradation. The resin is an anionic exchange resin in which a trimethyl ammonium group is introduced to a copolymer of 4-bromo-butoxymethyl styrene and divinyl benzene. The resin is excellent in economic performance, and can reduce the frequency for the exchange of cross-linked anionic exchangers. (T.M.).

  5. Composting of waste paint sludge containing melamine resin as affected by nutrients and gypsum addition and microbial inoculation

    International Nuclear Information System (INIS)

    Tian Yongqiang; Chen Liming; Gao Lihong; Michel, Frederick C.; Wan Caixia; Li Yebo; Dick, Warren A.

    2012-01-01

    Melamine formaldehyde resins have hard and durable properties and are found in many products, including automobile paints. These resins contain high concentrations of nitrogen and, if properly composted, can yield valuable products. We evaluated the effects of starter compost, nutrients, gypsum and microbial inoculation on composting of paint sludge containing melamine resin. A bench-scale composting experiment was conducted at 55 °C for 91 days and then at 30 °C for an additional 56 days. After 91 days, the composts were inoculated with a mixed population of melamine-degrading microorganisms. Melamine resin degradation after the entire 147 days of composting varied between 73 and 95% for the treatments with inoculation of microorganisms compared to 55–74% for the treatments without inoculation. Degradation was also enhanced by nutrients and gypsum additions. Our results infer that large scale composting of melamine resins in paint sludge is possible. - Highlights: ► Melamine resin in waste paint sludges could be efficiently composted at bench scale. ► Melamine resin degradation after 147 days of composting was 73–95% complete. ► Nutrients, gypsum and melamine-degrading microorganisms increased composting rate. ► Melamine degradation products first increased and then decreased in the compost. ► Final compost was enriched in nitrogen and other essential plant nutrients. - Melamine resin in waste paint sludges was efficiently composted at bench scale, with finished composts having low levels of heavy metals and enriched in plant nutrients.

  6. Development and testing of a method for coating metallic wastes by thermosetting resins

    International Nuclear Information System (INIS)

    Tassigny, C. de

    1985-01-01

    Decommissioning of nuclear installations produces large-sized metallic components difficult to be compacted or to be inserted in the usual containers. The purpose of this study was to develop a coating procedure for low and medium activity wastes, in order to: - fix the contamination of the waste by a first appropriate layer; - protect this layer from mechanical shocks by a second thick layer; - reduce the diffusion of radionuclides. The study has proven the feasibility of depositing epoxy resins by electrostatic spraying in a nuclear environment on steel sub-strata with an efficiency higher than 95%; showing that it is possible to work in closed ventilated rooms without risk of clogging the filters. Two layers of epoxy resin are sufficient to fix contamination with a factor of 8600; their thickness (about 30 μm) nevertheless limits their use to this application. To reduce diffusion of radionuclides (Co-60 and Cs-137), polyurethane resins were chosen with which it is possible to obtain a thick coating in only a short time. Their properties of fixing the contamination, retaining of radionuclides and resistance to impact, have been assessed. First application of the procedure on a contaminated metallic bellow coming from dismantling of nuclear reactor, has been carried out

  7. "We Was Regenerated Out": Regeneration, Recycling and Devaluing Communities

    Directory of Open Access Journals (Sweden)

    Luna Glucksberg

    2014-12-01

    Full Text Available This article looks at well documented processes of urban regeneration and community displacement in the inner-city through an innovative anthropological perspective focused on concepts of waste and value. Using the notion of symbolic devaluation of the working classes developed by Skeggs (1997; 2004, it traces their exclusion from recycling practices while at the same time the estates they live on are being regenerated. Raising questions about the parallels and contradictions between regeneration and recycling, it shows how symbolic devaluation of specifi c areas and their inhabitants are necessary precursors of the physical demolition and removal that characterize regeneration processes. Through an ethnographic approach, the deep connections between people and their waste, and people as waste, are exposed and questioned, showing how valuable middle class selves are produced through appropriate waste management procedures, i.e. individualized recycling, while inner-city, estate dwellers are remade into uncaring, unworthy citizens who cannot take part in this value-producing circuit.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-07

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

  10. Performances and improvement of copper-hydrazine complexation deoxidising resin

    International Nuclear Information System (INIS)

    Liu Fenfen; Zhang Hao; Sun Haijun; Liu Xiaojie

    2012-01-01

    Copper-hydrazine complexation deoxidising resin is tested to examine its performances including effluent water quality and capacity of deoxidisation. By the means of changing the resin type and regeneration, the deoxidising capacity of the resin can be improved to 13 times more than before. At the same time, physical performances of the resin are also greatly improved while maintaining its velocity of deoxidisation and effluent quality. (authors)

  11. PILOT-SCALE HYDRAULIC TESTING OF RESORCINOL FORMALDEHYDE ION EXCHANGE RESIN

    International Nuclear Information System (INIS)

    Adamson, D

    2007-01-01

    Savannah River National Laboratory (SRNL) performed pilot-scale hydraulic/chemical testing of spherical resorcinol formaldehyde (RF) ion exchange (IX) resin for the River Protection Project Hanford Tank Waste Treatment and Immobilization Plant (WTP) Project. The RF resin cycle testing was conducted in two pilot-scale IX columns, 1/4 and 1/2 scale. A total of twenty-three hydraulic/chemical cycles were successfully completed on the spherical RF resin. Seven of the cycles were completed in the 12-inch IX Column and sixteen cycles were completed in the 24-inch IX Column. Hydraulic testing showed that the permeability of the RF resin remained essentially constant, with no observed trend in the reduction of the permeability as the number of cycles increased. The permeability during the pilot-scale testing was 2 1/2 times better than the design requirements of the WTP full-scale system. The permeability of the resin bed was uniform with respect to changes in bed depth. Upflow Regeneration and Simulant Introduction in the IX columns revealed another RF resin benefit; negligible radial pressures to the column walls from the swelling of resin beads. In downflow of the Regeneration and Simulant Introduction steps, the resin bed particles pack tightly together and produce higher hydraulic pressures than that found in upflow. Also, upflow Simulant Introduction produced an ideal level bed for the twenty cycles completed using upflow Simulant Introduction. Conversely, the three cycles conducted using downflow Simulant Introduction produced an uneven bed surface with erosion around the thermowells. The RF resin bed in both columns showed no tendency to form fissures or pack more densely as the number of cycles increased. Particle size measurements of the RF resin showed no indication of particle size change (for a given chemical) with cycles and essentially no fines formation. Micrographs comparing representative bead samples before and after testing indicated no change in bead

  12. PILOT-SCALE HYDRAULIC TESTING OF RESORCINOL FORMALDEHYDE ION EXCHANGE RESIN

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, D

    2007-01-09

    Savannah River National Laboratory (SRNL) performed pilot-scale hydraulic/chemical testing of spherical resorcinol formaldehyde (RF) ion exchange (IX) resin for the River Protection Project Hanford Tank Waste Treatment & Immobilization Plant (WTP) Project. The RF resin cycle testing was conducted in two pilot-scale IX columns, 1/4 and 1/2 scale. A total of twenty-three hydraulic/chemical cycles were successfully completed on the spherical RF resin. Seven of the cycles were completed in the 12-inch IX Column and sixteen cycles were completed in the 24-inch IX Column. Hydraulic testing showed that the permeability of the RF resin remained essentially constant, with no observed trend in the reduction of the permeability as the number of cycles increased. The permeability during the pilot-scale testing was 2 1/2 times better than the design requirements of the WTP full-scale system. The permeability of the resin bed was uniform with respect to changes in bed depth. Upflow Regeneration and Simulant Introduction in the IX columns revealed another RF resin benefit; negligible radial pressures to the column walls from the swelling of resin beads. In downflow of the Regeneration and Simulant Introduction steps, the resin bed particles pack tightly together and produce higher hydraulic pressures than that found in upflow. Also, upflow Simulant Introduction produced an ideal level bed for the twenty cycles completed using upflow Simulant Introduction. Conversely, the three cycles conducted using downflow Simulant Introduction produced an uneven bed surface with erosion around the thermowells. The RF resin bed in both columns showed no tendency to form fissures or pack more densely as the number of cycles increased. Particle size measurements of the RF resin showed no indication of particle size change (for a given chemical) with cycles and essentially no fines formation. Micrographs comparing representative bead samples before and after testing indicated no change in bead

  13. PILOT-SCALE HYDRAULIC TESTING OF RESORCINOL FORMALDEHYDE ION EXCHANGE RESIN

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, D

    2006-11-08

    Savannah River National Laboratory (SRNL) performed pilot-scale hydraulic/chemical testing of spherical resorcinol formaldehyde (RF) ion exchange (IX) resin for the River Protection Project-Hanford Tank Waste Treatment & Immobilization Plant (WTP) Project. The RF resin cycle testing was conducted in two pilot-scale IX columns, 1/4 and 1/2 scale. A total of twenty-three hydraulic/chemical cycles were successfully completed on the spherical RF resin. Seven of the cycles were completed in the 12 inch IX Column and sixteen cycles were completed in the 24 inch IX Column. Hydraulic testing showed that the permeability of the RF resin remained essentially constant, with no observed trend in the reduction of the permeability as the number of cycles increased. The permeability during the pilot-scale testing was 2 1/2 times better than the design requirements of the WTP full-scale system. The permeability of the resin bed was uniform with respect to changes in bed depth. Upflow Regeneration and Simulant Introduction in the IX columns revealed another RF resin benefit; negligible radial pressures to the column walls from the swelling of resin beads. In downflow of the Regeneration and Simulant Introduction steps, the resin bed particles pack tightly together and produce higher hydraulic pressures than that found in upflow. Also, upflow Simulant Introduction produced an ideal level bed for the twenty cycles completed using upflow Simulant Introduction. Conversely, the three cycles conducted using downflow Simulant Introduction produced an uneven bed surface with erosion around the thermowells. The RF resin bed in both columns showed no tendency to form fissures or pack more densely as the number of cycles increased. Particle size measurements of the RF resin showed no indication of particle size change (for a given chemical) with cycles and essentially no fines formation. Micrographs comparing representative bead samples before and after testing indicated no change in bead

  14. Carbon-14 behavior in a cement-dominated environment: Implications for spent CANDU resin waste disposal

    International Nuclear Information System (INIS)

    Dayal, R.; Reardon, E.J.

    1994-01-01

    Cement based waste forms and concrete engineered barriers are expected to play a key role in providing 14 C waste containment and control 14 C migration for time periods commensurate with its hazardous life of about 50,000 years. The main thrust of this study was, therefore, to evaluate the performance of cement based waste forms with regard to 14 C containment. Of particular importance are the geochemical processes controlling 14 C solubility and release under anticipated cement dominated low and intermediate level waste repository conditions. Immobilization of carbonate-form exchange resin in grout involves transfer of sorbed 14 CO 3 2- ions, through exchange for hydroxyl ions from the grout slurry, followed by localized precipitation of solid calcium carbonate at the cement/resin interface in the grout matrix. Carbon-14 release behavior can be attributed to the dissolution characteristics and solubility of calcite present in the cement based waste form. The groundwater flow regime can exert a pronounced effect both on the near-field chemistry and the leaching behavior of 14 C. For a cement dominated repository, at relatively low-flow or stagnant groundwater conditions, the alkaline near-field chemical environments inhibits the release of 14 C from the cemented waste form. Under high flow conditions, the near-field environment is characterized by relatively neutral pH conditions which promote calcite dissolution, thus resulting in 14 C release from the waste form

  15. Study of phosphorous based resins for the uptake of plutonium from H2SO4 based analytical waste

    International Nuclear Information System (INIS)

    Seshadri, H.; Mohandas, Jaya; Srinivasan, S.; Kumar, T.; Rajan, S.K.

    2006-01-01

    This study indicates that phosphorous based resins can be conveniently employed for the uptake of plutonium from analytical wastes even in strong acid media and also in the presence of diverse ions like silver and chromium. It is also evident that phosphorous based resins have proved to be efficient even in sulphuric acid medium

  16. A new technology for separation and recovery of materials from waste printed circuit boards by dissolving bromine epoxy resins using ionic liquid

    International Nuclear Information System (INIS)

    Zhu, P.; Chen, Y.; Wang, L.Y.; Qian, G.Y.; Zhou, M.; Zhou, J.

    2012-01-01

    Highlights: ► WPCBs were heated in [EMIM + ][BF 4 − ] for recovering solider at 240 °C. ► The bromine epoxy resins in WPCBs were all dissolved in [EMIM + ][BF 4 − ] at 260 °C. ► Used [EMIM + ][BF 4 − ] is treated by water to obtain regeneration. - Abstract: Recovery of valuable materials from waste printed circuit boards (WPCBs) is quite difficult because WPCBs is a heterogeneous mixture of polymer materials, glass fibers, and metals. In this study, WPCBs was treated using ionic liquid (1-ethyl-3-methylimizadolium tetrafluoroborate [EMIM + ][BF 4 − ]). Experimental results showed that the separation of the solders went to completion, and electronic components (ECs) were removed in WPCBs when [EMIM + ][BF 4 − ] solution containing WPCBs was heated to 240 °C. Meanwhile, metallographic observations verified that the WPCBs had an initial delamination. When the temperature increased to 260 °C, the separation of the WPCBs went to completion, and coppers and glass fibers were obtained. The used [EMIM + ][BF 4 − ] was treated by water to generate a solid–liquid suspension, which was separated completely to obtain solid residues by filtration. Thermal analyses combined with infrared ray spectra (IR) observed that the solid residues were bromine epoxy resins. NMR (nuclear magnetic resonance) showed that hydrogen bond played an important role for [EMIM + ][BF 4 − ] dissolving bromine epoxy resins. This clean and non-polluting technology offers a new way to recycle valuable materials from WPCBs and prevent environmental pollution from WPCBs effectively.

  17. Radioactive liquid waste processing method

    International Nuclear Information System (INIS)

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

    1995-01-01

    Floor drainages are mixed with low electroconductive liquid wastes, and after filtering the mixed liquid wastes by a hollow thread membrane filters, they are subjected to a desalting treatment by a desalter. The mixing ratio of the floor drainages to the lower electroconductive liquid wastes is determined to not more than 50wt%. With such procedures, since ionic ingredients are further diluted by mixing the floor drainages to the low electroconductive liquid wastes, sufficient margin can be provided up to the saturation of the ion exchange resins of the desalter, to maintain the ion exchange performance for a long period of time. Further, the recovery of the amount of permeation water and a differential pressure of filtration upon back washing of the hollow thread membrane filters is facilitated, thereby enabling to perform regeneration easily at high efficiency. (T.M.)

  18. The ChemChar process for hazardous-waste treatment

    International Nuclear Information System (INIS)

    McGowin, A.E.

    1991-01-01

    The ChemChar Reverse-Burn Gasification Process has been studied for application to the thermal destruction of radioactive waste organic ion exchange resins. The resulting char was mixed with cement to form a dry, leach-resistant final disposal product. Successful regeneration of spent granular activated carbons was achieved with reverse-burn gasification. Regeneration parameters such as moisture content and supplemental fuel addition were investigated. The performance of regenerated carbon was evaluated by batch equilibrium and breakthrough assay and was comparable to that of the original. Surface areas were determined by the BET method. The fate of mercury during reverse-burn gasification was investigated. TRB Char adsorbent was used to remove mercury vapor emission from the process. The use of petroleum coke as a substrate for gasification of wastes was studied. Petroleum coke was activated by reverse-burn gasification to produce a highly porous, low surface area solid. Destruction efficiency for hexachlorobenzene on activated coke was considerably lower than on coal char, however, the addition of iron appeared to catalyze hexachlorobenzene gasification

  19. Solidifying power station resins and sludges

    International Nuclear Information System (INIS)

    Willis, A.S.D.; Haigh, C.P.

    1984-01-01

    Radioactive ion exchange resins and sludges arise at nuclear power stations from various operations associated with effluent treatment and liquid waste management. As the result of an intensive development programme, the Central Electricity Generating Board (CEGB) has designed a process to convert power station resins and sludges into a shielded, packaged solid monolithic form suitable for final disposal. Research and development, the generic CEGB sludge/resin conditioning plant and the CEGB Active Waste Project are described. (U.K.)

  20. Destruction of Ion-Exchange Resin In Waste From the HFIR, T1 and T2 Tanks Using Fenton's Reagent

    International Nuclear Information System (INIS)

    Taylor, P.A.

    2002-01-01

    The use of Fenton's reagent (hydrogen peroxide and a ferrous iron catalyst) has been tested as a method for destroying ion-exchange resin in radioactive waste from three underground storage tanks at the Oak Ridge National Laboratory. The resin in these wastes must be destroyed before they can be transferred to the Melton Valley Storage Tanks (MVSTs) prior to solidification and disposal at the Waste Isolation Pilot Plant. The reaction with ion-exchange resin requires a dilute acidic solution (pH = 3 to 5) and moderate temperatures (T = 60 to 100 C). Laboratory-scale tests of the process have been successfully completed using both simulants and actual waste samples. The ion-exchange resin is oxidized to carbon dioxide and inorganic salts. The reaction rate is quite slow for temperatures below 70 C but increases almost linearly as the temperature of the slurry increases from 70 to 90 C. Pilot-scale tests have demonstrated the process using larger samples of actual waste slurries. A sample from the High Flux Isotope Reactor (HFIR) tank, containing 500 mL of settled solids (resin and inorganic sludge) in a total volume of 1800 mL, was successfully treated to meet MVST waste acceptance requirements in 9 h of processing time, using 1650 mL of 50 wt% hydrogen peroxide. A composite sample from the T1 and T2 tanks, which contained 1000 mL of settled solids in a total volume of 2000 mL required 8 h of treatment, using 1540 mL of 50 wt% peroxide, to meet waste acceptance requirements. Hydrogen peroxide reaction rates were 0.71 to 0.74 g H 2 O 2 /L/min, with very low (<2000 mg/L) concentrations of peroxide in the slurry. The reaction produces mostly carbon dioxide gas during the early part of the treatment, when organic carbon concentrations in the slurry are high, and then produces increasing amounts of oxygen as the organic carbon concentration drops. Small amounts (<3 vol%) of carbon monoxide are also generated. The off-gas from the pilot-scale tests, which was 81 vol

  1. Feasible modifications for the low-level waste treatment plant at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Chilton, J.M.

    1984-06-01

    Aqueous, low-level, radioactive wastes at Oak Ridge National Laboratory (ORNL) contain small amounts of 60 Co, 90 Sr, 137 Cs, and trace amounts of other radionuclides. These wastes are processed by passage through beds of a strong-acid cation exchange resin, and the treated water is then discharged to the environment. Studies show that pretreatment of the waste with a weak-acid cation exchange resin would result in a significant decrease in regeneration reagents and a saving of manpower. This can be accomplished in the present plant by piping changes on the existing columns. The effluent from the cation treatment process contains all of the radionuclides that are present in anionic form. Routinely, this consists only of approximately one-half of the 60 Co. Under certain conditions, other anions (such as 131 I) could be present. Studies show that these can be removed by use of an anion exchange resin bed at the end of the process. This would require the construction of an additional column, if the head-end treatment described above is also installed. 2 references, 2 figures, 2 tables

  2. Physico-Chemical Studies Involving Incorporation of Radioactive and Industrial Waste In Cement-Epoxy Resin Matrix

    International Nuclear Information System (INIS)

    Sayed, M.S.; Hafez, N.

    1999-01-01

    Cement and epoxy resin as chemical additives are proposed to incorporate different types of wastes. The study was extended to prepare different mixtures of cement and epoxy resin in presence of some toxic ions. The studied ions were Cd II, Ni II, Cu II, Fe III, Ce IV, 154+152 Eu, phenol and toluene. The physical, mechanical and leaching properties of the mixtures were studied. The thermal analysis and infrared spectra were also investigated. It was observed that all the studied properties of the epoxy modified cement as a disposal matrix was improved

  3. SHALLOW SHELL RESIN VERSUS TRADITIONAL RESIN: A CASE STUDY FOR Cu(II REMOVAL

    Directory of Open Access Journals (Sweden)

    Özgür Arar

    2016-10-01

    Full Text Available A comparative study on Cu2+ removal by shallow shell resin (Purolite SST 60 and traditional strongly acidic cation exchange resin (Purolite PFC 100 was performed. Batch experiments were carried out as a function of  resin  dosage and  solution pH and contact time. Ion exchange reaction showed a pH depended feature.  Maximum removal of Cu2+ achieved  pH  from 2 to 5. Sorption isothermal data is well interpreted by the Langmuir equation. Additionally, kinetic experiments showed that the pseudo first-order model was suitable for such resins. The regeneration performance of shallow shell technology (SST resin is better than PFC 100.  A solution of 2M H2SO4 performed well in regenerationof SST 60 resin. On the other han maximum regeneration reached 80% for PFC 100 resin.Özet: Bu çalışmada, klasik iyon değiştirici reçine (Purolite PFC 100 ve  sığ kabuk  reçine (Purolite SST 60  ile Cu2+ giderilmesi incelenmiştir. Yapılan kesikli çalışmalarla Cu2+ giderilmesine, reçine miktarı, çözelti pH`ı ve temas süresinin etkisi incelenmiştir. Çözelti pH`ının 2 ile 5 arasında olduğu durumda Cu2+ iyonları tamamen giderilmiştir. Denge çalışmalarında elde edilen sonuçlar Langmuir izoterm modeline daha uygun olmuştur. Kinetik çalışmalarda elde edilen sonuçlar yalancı birinci mertebe kinetik modeline uygunluk göstermişir. SST 60 reçinesinin rejenerasyon verimi PFC 100 reçinesinden daha yüksektir. 2M H2SO4 ile SST 60 reçinesi tamamen rejenere edilmiştir.

  4. Regeneration and localization of radioactive waste in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Egorov, N.N.; Kudryavtsev, E.G.; Nikipelov, B.V.; Polyakov, A.S.; Zakharkin, B.S.; Mamaev, L.A.

    1993-01-01

    Normal functioning of the nuclear-power industry is only possible with a closed fuel cycle, including regeneration of the spent fuel from atomic power plants, the production and recycling of the secondary fuel, and localization of the radioactive waste. Despite the diversity of contemporary attitudes toward the structure of the nuclear fuel cycle around the world, the closure of the fuel cycle has been fundamental to the atomic-power industry in the USSR since the very beginning, and has taken on even greater significance in Russia today. From the beginning, the idea of a closed fuel cycle has been based essentially on one fundamental criterion: the concept of expanded productivity on the basis of fuel regeneration, i.e., the economic factor. Important as economic factors are, safety issues have taken on great significance in recent years: not only power-station reactors but all the ancillary stages of the fuel cycle must meet fundamentally new reliability, safety, and environmental hazards. The RT-1 plant is a versatile operation, regenerating spent fuel from VVER-440, BN-350, and BN-600 reactors, nuclear icebreakers and submarines, research reactors, and other power units. The plant can reprocess 400 ton/year of basic VVER-440 fuel. World-class modern processes have been introduced at the plant, meeting the necessary quality standards: zonal planning, remote operation to eliminate direct contact of the staff with radioactive material, extensive monitoring and control systems, multistage gas-purification systems, and new waste-treatment methods

  5. Low-level waste management - suggested solutions for problem wastes

    International Nuclear Information System (INIS)

    Pechin, W.H.; Armstrong, K.M.; Colombo, P.

    1984-01-01

    Problem wastes are those wastes which are difficult or require unusual expense to place into a waste form acceptable under the requirements of 10 CFR 61 or the disposal site operators. Brookhaven National Laboratory has been investigating the use of various solidification agents as part of the DOE Low-Level Waste Management Program for several years. Two of the leading problem wastes are ion exchange resins and organic liquids. Ion exchange resins can be solidified in Portland cement up to about 25 wt % resin, but waste forms loaded to this degree exhibit significantly reduced compressive strength and may disintegrate when immersed in water. Ion exchange resins can also be incorporated into organic agents. Mound Laboratory has been investigating the use of a joule-heated glass melter as a means of disposing of ion exchange resins and organic liquids in addition to other combustible wastes

  6. Oxidative degradation of low and intermediate level Radioactive organic wastes 2. Acid decomposition on spent Ion-Exchange resins

    Energy Technology Data Exchange (ETDEWEB)

    Ghattas, N K; Eskander, S B [Radioisotope dept., atomic energy authority, (Egypt)

    1995-10-01

    The present work provides a simplified, effective and economic method for the chemical decomposition of radioactively contaminated solid organic waste, especially spent ion - exchange resins. The goal is to achieve volume reduction and to avoid technical problems encountered in processes used for similar purposes (incineration, pyrolysis). Factors efficiency and kinetics of the oxidation of the ion exchange resins in acid medium using hydrogen peroxide as oxidant, namely, duration of treatment and the acid to resin ratio were studied systematically on a laboratory scale. Moreover the percent composition of the off-gas evolved during the decomposition process was analysed. 3 figs., 5 tabs.

  7. Oxidative degradation of low and intermediate level Radioactive organic wastes 2. Acid decomposition on spent Ion-Exchange resins

    International Nuclear Information System (INIS)

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

    1995-01-01

    The present work provides a simplified, effective and economic method for the chemical decomposition of radioactively contaminated solid organic waste, especially spent ion - exchange resins. The goal is to achieve volume reduction and to avoid technical problems encountered in processes used for similar purposes (incineration, pyrolysis). Factors efficiency and kinetics of the oxidation of the ion exchange resins in acid medium using hydrogen peroxide as oxidant, namely, duration of treatment and the acid to resin ratio were studied systematically on a laboratory scale. Moreover the percent composition of the off-gas evolved during the decomposition process was analysed. 3 figs., 5 tabs

  8. Removal of Uranium by Exchanger Resins from Soil Washing Solution

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Soo; Han, G. S.; Kim, G. N.; Koo, D. S.; Jeong, J. W.; Moon, J. K. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Uranyl ions in the acidic waste solution were sorbed on AM-resin resin with a high sorption efficiency, and desorbed from the resin by a batch-type washing with a 60 .deg. C heated 0.5 M Na{sub 2}CO{sub 3} solution. However, the uranium dissolved in the sulfuric acid solution was not sorbed onto the strong anion exchanger resins. Our group has developed a decontamination process with washing and electrokinetic methods for uranium-contaminated (U-contaminated) soil. However, this process generates a large amount of waste solution containing various metal ions. If the uranium selectively removed from the waste solution, a very small amount of the 2nd waste would be generated. Thus, selective sorption of uranium by ion exchange resins was examined in this study.

  9. ION EXCHANGE MODELING FOR REMOVAL OF CESIUM FROM HANFORD WASTE USING SUPERLIG 644 RESIN

    International Nuclear Information System (INIS)

    Hamm, L

    2004-01-01

    The expected performance of a proposed ion exchange column using SuperLig(reg s ign) 644 resin for the removal of cesium from Hanford high level radioactive alkaline waste is discussed. This report represents a final report on the ability and knowledge with regard to modeling the Cesium-SuperLig(reg s ign) 644 resin ion exchange system. Only the loading phase of the cycle process is addressed within this report. Pertinent bench-scale column tests and batch equilibrium experiments are addressed. The methodology employed and sensitivity analyses are also included (i.e., existing methodology employed is referenced to prior developmental efforts while updated methodology is discussed). Pilot-scale testing is not assessed since no pilot-scale testing was available at the time of this report. Column performance predictions are made considering three selected feed compositions under nominal operating conditions. The sensitivity analyses provided help to identify key parameters that aid in resin procurement acceptance criteria. The methodology and application presented within this report reflect the expected behavior of SuperLig(reg s ign) 644 resin manufactured at the production-scale (i.e, 250 gallon batch size level). The primary objective of this work was, through modeling and verification based on experimental assessments, to predict the cesium removal performance of SuperLig(reg s ign) 644 resin for application in the RPP pretreatment facility

  10. Boron removal from aqueous solutions by ion-exchange resin: Column sorption-elution studies

    International Nuclear Information System (INIS)

    Koese, T. Ennil; Oztuerk, Nese

    2008-01-01

    A column sorption-elution study was carried out by using a strong base anion-exchange resin (Dowex 2 x 8) for the removal of boron from aqueous solutions. The breakthrough curve was obtained as a function of feed flow rate and the total and breakthrough capacity values of the resin were calculated. The boron on the resin was quantitatively eluted with 0.5 M HCl solution at different flow rates. Three consecutive sorption-elution-washing-regeneration-washing cycles were applied to the resin in order to investigate the reusability of the ion-exchange resin. Total capacity values remained almost the same after three sorption-elution-regeneration cycles. The Thomas and the Yoon-Nelson models were applied to experimental data to predict the breakthrough curves and to determine the characteristic column parameters required for process design. The results proved that the models would describe the breakthrough curves well

  11. Gamma irradiation-induced modifications of resins found in nuclear waste embedding processes; Modifications induites par irradiation gamma dans les differentes resines rencontrees dans le traitement des dechets nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Allali, H [Faculte des Science et Techniques, Settat (Morocco); Debre, O; Lambert, M; Nsouli, B; Thomas, J P [Inst. de Physique Nucleaire, Lyon-1 Univ., 69 - Villeurbanne (France); Collaboration: IPN-Lyon, Laboratoire d` Etude des Materiaux Plastiques et des Biomateriaux, URA 507, UCBL, Agence Nationale pour la Gestion des Dechets Radiactifs

    1999-12-31

    The various resins involved in nuclear waste disposal are subject to gamma irradiation-induced modifications. From Time-of-Flight Mass Spectrometry of Secondary ions emitted under High Energy Projectiles bombardment (HSF-SIMS) of such materials the following results are obtained: the embedding epoxy resin DGEBA-DDM does not exhibit significant bulk changes in chemical structure, whatever the dose rate and irradiation medium (air or water), at least up to 2 MGy. However, oxidation processes are well observed at the very surface. Under the same conditions Ion exchange resins to be embedded are subjected to scissions of their functional sites, leading to fixed ion release. Dehydration under irradiation is observed pointing out the crucial role of water in ion transport outside of the material. (authors)

  12. On-site Destruction of Radioactive Oily Wastes Using Adsorption Coupled with Electrochemical Regeneration - 12221

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N.W. [Arvia Technology Ltd, Daresbury Innovation Centre, Daresbury, WA4 4FS (United Kingdom); Wickenden, D.A. [Magnox Ltd, Berkeley Centre, Gloucestershire, GL13 9PB (United Kingdom); Roberts, E.P.L. [School of Chemical Engineering, University of Manchester, Manchester, M60 1QD (United Kingdom)

    2012-07-01

    Arvia{sup R}, working with Magnox Ltd, has developed the technology of adsorption coupled with electrochemical regeneration for the degradation of orphan radioactive oil wastes. The process results in the complete destruction of the organic phase where the radioactivity is transferred to liquid and solid secondary wastes that can then be processed using existing authorised on-site waste-treatment facilities.. Following on from successful laboratory and pilot scale trials, a full scale, site based demonstrator unit was commissioned at the Magnox Trawsfynydd decommissioning site to destroy 10 l of LLW and ILW radioactive oils. Over 99% of the emulsified oil was removed and destroyed with the majority of activity (80 - 90%) being transferred to the aqueous phase. Secondary wastes were disposed of via existing routes with the majority being disposed of via the sites active effluent treatment plant. The regeneration energy required to destroy a litre of oil was 42.5 kWh/l oil. This on-site treatment approach eliminates the risks and cost associated with transporting the active waste oils off site for incineration or other treatment. The Arvia{sup R} process of adsorption coupled with electrochemical regeneration has successfully demonstrated the removal and destruction of LLW and ILW radioactive oils on a nuclear site. Over 99.9% of the emulsified oil was removed, with the majority of the radioactive species transferred to the aqueous, supernate, phase (typically 80 - 90 %). The exception to this is Cs-137 which appears to be more evenly distributed, with 43% associated with the liquid phase and 33 % with the Nyex, the remainder associated with the electrode bed. The situation with Plutonium may be similar, but this requires confirmation, hence further work is underway to understand the full nature of the electrode bed radioactive burden and its distribution within the body of the electrodes. - Tritium gaseous discharges were negligible; hence no off-gas treatment

  13. Integrated bicarbonate-form ion exchange treatment and regeneration for DOC removal: Model development and pilot plant study.

    Science.gov (United States)

    Hu, Yue; Boyer, Treavor H

    2017-05-15

    The application of bicarbonate-form anion exchange resin and sodium bicarbonate salt for resin regeneration was investigated in this research is to reduce chloride ion release during treatment and the disposal burden of sodium chloride regeneration solution when using traditional chloride-form ion exchange (IX). The target contaminant in this research was dissolved organic carbon (DOC). The performance evaluation was conducted in a completely mixed flow reactor (CMFR) IX configuration. A process model that integrated treatment and regeneration was investigated based on the characteristics of configuration. The kinetic and equilibrium experiments were performed to obtain required parameters for the process model. The pilot plant tests were conducted to validate the model as well as provide practical understanding on operation. The DOC concentration predicted by the process model responded to the change of salt concentration in the solution, and showed a good agreement with pilot plant data with less than 10% difference in terms of percentage removal. Both model predictions and pilot plant tests showed over 60% DOC removal by bicarbonate-form resin for treatment and sodium bicarbonate for regeneration, which was comparable to chloride-form resin for treatment and sodium chloride for regeneration. Lastly, the DOC removal was improved by using higher salt concentration for regeneration. Copyright © 2017 Elsevier Ltd. All rights reserved.

  14. Summary of pilot-scale activities with resorcinol ion exchange resin

    International Nuclear Information System (INIS)

    Cicero, C.A.; Bickford, D.F.; Sargent, T.N.; Andrews, M.K.; Bibler, J.P.; Bibler, N.E.; Jantzen, C.M.

    1995-01-01

    The Mixed Waste Focus Area (MWFA) of the Department of Energy (DOE) is currently investigating vitrification technology for treatment of low level mixed wastes (LLMW). They have chartered the Savannah River Technology Center (SRTC) to study vitrification of the wastes through an Office of Technology Development (OTD) Technical Task Plan (TTP). SRTC's efforts have included crucible-scale studies and pilot scale testing on simulated LLMW sludges, resins, soils, and other solid wastes. Results from the crucible-scale studies have been used as the basis for the pilot-scale demonstrations. As part of the fiscal year (FY) 1995 activities, SRTC performed crucible-scale studies with organic resins. This waste stream was selected because of the large number of DOE sites, as well as commercial industries, that use resins for treatment of liquid wastes. Pilot-scale studies were to be completed in FY 1995, but could not be due to a reduction in funding. Instead, a compilation of pilot-scale tests with organic resins performed under the guidance of SRTC was provided in this report. The studies which will be discussed used a resorcinol- formaldehyde resin loaded with non-radioactive cesium, which was fed with simulated wastewater treatment sludge feed. The first study was performed at the SRTC in the mini-melter, 1/100th scale of the Defense Waste Processing Facility (DWPF) melter, and also involved limited crucible-scale studies to determine the resin loading obtainable. The other study was performed at the DOE/Industrial Center for Vitrification Research (Center) and involved both crucible and pilot-scale testing in the Stir-Melter stirred-melter. Both studies were successful in vitrifying the resin in simulated radioactive sludge and glass additive feeds

  15. Kinetic study of ion exchange in phosphoric acid chelating resin

    International Nuclear Information System (INIS)

    Brikci-Nigassa, Mounir; Hamouche, Hafida

    1995-11-01

    Uranium may be recovered as a by product of wet phosphoric acid using a method based on specific ion exchange resins. These resins called chelates contain amino-phosphonic functional groups. The resin studied in this work is a purolite S-940; uranium may be loaded on this resin from 30% P2O5 phosphoric acid in its reduced state. The influence of different parameters on the successive steps of the process have been studied in batch experiments: uranium reduction, loading and oxydation. Uranium may be eluted with ammonium carbonate and the resin regeneration may be done with hydrochloric acid.Ferric ions reduce the effective resin capacity considerably and inert fixation conditions are proposed to enhance uranium loading

  16. Deactivation and regeneration of ZSM-5 zeolite in catalytic pyrolysis of plastic wastes

    International Nuclear Information System (INIS)

    Lopez, A.; Marco, I. de; Caballero, B.M.; Adrados, A.; Laresgoiti, M.F.

    2011-01-01

    Highlights: → Pyrolysis transforms plastic wastes in valuable liquids and gases useful as fuels or source of chemicals. → The use of ZSM-5 zeolite in pyrolysis favours the production of gases and of lighter and more aromatic liquids. → ZSM-5 zeolite is almost completely deactivated after one plastics pyrolysis experiment. → ZSM-5 zeolite used in plastic wastes pyrolysis can be regenerated by burning the deposited coke in an air stream. → Regenerated ZSM-5 recovers its activity and produces liquids and gases equivalent to those obtained with fresh catalyst. - Abstract: In this work, a study of the regeneration and reuse of ZSM-5 zeolite in the pyrolysis of a plastic mixture has been carried out in a semi-batch reactor at 440 deg. C. The results have been compared with those obtained with fresh-catalyst and in non-catalytic experiments with the same conditions. The use of fresh catalyst produces a significant change in both the pyrolysis yields and the properties of the liquids and gases obtained. Gases more rich in C3-C4 and H 2 are produced, as well as lower quantities of aromatic liquids if compared with those obtained in thermal decomposition. The authors have proved that after one pyrolysis experiment the zeolite loses quite a lot of its activity, which is reflected in both the yields and the products quality; however, this deactivation was found to be reversible since after regeneration heating at 550 deg. C in oxygen atmosphere, this catalyst recovered its initial activity, generating similar products and in equivalent proportions as those obtained with fresh catalyst.

  17. Waste management implications of irradiated nuclear fuel storage

    International Nuclear Information System (INIS)

    Larrick, A.P.; Schneidmiller, D.

    1977-01-01

    Irradiated nuclear fuel is stored underwater in large recirculating basins prior to reprocessing in chemical separations plants. A small number of the fuel rods contain minor defects which allow fission products to seep into the basin water. The predominant radionuclides leached into the water are dependent upon the decay time after removal from the reactor. Freshly discharged fuel releases short half-lived radioiodine which presents exposure and airborne release problems on a short-term basis but does not impose significant long-term waste management problems. After a reasonable decay period, the major radionuclides present are 137 Cs, 144 Ce, 90 Sr, 3 H, and 239 / 240 Pu. Each of the radionuclides imposes specific waste management problems which require individual consideration in their control, packaging and shipment, and permanent disposal. These considerations are reviewed in this paper for general industry practices and specific illustrative examples are presented. Control of radionuclide concentrations and water purity are normally achieved by filtration and ion exchange treatment. Cartridge-type filters are the most common but improved filtration techniques that minimize personnel exposure, particularly during filter replacements, are now being adopted. Mixed bed ion exchange resins are most commonly utilized for soluble radionuclide removal, particularly for basins filled with demineralized water. Cesium-specific exchange media are employed at basins where demineralized water is not employed; these media operate for very long periods of time since they are not depleted by the normal dissolved non-radioactive water impurities. The resins are either buried when depleted or regenerated and the regeneration solutions concentrated for burial. Resin run lengths are usually determined by ionic or radiochemical depletion of the resin or in some cases by limiting radionuclide concentrations specified by shipping regulations or established ion column dose rates

  18. Properties of the Carboxylate ion exchange resins

    International Nuclear Information System (INIS)

    Allard, Bert; Dario, Maarten; Boren, Hans; Torstenfelt, Boerje; Puigdomenech, Ignasi; Johansson, Claes

    2002-09-01

    Weakly acidic, carboxylic resin has been selected, together with strong base anion resins, for water purification at the Forsmark 1 and 2 reactors. For the strong (but not the weak) ion exchange resin the Nuclear Power Inspectorate has given permission to dispose the spent resins in the SFR 1 (the Final Repository for Radioactive Operational Waste). This report gives a review of the carboxylic resins and comes to the conclusion that the resins are very stable and that there should not exist any risks for increased leaching of radionuclides from SFR 1 if these resins are disposed (compared to the strong resins)

  19. Preliminary flowsheet: Ion exchange for separation of cesium from Hanford tank waste using resorcinol-formaldehyde resin

    International Nuclear Information System (INIS)

    Penwell, D.L.

    1994-01-01

    This preliminary flowsheet document describes an ion exchange process which uses resorcinol-formaldehyde (R-F) resin to remove cesium from Hanford tank waste. The flowsheet describes one possible equipment configuration, and contains mass balances based on that configuration with feeds of Neutralized Current Acid Waste, and Double Shell Slurry Feed. The flowsheet also discusses process alternatives, unresolved issues, and development needs associated with the ion exchange process. It is expected that this flowsheet will evolve as open issues are resolved and progress is made on development needs. This is part of the Tank Waste Remediation Program at Hanford. 26 refs, 6 figs, 25 tabs

  20. Application of titanates, niobates, and tantalates to neutralized defense waste decontamination: materials properties, physical forms, and regeneration techniques. Final report

    International Nuclear Information System (INIS)

    Dosch, R.G.

    1981-01-01

    A study of the application of sodium titanate (ST) to the decontamination of neutralized defense waste has been completed. The work was directed at Sr removal from dissolved salt cake, simulated in this work with a 6.0 N NaNO 3 - 0.6 N NaOH solution. Three physical forms of the titanates were developed including powder, pellets, and titanate-loaded resin beads and all were found to be superior to conventional organic ion exchange in this application. When spent, the titanate materials can be calcined to an oxide from which is a stable waste form in itself or can be added directly to a glass melter to become part of a vitrified waste form. Radiation stability of titanate powder and resin forms was assessed in tests in which these materials were exposed to 60 Co radiation. The strontium exchange capacity of the powder remained constant through a dose of 3 x 10 7 rads and retained 50% capacity after a dose of 2 x 10 9 rads. The primary mechanism involved in loss of capacity was believed to be heating associated with the irradiation. The resin forms were unchanged through a dose of 5 x 10 8 rads and retained 30% capacity after a dose of 2 x 10 9 rads. The latter dose resulted in visible degradation of the resin matrix. Anion exchange resins loaded with sodium niobate and sodium tantalate were also prepared by similar methods and evaluated for this application. These materials had Sr sorption properties comparable to the titanate material; however, they would have to provide a significant improvement to justify their higher cost

  1. Analysis by Fourier Transform Infrared (FTIR) of the gamma radiation effect on epoxy resin, used as solidification agent of radioactive wastes

    International Nuclear Information System (INIS)

    Liu, C.H.; Riella, H.G.; Guedes, S.M.L.

    1995-01-01

    The effects of gamma radiation on Epoxy resin, used as solidification agent of radioactive wastes, were studied by Fourier Transform Infrared (FTIR). The spectra showed no significant modifications on Epoxy resin functional groups (irradiated with dose from 0 to 1 MGy). Up to 1 MGy Epoxy resin did not oxidize, confirming the Epoxy good radiation strength. The presence of aromatic chain and amine group, mainly tertiary amine, give good radiolytic stability to the Epoxy, increasing the interest to use this material in nuclear facilities. (author). 3 refs, 2 figs

  2. Experience in industrial operation of the plant for immobilizing radioactive wastes in thermosetting resins at the Ardennes Nuclear Power Station

    International Nuclear Information System (INIS)

    Haller, P.; Romestain, P.; Bruant, J.P.

    1983-01-01

    The French Atomic Energy Commission (CEA) has developed, at the Grenoble Centre for Nuclear Studies, a procedure for immobilizing low- and intermediate-level wastes in thermosetting resins of the polyester or epoxy types. To demonstrate feasibility on an industrial scale, a pilot plant has been set up at the effluent treatment station of the Ardennes Franco-Belgium Nuclear Power Station (SENA), which is a 305 MW(e) PWR type. Assembly work began in January 1979. After a period devoted to final adjustments and operation with inactive products, conditioning of active products began in January 1981. In the paper, the methods of conditioning the three types of waste (evaporation concentrates, ion exchange resins and filter cartridges) are described, experience of the start-up and operation of the plant is reported and the principal results of coating characterization tests are given. The results of tests on active and inactive products show that the characteristics of the materials obtained on an industrial scale match those of laboratory products and confirm their high quality with regard to mechanical behaviour, fire resistance, homogeneity and low-leachability. Industrial experience and economic comparisons show that the process of immobilizing waste from nuclear power stations in thermosetting resins offers an extremely interesting alternative to classical methods of conditioning. (author)

  3. Spherical Resorcinol-Formaldehyde Resin for the Removal of Cesium from Hanford Tank Waste

    International Nuclear Information System (INIS)

    Wilmarth, William

    2006-01-01

    This power-point presentation highlights that spherical RF continues to out-perform baseline technology (SuperLig(registered trademark)644) due to several factors: increased radiation stability from Cs Capacity perspective, outstanding hydraulic performance, and flammable gas generation being the same as the baseline technology. Actual waste testing at Pacific Northwest National Lab. is going well, and isotherm modeling is continuing at Savannah River National Lab. Regulatory analysis of spent resin is planned for early summer 2006

  4. Controlled catalytic and thermal sequential pyrolysis and hydrolysis of phenolic resin containing waste streams to sequentially recover monomers and chemicals

    Science.gov (United States)

    Chum, H.L.; Evans, R.J.

    1992-08-04

    A process is described for using fast pyrolysis in a carrier gas to convert a waste phenolic resin containing feedstreams in a manner such that pyrolysis of said resins and a given high value monomeric constituent occurs prior to pyrolyses of the resins in other monomeric components therein comprising: selecting a first temperature program range to cause pyrolysis of said resin and a given high value monomeric constituent prior to a temperature range that causes pyrolysis of other monomeric components; selecting, if desired, a catalyst and a support and treating said feedstreams with said catalyst to effect acid or basic catalyzed reaction pathways to maximize yield or enhance separation of said high value monomeric constituent in said first temperature program range to utilize reactive gases such as oxygen and steam in the pyrolysis process to drive the production of specific products; differentially heating said feedstreams at a heat rate within the first temperature program range to provide differential pyrolysis for selective recovery of optimum quantity of said high value monomeric constituent prior to pyrolysis of other monomeric components therein; separating said high value monomeric constituent; selecting a second higher temperature program range to cause pyrolysis of a different high value monomeric constituent of said phenolic resins waste and differentially heating said feedstreams at said higher temperature program range to cause pyrolysis of said different high value monomeric constituent; and separating said different high value monomeric constituent. 11 figs.

  5. Eichrom's ABEC trademark resins: Alkaline radioactive waste treatment, radiopharmaceutical, and potential hydrometallurgical applications

    International Nuclear Information System (INIS)

    Bond, A.H.; Gula, M.J.; Chang, F.; Rogers, R.D.

    1997-01-01

    Eichrom's ABEC trademark resins selectivity extract certain anions from high ionic strength acidic, neutral, or strongly alkaline media, and solute stripping can be accomplished by eluting with water. ABEC resins are stable to pH extreme and radiolysis and operate in high ionic strength and/or alkaline solutions where anion-exchange is often ineffective. Potential applications of the ABEC materials include heavy metal and ReO 4 - separations in hydrometallurgy and purification of perrhenate iodide, and iodate in radiopharmaceutical production. Separation of 99m TcO 4 - from its 99 MoO 4 2- parent and stripping with water or physiological saline solution have been demonstrated for radiopharmaceutical applications. Removal of 99 TcO 4 - and 129 I - from alkaline tank wastes has also been successfully demonstrated. The authors will discuss the scale-up studies, process-scale testing, and market development of this new extraction material

  6. Method of solidifying radioactive wastes

    International Nuclear Information System (INIS)

    Maeda, Masahiko; Kira, Satoshi; Watanabe, Naotoshi; Nagaoka, Takeshi; Akane, Junta.

    1982-01-01

    Purpose: To obtain solidification products of radioactive wastes having sufficient monoaxial compression strength and excellent in water durability upon ocean disposal of the wastes. Method: Solidification products having sufficient strength and filled with a great amount of radioactive wastes are obtained by filling and solidifying 100 parts by weight of chlorinated polyethylene resin and 100 - 500 parts by weight of particular or powderous spent ion exchange resin as radioactive wastes. The chlorinated polyethylene resin preferably used herein is prepared by chlorinating powderous or particulate polyethylene resin in an aqueous suspending medium or by chlorinating polyethylene resin dissolved in an organic solvent capable of dissolving the polyethylene resin, and it is crystalline or non-crystalline chlorinated polyethylene resin comprising 20 - 50% by weight of chlorine, non-crystalline resin with 25 - 40% by weight of chlorine being particularly preferred. (Horiuchi, T.)

  7. Advances in the disposal of radioactive ion exchange resins

    International Nuclear Information System (INIS)

    McCoy, S.B.

    1983-01-01

    During the last several years, more stringent regulations have been imposed on the disposal of low-level radioactive wastes. In particular, the disposal of high-activity ion exchange resins has been affected by the recent requirements intended to enhance waste stability. High-activity resins must now be either solidified or placed in a ''high-integrity'' container. The allowable levels of free liquids in the containers have also been reduced. Solidification of resins has long been applied at nuclear power stations, but new designs in high-integrity containers and dewatering techniques to enhance the waste stability and ensure regulatory compliance have been developed and are being introduced for use at power stations

  8. Evaluation of adsorbent and ion exchange resins for removal of organic matter from petroleum refinery wastewaters aiming to increase water reuse.

    Science.gov (United States)

    de Abreu Domingos, Rodrigo; da Fonseca, Fabiana Valéria

    2018-05-15

    The oil refinery industry seeks solutions to reduce its water uptake and consumption by encouraging the reuse of internal streams and wastewater from treatment systems. After conventional treatment the petroleum refinery wastewater still contains a considerable quantity of recalcitrant organics and the adsorption on activated carbon is currently used in Brazilian refineries, although it is still expensive due to the difficulty of its regeneration. This study evaluated the use of adsorbent and ion exchange resins for the removal of organic matter from refinery wastewater after conventional treatment in order to verify its feasibility, applying successive resin regenerations and comparing the results with those obtained for activated carbon process. Adsorption isotherms experiments were used to evaluate commercial resins, and the most efficient was subjected to column experiments, where absorbance (ABS) and total organic carbon (TOC) removal were measured. The adsorption isotherm of the best resin showed an adsorptive capacity that was 55% lower than that of activated carbon. On the other hand, the column experiments indicated good removal efficiency, and the amount of TOC in the treated wastewater was as good as has been reported in the literature for activated carbon. The regeneration efficiency of the retained organics ranged from 57 to 94%, while regenerant consumption ranged from 12 to 79% above the amount recommended by the resin supplier for the removal of organic material from natural sources, showing the great resistance of these recalcitrant compounds to desorption. Finally, an estimate of the service life of the resin using intermediate regeneration conditions found it to be seven times higher than that of activated carbon when the latter is not regenerated. Copyright © 2018 Elsevier Ltd. All rights reserved.

  9. Spray drying of bead resins: feasibility tests

    International Nuclear Information System (INIS)

    Gay, R.L.; Grantham, L.F.; Jones, L.J.

    1984-01-01

    Rockwell International has developed a volume reduction system for low-level reactor wastes based on drying the wastes in a heated-air spray dryer. The drying of slurries of sodium sulfate, boric acid, and powdered ion exchange resins was demonstrated in previous tests. The drying of bead ion exchange resins can be especially difficult due to the relatively large size of bead resins (about 500 to 800 microns) and their natural affinity for water. This water becomes part of the pore structure of the resins and normally comprises 50 t 60 wt % of the resin weight. A 76-cm-diameter spray dryer was used for feasibility tests of spray drying of cation and anion bead resins. These resins were fed to the dryer in the as-received form (similar to dewatered resins) and as slurries. A dry, free-flowing product was produced in all the tests. The volume of the spray-dried product was one-half to one-third the volume of the as-received material. An economic analysis was made of the potential cost savings that can be achieved using the Rockwel spray dryer system. In-plant costs, transportation costs, and burial costs of spray-dried resins were compared to similar costs for disposal of dewatered resins. A typical utility producing 170 m 3 (6,000 ft 3 ) per year of dewatered resins can save $600,000 to $700,000 per year using this volume reduction system

  10. Preliminary Ion Exchange Modeling for Removal of Cesium from Hanford Waste Using SuperLig 644 Resin

    International Nuclear Information System (INIS)

    Hamm, L.L.

    2000-01-01

    A proposed facility is being designed for the immobilization of Hanford high-level radioactive waste. One unit process in the facility is designed to remove radioactive cesium by ion-exchange from the strongly alkaline aqueous phase. A resin specifically designed with high selectivity of cesium under alkaline conditions is being investigated. The resin also is elutable under more acidic conditions. The proposed design of the facility consists of two sets of two packed columns placed in series (i.e., a lead column followed by a lag (guard) column configuration). During operation, upon reaching a specified cesium concentration criterion at the exit of the lag column, operation is switched to the second set of lead and lag columns. The cesium-loaded lead column is processed (i.e., washed and eluted) and switched to the lag position. the previous lag column is then placed in the lead position (without eluting) and the system is ready for use in the next cycle. For a well designed process, the loading and elution processes result in significant volume reductions in aqueous high-level waste

  11. Incineration of spent ion exchange resin

    International Nuclear Information System (INIS)

    Hasegawa, Chiaki

    1990-01-01

    It is a pressing need to reduce radioactive waste which is generated from the maintenance and operation of a nuclear power plant. Incineration of low level combustible solid waste such as polyethylene seats, paper and others have been successfully performed since 1984 at the Shimane Nuclear Power Station. Furthermore, for extending incineration treatment to spent ion exchange resin, the incineration test was carried out in 1989. However, as the cation exchange resin contains sulfur and then incineration generates SOx gases, so the components of this facility will be in a corrosive environment. We surveyed incineration conditions to improve the corrosive environment at the exhaust gas treatment system. This paper includes these test results and improved method to incinerate spent ion exchange resin. (author)

  12. Enzymatic transesterification of waste vegetable oil to produce biodiesel.

    Science.gov (United States)

    Lopresto, C G; Naccarato, S; Albo, L; De Paola, M G; Chakraborty, S; Curcio, S; Calabrò, V

    2015-11-01

    An experimental study on enzymatic transesterification was performed to produce biodiesel from waste vegetable oils. Lipase from Pseudomonas cepacia was covalently immobilized on a epoxy-acrylic resin support. The immobilized enzyme exhibited high catalytic specific surface and allowed an easy recovery, regeneration and reutilisation of biocatalyst. Waste vegetable oils - such as frying oils, considered not competitive with food applications and wastes to be treated - were used as a source of glycerides. Ethanol was used as a short chain alcohol and was added in three steps with the aim to reduce its inhibitory effect on lipase activity. The effect of biocatalyst/substrate feed mass ratios and the waste oil quality have been investigated in order to estimate the process performances. Biocatalyst recovery and reuse have been also studied with the aim to verify the stability of the biocatalyst for its application in industrial scale. Copyright © 2015 Elsevier Inc. All rights reserved.

  13. Removal of Pb2+ from aqueous solutions by a high-efficiency resin

    International Nuclear Information System (INIS)

    Guo, Hao; Ren, Yongzheng; Sun, Xueliang; Xu, Yadi; Li, Xuemei; Zhang, Tiancheng; Kang, Jianxiong; Liu, Dongqi

    2013-01-01

    The removal of Pb 2+ from aqueous solution by 732 cation-exchange resin in sodium type (732-CR) has been studied in batch experiments at varying pH (2.0–8.0), Pb 2+ concentration (50–200 mg/L), contact time (5–300 min), temperature (288–308 K) and resin dose (0.125–0.75 g/L). The experimental data show that the ion-exchange process was dependent on pH and temperature, the optimal exchange capacity was found at pH 4.0, and higher temperature was beneficial to lead sorption. Kinetic data indicate that the ion-exchange process followed a pseudo-first order model. The equilibrium exchange capacity could be reached at approximately 4 h, and the maximum sorption capacity of Pb 2+ at pH 4.0 was 396.8 mg/g resin. The equilibrium data were evaluated with Langmuir and Freundlich model, and were best fitted with Langmuir model. The thermodynamic parameters for removal of Pb 2+ indicate that the reaction was spontaneous and endothermic. Additionally, column tests were conducted by using both synthetic solution and effluents from lead battery industry. The regeneration of resin was performed for two sorption-regeneration cycles by 1 M NaOH, and the results show that effective regeneration was achieved by this method.

  14. Bituminization of simulated PWR type reactor wastes, boric evaporator bottons and ion exchange resins, carried out in CNEN/SP using commercial bitumen available in the Brazilian market

    International Nuclear Information System (INIS)

    Grosche Filho, C.E.; Chandra, U.

    1986-01-01

    The first results of the study of bituminization of simulated PWR wastes, boric evaporator bottons and spent ion-exclange resins (OH - , H + ) and incinerated ash-wates are presented and discussed. The study consisted of characterization of the commercial bitumen, locally available and bitumen wastes products of varying whight compositions. The characterization was carried out using standard analysis methods of ABNT and ASTM, and included measurement of, penetration, softening point and flash point. In addition, the bitumen samples were analized for their resin and asphaltene contents. For leaching studies, wastes products of bitumen and resin loaded with 134 Cs were utilized. The method used was according to the ISO norms. The simulation of the industrial process was carried out using an extruder-evaporator typically used in the plastic industries offered by Industria de Maquinas Miotto Ltda., Sao Bernardo do Campo - SP. (Author) [pt

  15. Diclofenac removal in urine using strong-base anion exchange polymer resins.

    Science.gov (United States)

    Landry, Kelly A; Boyer, Treavor H

    2013-11-01

    One of the major sources of pharmaceuticals in the environment is wastewater effluent of which human urine contributes the majority of pharmaceuticals. Urine source separation has the potential to isolate pharmaceuticals at a higher concentration for efficient removal as well as produce a nutrient byproduct. This research investigated the efficacy of using strong-base anion exchange polymer resins to remove the widely detected and abundant pharmaceutical, diclofenac, from synthetic human urine under fresh and ureolyzed conditions. The majority of experiments were conducted using a strong-base, macroporous, polystyrene resin (Purolite A520E). Ion-exchange followed a two-step removal rate with rapid removal in 1 h and equilibrium removal in 24 h. Diclofenac removal was >90% at a resin dose of 8 mL/L in both fresh and ureolyzed urine. Sorption of diclofenac onto A520E resin was concurrent with desorption of an equivalent amount of chloride, which indicates the ion-exchange mechanism is occurring. The presence of competing ions such as phosphate and citrate did not significantly impact diclofenac removal. Comparisons of three polystyrene resins (A520E, Dowex 22, Dowex Marathon 11) as well as one polyacrylic resin (IRA958) were conducted to determine the major interactions between anion exchange resin and diclofenac. The results showed that polystyrene resins provide the highest level of diclofenac removal due to electrostatic interactions between quaternary ammonium functional groups of resin and carboxylic acid of diclofenac and non-electrostatic interactions between resin matrix and benzene rings of diclofenac. Diclofenac was effectively desorbed from A520E resin using a regeneration solution that contained 4.5% (m/m) NaCl in an equal-volume mixture of methanol and water. The greater regeneration efficiency of the NaCl/methanol-water mixture over the aqueous NaCl solution supports the importance of non-electrostatic interactions between resin matrix and benzene rings

  16. Separation of 99Tc from low level radioactive liquid waste using anion exchange resin

    International Nuclear Information System (INIS)

    Sonar, N.L.; Mittal, V.K.; Dhara, Amrita; Thakur, D.A.; Valsala, T.P.; Vishwaraj, I.

    2016-01-01

    Technetium-99 is one of the fission products with very high yield (∼6%) in thermal neutron induced fission of 235 U. 99 Tc exists as pertechnate ( 99 TcO 4 ) ion in reprocessing streams. The high solubility in water and high mobility of pertechnate ions, coupled with very high half life of 99 Tc (t1/2 = 2 × 105 y, âmax = 290 KeV) makes it a potential candidate for long term hazard to the environment. Major radionuclides present in the intermediate level waste (ILW) generated at reprocessing plant is conventionally treated by ion exchange method for removal of 137 Cs. The Low level effluent waste (LLW) from the IX column contains 99 Tc as a major isotope. Though the concentration of 99 Tc in the waste is in ppm level, the presence of molar level of competing nitrates makes its separation very difficult. Many efforts have been reported on selective separation of 99 Tc from various waste streams. In this paper, separation of 99 Tc from ion exchange column effluent waste stream using selected commercially available anion exchange resins has been detailed

  17. Management of regenerant effluent waste at reprocessing plant, Tarapur- a new approach

    Energy Technology Data Exchange (ETDEWEB)

    Chandra, Munish; Bajpai, D D; Mudaiya, Avinash; Varadarajan, N [Power Reactor Fuel Reprocessing Plant, Tarapur (India)

    1994-06-01

    Power Reactor Fuel Reprocessing (PREFRE) Plant, Tarapur has been processing zircaloy clad spent fuel arising from PHWR namely RAPS and MAPS. The plant has been provided with a water pool to receive and store the irradiated fuel assemblies from the reactor site for an interim period before they are taken up for chop-leach and further reprocessing by PUREX process. This paper highlights the important and innovative modifications like introduction of a cation exchanger for water polishing and using nitric acid as regenerant. The regenerant effluent (nitric acid) is recycled to the main process cells where it is mixed and further treated along with process waste stream. This is a step towards minimising effluent generation. The paper describes the advantages of modified system like operational simplification, manpower, man-rem saving and minimising release of activity to environment. 3 figs., 4 tabs.

  18. Resin-based preparation of HTGR fuels: operation of an engineering-scale uranium loading system

    International Nuclear Information System (INIS)

    Haas, P.A.

    1977-10-01

    The fuel particles for recycle of 233 U to High-Temperature Gas-Cooled Reactors are prepared from uranium-loaded carboxylic acid ion exchange resins which are subsequently carbonized, converted, and refabricated. The development and operation of individual items of equipment and of an integrated system are described for the resin-loading part of the process. This engineering-scale system was full scale with respect to a hot demonstration facility, but was operated with natural uranium. The feed uranium, which consisted of uranyl nitrate solution containing excess nitric acid, was loaded by exchange with resin in the hydrogen form. In order to obtain high loadings, the uranyl nitrate must be acid deficient; therefore, nitric acid was extracted by a liquid organic amine which was regenerated to discharge a NaNO 3 or NH 4 NO 3 solution waste. Water was removed from the uranyl nitrate solution by an evaporator that yielded condensate containing less than 0.5 ppM of uranium. The uranium-loaded resin was washed with condensate and dried to a controlled water content via microwave heating. The loading process was controlled via in-line measurements of the pH and density of the uranyl nitrate. The demonstrated capacity was 1 kg of uranium per hour for either batch loading contractors or a continuous column as the resin loading contractor. Fifty-four batch loading runs were made without a single failure of the process outlined in the chemical flowsheet or any evidence of inability to control the conditions dictated by the flowsheet

  19. Corrosion of steel drums containing cemented ion-exchange resins as intermediate level nuclear waste

    Science.gov (United States)

    Duffó, G. S.; Farina, S. B.; Schulz, F. M.

    2013-07-01

    Exhausted ion-exchange resins used in nuclear reactors are immobilized by cementation before being stored. They are contained in steel drums that may undergo internal corrosion depending on the presence of certain contaminants. The objective of this work is to evaluate the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins with different aggressive species. The corrosion potential and the corrosion rate of the steel, and the electrical resistivity of the matrix were monitored for 900 days. Results show that the cementation of ion-exchange resins seems not to pose special risks regarding the corrosion of the steel drums. The corrosion rate of the steel in contact with cemented ion-exchange resins in the absence of contaminants or in the presence of 2.3 wt.% sulphate content remains low (less than 0.1 μm/year) during the whole period of the study (900 days). The presence of chloride ions increases the corrosion rate of the steel at the beginning of the exposure but, after 1 year, the corrosion rate drops abruptly reaching a value close to 0.1 μm/year. This is probably due to the lack of water to sustain the corrosion process. When applying the results obtained in the present work to estimate the corrosion depth of the steel drums containing the cemented radioactive waste after a period of 300 years, it is found that in the most unfavourable case (high chloride contamination), the corrosion penetration will be considerably lower than the thickness of the wall of the steel drums. Cementation of ion-exchange resins does not seem to pose special risks regarding the corrosion of the steel drums that contained them; even in the case the matrix is highly contaminated with chloride ions.

  20. Method of processing decontaminating liquid waste

    International Nuclear Information System (INIS)

    Kusaka, Ken-ichi

    1989-01-01

    When decontaminating liquid wastes are processed by ion exchange resins, radioactive nuclides, metals, decontaminating agents in the liquid wastes are captured in the ion exchange resins. When the exchange resins are oxidatively deomposed, most of the ingredients are decomposed into water and gaseous carbonic acid and discharged, while sulfur ingredient in the resins is converted into sulfuric acid. In this case, even less oxidizable ingredients in the decontaminating agent made easily decomposable by oxidative decomposition together with the resins. The radioactive nuclides and a great amount of iron dissolved upon decontamination in the liquid wastes are dissolved in sulfuric acid formed. When the sulfuric acid wastes are nuetralized with sodium hydroxide, since they are formed into sodium sulfate, which is most popular as wastes from nuclear facilities, they can be condensated and solidified by existent waste processing systms to thereby facilitate the waste processing. (K.M.)

  1. Properties of the Carboxylate ion exchange resins; Karboxylatjonbytarmassans egenskaper

    Energy Technology Data Exchange (ETDEWEB)

    Allard, Bert; Dario, Maarten [Oerebro Univ. (Sweden); Boren, Hans [Linkoepings Univ. (Sweden); Torstenfelt, Boerje [Swedpower, Stockholm (Sweden); Puigdomenech, Ignasi; Johansson, Claes [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)

    2002-09-01

    Weakly acidic, carboxylic resin has been selected, together with strong base anion resins, for water purification at the Forsmark 1 and 2 reactors. For the strong (but not the weak) ion exchange resin the Nuclear Power Inspectorate has given permission to dispose the spent resins in the SFR 1 (the Final Repository for Radioactive Operational Waste). This report gives a review of the carboxylic resins and comes to the conclusion that the resins are very stable and that there should not exist any risks for increased leaching of radionuclides from SFR 1 if these resins are disposed (compared to the strong resins)

  2. Tests of the use of cation exchange organic resins for the decontamination of radioactive aqueous effluents

    International Nuclear Information System (INIS)

    Bourdrez, Jean; Girault, Jacques; Wormser, Gerald

    1962-01-01

    The authors report tests performed in laboratory and results obtained during an investigation of the use of synthetic ion exchangers for the decontamination of radioactive effluents of moderate activity level and with a non neglectable salt loading. Resins are used under sodium form and regenerated after each fixing operation. Once decontaminated and free of its disturbing ions, the regenerating agent (NaCl) is used for several operations. The authors present the used resins, the treated effluents, describe the tests, and discuss the obtained results [fr

  3. Ion Exchange Column Tests Supporting Technetium Removal Resin Maturation

    Energy Technology Data Exchange (ETDEWEB)

    Nash, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hamm, L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Smith, F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Morse, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2013-12-20

    The primary treatment of the tank waste at the DOE Hanford site will be done in the Waste Treatment and Immobilization Plant, currently under construction. The baseline plan for this facility is to treat the waste, splitting it into High Level Waste (HLW) and Low Activity Waste (LAW). Both waste streams are then separately vitrified as glass and sealed in canisters. The LAW glass will be disposed on site. There are currently no plans to treat the waste to remove technetium, so its disposition path is the LAW glass. Due to the soluble properties of pertechnetate and long half-life of 99Tc, effective management of 99Tc is important. Options are being explored to immobilize the supplemental LAW portion of the tank waste, as well as to examine the volatility of 99Tc during the vitrification process. Removal of 99Tc, followed by off-site disposal has potential to reduce treatment and disposal costs. A conceptual flow sheets for supplemental LAW treatment and disposal that could benefit from technetium removal will specifically examine removing 99Tc from the LAW feed stream to supplemental immobilization. SuperLig® 639 is an elutable ion exchange resin. In the tank waste, 99Tc is predominantly found in the tank supernate as pertechnetate (TcO4-). Perrhenate (ReO4-) has been shown to be a good non-radioactive surrogate for pertechnetate in laboratory testing for this ion exchange resin. This report contains results of experimental ion exchange distribution coefficient and column resin maturation kinetics testing using the resin SuperLig® 639a to selectively remove perrhenate from simulated LAW. This revision includes results from testing to determine effective resin operating temperature range. Loading tests were performed at 45°C, and the computer modeling was updated to include the temperature effects. Equilibrium contact testing indicated that this batch of

  4. Potential problems associated with ion-exchange resins used in the decontamination of light-water reactor systems

    International Nuclear Information System (INIS)

    Soo, P.; Adams, J.W.; Kempf, C.R.

    1987-01-01

    During a typical decontamination event, ion-exchange resin beds are used to remove corrosion products (radioactive and nonradioactive) and excess decontamination reagents from waste streams. The spent resins may be solidified in a binder, such as cement, or sealed in a high-integrity container (HIC) in order to meet waste stability requirements specified by the Nuclear Regulatory Commission. Lack of stability of low-level waste in a shallow land burial trench may lead to trench subsidence, enhanced water infiltration and waste leaching, which would result in accelerated transport of radionuclides and the complexing agents used for decontamination. The current program is directed at investigating safety problems associated with the handling, solidification and containerization of decontamination resin wastes. The three tasks currently underway include freeze-thaw cycling of cementitious and vinyl ester-styrene forms to determine if mechanical integrity is compromised, a study of the corrosion of container materials by spent decontamination waste resins, and investigations of resin degradation mechanisms

  5. Organic diagenesis in commercial nuclear wastes

    International Nuclear Information System (INIS)

    Toste, A.P.; Lechner-Fish, T.J.

    1988-01-01

    The nuclear industry currently faces numerous challenges. Large volumes of already existing wastes must be permanently disposed using environmentally acceptable technologies. Numerous criteria must be addressed before wastes can be permanently disposed. Waste characterization is certainly one of the key criteria for proper waste management. some wastes are complex melting pots of inorganics, radiochemicals, and, occasionally, organics. It is clear, for example, that organics have been used extensively in nuclear operations, such as waste reprocessing, and continue to be used widely as solvents, decontamination agents, etc. The authors have analyzed the organic content of many kinds of nuclear wastes, ranging from commercial to defense wastes. In this paper, the finale analyses are described of three commercial wastes: one waste from a pressurized water reactor (PWR) and two wastes from a boiling water reactor (BWR). The PWR waste is a boric acid concentrate waste. The two BWR wastes, BWR wastes Nos. 1 and 2, are evaporator concentrates of liquid wastes produced during the regeneration of ion-exchange resins used to purify reactor process water. In preliminary analyses, which were reported previously, a few know organics and myriad unknowns were detected. Recent reexamination of mass-spectral data, coupled with reanalysis of the wastes, has resulted in the firm identification of the unknowns. Most of the compounds, over thirty distinct organics, are derived from the degradation, or diagenesis, of source-term organics, revealing, for the first time, that organic diagenesis in commercial wastes is both vigorous and varied

  6. Method of solidifying powderous wastes

    International Nuclear Information System (INIS)

    Kakimoto, Akira; Miyake, Takashi; Sato, Shuichi; Inagaki, Yuzo.

    1985-01-01

    Purpose: To improve the properties of solidification products, in the case of solidifying powderous wastes with thermosetting resins. Method. A solvent for the solution of the thermosetting resin is admixed with the powderous wastes into a paste-like form prior to adding the resin to the wastes, which are then mixed with the resin solution. As the result, those solidification products having the specific gravity and the compression strength more excellent than those of the conventional ones, and much higher than the reference values can be obtained. (Kamimura, M.)

  7. New system applying image processor to automatically separate cation exchange resin and anion exchange resin for condensate demineralizer

    International Nuclear Information System (INIS)

    Adachi, Tsuneyasu; Nagao, Nobuaki; Yoshimori, Yasuhide; Inoue, Takashi; Yoda, Shuji

    2014-01-01

    In PWR plant, condensate demineralizer is equipped to remove corrosive ion in condensate water. Mixed bed packing cation exchange resin (CER) and anion exchange resin (AER) is generally applied, and these are regenerated after separation to each layer periodically. Since the AER particle is slightly lighter than the CER particle, the AER layer is brought up onto the CER layer by feeding water upward from the bottom of column (backwashing). The separation performance is affected by flow rate and temperature of water for backwashing, so normally operators set the proper condition parameters regarding separation manually every time for regeneration. The authors have developed the new separation system applying CCD camera and image processor. The system is comprised of CCD camera, LED lamp, image processor, controller, flow control valves and background color panel. Blue color of the panel, which is corresponding to the complementary color against both ivory color of AER and brown color of CER, is key to secure the system precision. At first the color image of the CER via the CCD camera is digitized and memorized by the image processor. The color of CER in the field of vision of the camera is scanned by the image processor, and the position where the maximum difference of digitized color index is indicated is judged as the interface. The detected interface is able to make the accordance with the set point by adjusting the flow rate of backwashing. By adopting the blue background panel, it is also possible to draw the AER out of the column since detecting the interface of the CER clearly. The system has provided the reduction of instability factor concerning separation of resin during regeneration process. The system has been adopted in two PWR plants in Japan, it has been demonstrating its stable and precise performance. (author)

  8. Phenolic cation exchange resin material for recovery of cesium and strontium

    Science.gov (United States)

    Ebra, Martha A.; Wallace, Richard M.

    1983-01-01

    A phenolic cation exchange resin with a chelating group has been prepared by reacting resorcinol with iminodiacetic acid in the presence of formaldehyde at a molar ratio of about 1:1:6. The material is highly selective for the simultaneous recovery of both cesium and strontium from aqueous alkaline solutions, such as, aqueous alkaline nuclear waste solutions. The organic resins are condensation polymers of resorcinol and formaldehyde with attached chelating groups. The column performance of the resins compares favorably with that of commercially available resins for either cesium or strontium removal. By combining Cs.sup.+ and Sr.sup.2+ removal in the same bed, the resins allow significant reduction of the size and complexity of facilities for processing nuclear waste.

  9. Study on Concrete Containing Recycled Aggregates Immersed in Epoxy Resin

    Directory of Open Access Journals (Sweden)

    Adnan Suraya Hani

    2017-01-01

    Full Text Available In recent decades, engineers have sought a more sustainable method to dispose of concrete construction and demolition waste. One solution is to crush this waste concrete into a usable gradation for new concrete mixes. This not only reduces the amount of waste entering landfills but also alleviates the burden on existing sources of quality natural concrete aggregates. There are too many kinds of waste but here constructions waste will be the priority target that should be solved. It could be managed by several ways such as recycling and reusing the concrete components, and the best choice of these components is the aggregate, because of the ease process of recycle it. In addition, recycled aggregates and normal aggregates were immersed in epoxy resin and put in concrete mixtures with 0%, 5%, 10% and 20% which affected the concrete mixtures properties. The strength of the concrete for both normal and recycled aggregates has increased after immersed the aggregates in epoxy resin. The percentage of water absorption and the coefficient of water permeability decreased with the increasing of the normal and the recycled aggregates immersed in epoxy resin. Generally the tests which have been conducted to the concrete mixtures have a significant results after using the epoxy resin with both normal and recycled aggregates.

  10. Method of solidifying radioactive waste

    International Nuclear Information System (INIS)

    Hasegawa, Akira; Mihara, Shigeru; Yamashita, Koji; Sauda, Kenzo.

    1988-01-01

    Purpose: To obtain satisfactory plastic solidification products rapidly and more conveniently from radioactive wastes. Method: liquid wastes contain, in addition to sodium sulfate as the main ingredient, nitrates hindering the polymerizing curing reactions and various other unknown ingredients, while spent resins contain residual cationic exchange groups hindering the polymerizing reaction. Generally, as the acid value of unsaturated liquid polyester resins is lower, the number of terminal alkyd resins is small, formation of nitrates is reduced and the polymerizing curing reaction is taken place more smoothly. In view of the above, radioactive wastes obtained by dry powderization or dehydration of radioactive liquid wastes or spent resins are polymerized with unsaturated liquid polyester resins with the acid value of less than 13 to obtain plastic solidification. Thus, if the radioactive wastes contain a great amount of polymerization hindering material such as NaNO 2 , they can be solidified rapidly and conveniently with no requirement for pre-treatment. (Kamimura, Y.)

  11. Fully-differential spectrophotometry determination of trace thorium in uranium-containing waste water separated by CL-TBP levextrel resin

    International Nuclear Information System (INIS)

    You Jiannan

    2000-01-01

    A method for separation by CL-TBP levextrel resin and determination of trace thorium in uranium-containing waste water by fully-differential spectrophotometry is developed. In 4 mol/L HNO 3 medium, in presence of tartaric acid, CL-TBP levextrel resin is used for adsorption of thorium and separating from other elements. The thorium on the resin is stripped by 4 mol/L HCl, with oxalic acid and urea as screening agent, thorium forms red complex with arsenazo III. The maximum absorption of the complex is at 668 nm, and the molar absorptivity is 1.27 x 10 5 L/(mol·cm) . The complex can be steady for 2.5 h. By regulating micro-current of differential spectrophotometry, the method can realize determination with high precision. Sensitivity of this method increase 10 times than usual spectrophotometry. The relative standard deviation is better than +- 5% and recovery of thorium is 99%-107%

  12. Preparation and Characterization of UPR/ LNR/ Glass Fiber Composite by using Unsaturated Polyester Resin (PET) from PET Wastes

    International Nuclear Information System (INIS)

    Siti Farhana Hisham; Ishak Ahmad; Rusli Daik

    2011-01-01

    UPR/ LNR/ glass fibre composite had been prepared by using unsaturated polyester resin (UPR) based from recycled PET product. PET waste was recycled by glycolysis process and the glycides product was then reacted with maleic anhydride to produce unsaturated polyester resin. The preparation of UPR/ LNR blends were conducted by varying the amount of LNR addition to the resin ranging from 0-7.5 % (wt). The composition of UPR/LNR blend with good mechanical properties had been selected as a matrix of the glass fiber reinforced composite. Glass fibre was also treated by (3-Amino propil)triethoxysilane as a coupling agent. From the result, the addition of 2.5 % LNR in UPR had showed the optimum mechanical and morphological properties where the elastomer particle's were well dispersed in the matrix with smaller size. The silane treatment on the glass fiber increased the tensile and impact strength values of the UPR/ LNR/ GF composite compared to untreated fiber reinforcement. (author)

  13. Physical and chemical durability of cement impregnated epoxy resin

    International Nuclear Information System (INIS)

    Suryantoro

    1997-01-01

    Immobilization of simulation radioactive waste contains Cs and Sr with cement impregnated epoxy resin has been done. Low level liquid waste in 30% weight mixed cement homogeneously and then set in its curing time about 28 days. Waste from was impregnated with epoxy resin (Bisphenol-A-diglycidylether) and use Triethylenteramin as catalyst. the sample of cement impregnated epoxy resin 2.5 cm x 2.5 cm in diameter and length was tested by Paul Weber. The compressive strength was obtained of 4.08 kN.cm - 2. The sochxlet apparatus was run on flow rate of 300 ml/hour at 100 o C and during 24 hours. The leaching rate of Cs was round on 5.5 x 10 - 4 g.cm - 2.d - 1 and Sr was 6.1 x 10 - 4 g.cm - 2.d - 1 (author)

  14. Selective recovery of a pyridine derivative from an aqueous waste stream containing acetic acid and succinonitrile with solvent impregnated resins

    NARCIS (Netherlands)

    Bokhove, J.; Visser, T.J.; Schuur, Boelo; de Haan, A.B.

    2015-01-01

    Solvent impregnated resins (SIRs) were evaluated for the recovery of pyridine derivatives from an aqueous waste-stream containing also acetic acid and succinonitrile. For this purpose, a new solvent was developed, synthesized and impregnated in Amberlite XAD4. Sorption studies were used to determine

  15. Westinghouse Modular Grinding Process - Enhancement of Volume Reduction for Hot Resin Supercompaction - 13491

    Energy Technology Data Exchange (ETDEWEB)

    Fehrmann, Henning [Westinghouse Electric Germany GmbH, Dudenstr. 44, D-68167 Mannheim (Germany); Aign, Joerg [Westinghouse Electric Germany GmbH, Global D and D and Waste Management, Tarpenring 6, D-22419 Hamburg (Germany)

    2013-07-01

    In nuclear power plants (NPP) ion exchange (IX) resins are used in several systems for water treatment. Spent resins can contain a significant amount of contaminates which makes treatment for disposal of spent resins mandatory. Several treatment processes are available such as direct immobilization with technologies like cementation, bitumisation, polymer solidification or usage of a high integrity container (HIC). These technologies usually come with a significant increase in final waste volume. The Hot Resin Supercompaction (HRSC) is a thermal treatment process which reduces the resin waste volume significantly. For a mixture of powdered and bead resins the HRSC process has demonstrated a volume reduction of up to 75 % [1]. For bead resins only the HRSC process is challenging because the bead resins compaction properties are unfavorable. The bead resin material does not form a solid block after compaction and shows a high spring back effect. The volume reduction of bead resins is not as good as for the mixture described in [1]. The compaction properties of bead resin waste can be significantly improved by grinding the beads to powder. The grinding also eliminates the need for a powder additive.Westinghouse has developed a modular grinding process to grind the bead resin to powder. The developed process requires no circulation of resins and enables a selective adjustment of particle size and distribution to achieve optimal results in the HRSC or in any other following process. A special grinding tool setup is use to minimize maintenance and radiation exposure to personnel. (authors)

  16. Commercial Ion Exchange Resin Vitrification in Borosilicate Glass

    International Nuclear Information System (INIS)

    Cicero-Herman, C.A.; Workman, P.; Poole, K.; Erich, D.; Harden, J.

    1998-05-01

    Bench-scale studies were performed to determine the feasibility of vitrification treatment of six resins representative of those used in the commercial nuclear industry. Each resin was successfully immobilized using the same proprietary borosilicate glass formulation. Waste loadings varied from 38 to 70 g of resin/100 g of glass produced depending on the particular resin, with volume reductions of 28 percent to 68 percent. The bench-scale results were used to perform a melter demonstration with one of the resins at the Clemson Environmental Technologies Laboratory (CETL). The resin used was a weakly acidic meth acrylic cation exchange resin. The vitrification process utilized represented a approximately 64 percent volume reduction. Glass characterization, radionuclide retention, offgas analyses, and system compatibility results will be discussed in this paper

  17. Evaluation of the resin oxidation process using Fenton's reagent

    International Nuclear Information System (INIS)

    Araujo, Leandro G.; Goes, Marcos M.; Marumo, Julio T.

    2013-01-01

    The ion exchange resin is considered radioactive waste after its final useful life in nuclear reactors. Usually, this type of waste is treated with the immobilization in cement Portland, in order to form a solid monolithic matrix, reducing the possibility of radionuclides release in to environment. Because of the characteristic of expansion and contraction of the resins in presence of water, its incorporation in the common Portland cement is limited in 10% in direct immobilization, causing high costs in the final product. A pre-treatment would be able to reduce the volume, degrading the resins and increasing the load capacity of this material. This paper is about a method of degradation of ion spent resins from the nuclear research reactor of Nuclear and Energy Research Institute (IPEN/CNEN-SP), Brazil, using the Fenton's reagent. The resin evaluated was a mixture of cationic and anionic resins. The reactions were conducted by varying the concentration of the catalyst (25 to 80 mM), with and without external heat. The time of reaction was two hours. The concentration of 50 mM of catalyst was the most effective in degrading approximately 99%. The resin degradation was confirmed by the presence of CaCO 3 as a white precipitate resulting from the reaction between the Ca(OH) 2 and the CO 2 from the resin degradation. It was possible to degrade the resins without external heating. The calcium carbonates showed no correlation with the residual resin mass. (author)

  18. The use of regeneration profiles as a tool to optimise the ...

    African Journals Online (AJOL)

    DRINIE

    2003-10-04

    Oct 4, 2003 ... change in the structure of the solid. At the start of a service cycle the predominant ionic ..... The reason for these changes was to remove as many of the hardness salts exchanged by the WAC resin as ... higher average top acid strength for the 2nd regeneration of 0.6%, against 0.5% for the 1st regeneration.

  19. Preparation of titanates and zeolites and their uses in radioactive waste management, particularly in the treatment of spent resins

    International Nuclear Information System (INIS)

    Hultgren, Aa.; Airola, C.; Forsberg, S.; Faelt, L.

    1983-05-01

    Work on the preparation of titanates and zeolites was started and their possible uses in the management of radioactive wastes proposed in the first years of the 1970's by the Department of Nuclear Chemistry at the Royal Institute of Technology in Stockholm and the Chemical Center of the University of Lund, respectively. The major part of these efforts was sponsored by the National Council for Radioactive Waste (Prav), while the concluding phase including an over-all system design study adapted to the Swedish nuclear power programme for an economic evaluation was sponsored by the KBS Division of the Swedish Nuclear Fuel Supply Co. The preparation work resulted in processes ready for industrial scale production of sorbents of qualities adequate for applications in radioactive waste treatment. The essential effort was devoted to the treatment of spent resins from nuclear power plants by transfer to their radioactive contents to titanates and zeolites, which can then be sintered to stabel ceramic bodies (the PILO process). The economic evaluation indicated a significant economic incentive for the introduction of the PILO process if an incineration step is included for all types of spent resins. The essential efforts and results from this programme are summarized in this report. (Authors)

  20. Utilization of cotton waste for regenerated cellulose fibres: Influence of degree of polymerization on mechanical properties.

    Science.gov (United States)

    De Silva, Rasike; Byrne, Nolene

    2017-10-15

    Cotton accounts for 30% of total fibre production worldwide with over 50% of cotton being used for apparel. In the process from cotton bud to finished textile product many steps are required, and significant cotton waste is generated. Typically only 30% of pre consumer cotton is recycled. Here we use cotton waste lint to produce regenerated cellulose fibres (RCF). We find the RCF from waste cotton lint had increased mechanical properties compared to RCF produced from wood pulp. We show that this is likely linked to the higher degree of polymerization (DP) of waste cotton lint. An ionic liquid is used to dissolve the cotton lint and the rheology of the spinning is measured. The properties of the RCF are characterized and compared to wood pulp RCF. Copyright © 2017 Elsevier Ltd. All rights reserved.

  1. Thermal behaviour of used resin during conditioning process

    International Nuclear Information System (INIS)

    Arsene, C.

    2016-01-01

    In the nuclear power plants using light water and heavy water as coolant, as well as in most waste treatment installations, the ion-exchange resins are used to purify water circuits. Since the resins retain both radionuclide and chemical impurities, it represents a low- and intermediate- radioactive waste that requires special management for storage and disposal. From experimental studies it was found that the conditioning of the used resin in bitumen has several advantages. But there are some disadvantages, too, one being the significant amount of gas produced during the bituminization process because of the high temperature (1200C). Besides water vapours, the condensable gas mixture (formed by a liquid fraction and an oil fraction) contains products generated from the partial decomposition of the resin and release of degradation products of bitumen: dimethyl and trimethylamine, methanol - compounds resulting from the destruction of functional groups and hydrocarbon fraction formed by n-paraffins (C6-C32), iso-paraffins and aromatics. (authors)

  2. Quantitative Analysis of Radionuclide for the Used Resin of the Primary Purification System in HANARO

    International Nuclear Information System (INIS)

    Lee, Mun; Kim, Myong Seop; Park, Se Il; Kim, Tae Whan; Kim, Dong Hun; Kim, Young Chil

    2005-01-01

    In HANARO, a 30 MW research reactor, the ion exchange resin has been used for the purification of the primary coolant system. The resin used in the primary coolant purification system is replaced with new one once every 3 months during 30 MW reactor operation. The extracted resin from the primary coolant purification system is temporarily stored in a shielding treatment of the reactor hall for radiation cooling. After the radiation level of resin decreases enough to be handled for the waste disposal, it is put into the waste drum, and delivered to the waste facility in KAERI. Recently, in this procedure, the quantitative analysis of radionuclide which is contained in resin is required to have more quantitative data for the disposal. Therefore, in this work, a preliminary study was performed to find a sampling method for the representation of the characteristics of radionuclide in the spent resin

  3. Vitrification of ion-exchange (IEX) resins: Advantages and technical challenges

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Peeler, D.K.; Cicero, C.A.

    1995-01-01

    Technologies are being developed by the US Department of Energy's (DOE) Savannah River Site (SRS) in conjunction with the Electric Power Research Institute (EPRI) and the commercial sector to convert low-level radioactive ion exchange (IEX) resin wastes from the nuclear utilities to solid stabilized waste forms for permanent disposal. One of the alternative waste stabilization technologies is vitrification of the resin into glass. Wastes can be vitrified at elevated temperatures by thermal treatment. One alternative thermal treatment is conventional Joule heated melting. Vitrification of wastes into glass is an attractive option because it atomistically bonds both hazardous and radioactive species in the glass structure, and volume reduces the wastes by 70-80%. The large volume reductions allow for large associated savings in disposal and/or long term storage costs

  4. Incineration of spent ion exchange resins in a triphasic mixture at Belgoprocess

    International Nuclear Information System (INIS)

    Deckers, J.; Luycx, P.

    2003-01-01

    Up to 1998, spent ion exchange resins have been fed to the incinerator in combination with various other solid combustible wastes at Belgoprocess. However, thanks to sustained efforts to reduce radioactive waste production in all nuclear facilities in Belgium, the annual production of solid combustible waste is now much too small to allow this practice to be continued. Since the incinerator at Belgoprocess is not capable of treating spent ion exchange resins as such, it was decided to adopt the use of foam as a carrier to feed the resins to the incinerator. The mixture is a pseudohomogeneous charged foam, ensuring easy handling and allowing incineration in the existing furance, while a number of additives may be included, such as oil to increase the calorific value of the mixture and accelerate combustion. The first incineration campaign of spent ion exchange resins in a triphasic foam mixture, in conjunction with other liquid and solid combustible wastes, will be started in January 2000. The foam, comprising 70% by weight of resins, 29% by weight of water and 1% by weight of surfactant will be pulverized in the incinerator through an injection lance, at a feed rate of 40 to 100 kg/h. The incinerator and associated off-gas treatment system can be operated at standard conditions. Belgoprocess is the subsidiary of the Belgian national agency for the management of radioactive waste, known by its Dutch and French acronyms, NIRAS and ONDRAF respectively. The company ensures the treatment, conditioning and interim storage of nearly all radioactive waste produced in Belgium. (orig.)

  5. Perchlorate adsorption and desorption on activated carbon and anion exchange resin.

    Science.gov (United States)

    Yoon, In-Ho; Meng, Xiaoguang; Wang, Chao; Kim, Kyoung-Woong; Bang, Sunbaek; Choe, Eunyoung; Lippincott, Lee

    2009-05-15

    The mechanisms of perchlorate adsorption on activated carbon (AC) and anion exchange resin (SR-7 resin) were investigated using Raman, FTIR, and zeta potential analyses. Batch adsorption and desorption results demonstrated that the adsorption of perchlorate by AC and SR-7 resin was reversible. The reversibility of perchlorate adsorption by the resin was also proved by column regeneration test. Solution pH significantly affected perchlorate adsorption and the zeta potential of AC, while it did not influence perchlorate adsorption and the zeta potential of resin. Zeta potential measurements showed that perchlorate was adsorbed on the negatively charged AC surface. Raman spectra indicated the adsorption resulted in an obvious position shift of the perchlorate peak, suggesting that perchlorate was associated with functional groups on AC at neutral pH through interactions stronger than electrostatic interaction. The adsorbed perchlorate on the resin exhibited a Raman peak at similar position as the aqueous perchlorate, indicating that perchlorate was adsorbed on the resin through electrostatic attraction between the anion and positively charged surface sites.

  6. The dissolution of organic ion exchange resins using iron-catalysed hydrogen peroxide

    International Nuclear Information System (INIS)

    Hawkings, N.; Horton, K.D.; Snelling, K.W.

    1980-10-01

    Feasibility studies have been made of the dissolution/partial decomposition of radioactive waste resins by means of iron-catalysed hydrogen peroxide. They have shown that the procedure is limited in its application and successfully treats only polystyrene/divinylbenzene-based resins. Evaporation of the final solution produces a solid residue which is difficult to handle and results in only a relatively small reduction in volume. It is concluded that the method could be used to dissolve specific waste resins for easier handling and disposal, but is not of general applicability. (author)

  7. Method of controlling radioactive waste processing systems

    International Nuclear Information System (INIS)

    Mikawa, Hiroji; Sato, Takao.

    1981-01-01

    Purpose: To minimize the pellet production amount, maximize the working life of a solidifying device and maintaining the mechanical strength of pellets to a predetermined value irrespective of the type and the cycle of occurrence of the secondary waste in the secondary waste solidifying device for radioactive waste processing systems in nuclear power plants. Method: Forecasting periods for the type, production amount and radioactivity level of the secondary wastes are determined in input/output devices connected to a control system and resulted signals are sent to computing elements. The computing elements forecast the production amount of regenerated liquid wastes after predetermined days based on the running conditions of a condensate desalter and the production amounts of filter sludges and liquid resin wastes after predetermined days based on the liquid waste processing amount or the like in a processing device respectively. Then, the mass balance between the type and the amount of the secondary wastes presently stored in a tank are calculated and the composition and concentration for the processing liquid are set so as to obtain predetermined values for the strength of pellets that can be dried to solidify, the working life of the solidifying device itself and the radioactivity level of the pellets. Thereafter, the running conditions for the solidifying device are determined so as to maximize the working life of the solidifying device. (Horiuchi, T.)

  8. Leaching of 60 Co and 137 Cs from spent ion exchange resins in ...

    Indian Academy of Sciences (India)

    Cement; radioactive waste; composite; waste management. Abstract. The leaching rate of 60Co and 137Cs from the spent cation exchange resins in cement–bentonite matrix has been studied. The solidification matrix was a standard Portland cement mixed with 290–350 (kg/m3) spent cation exchange resins, with or ...

  9. Column chromatographic separation of Y3+ from Sr2+ by polymeric ionizable crown ether resins

    International Nuclear Information System (INIS)

    Wood, D.J.; Elshani, S.; Wai, C.M.; Bartsch, R.A.; Huntley, M.; Hartenstein, S.

    1993-01-01

    Condensation polymers containing subunits of crown ether carboxylic acid monomers are effective stationary phases for the chromatographic separation of Y 3+ and Sr 2+ . The pH range and metal loading capacities for the resins have been determined under equilibrium conditions. The resin can be regenerated for repeated use without losing its separation capability. Altering the molecular structure of the monomer (sym-dibenzo-16-crown-5-oxyacetic acid) by an alkyl substitution on the macrocyclic cavity decreases the loading capacity of the resin

  10. Treatment of spent ion exchange resins IAEA research coordination programme

    International Nuclear Information System (INIS)

    Balu, K.; Bhatia, S.C.; Wattal, P.K.; Chanana, N.

    1981-09-01

    Spent ion-exchange resins arising from steam condensate systems, reactor coolant clean-up systems and rad-waste procession, are considered as a specific solid waste management problem. This is the second report on the product characterisation with respect to thermal properties, flammability, bio-organic degradation and leaching behaviours. All these studies are based on polyester-styrene polymer as a matrix for fixation of these spent Ix-resins. Choice of this matrix was dealt with in the first report. (author)

  11. Radioactive spent resins conditioning by the hot super-compaction process

    International Nuclear Information System (INIS)

    Roth, Andreas; Centner, Baudouin; Lemmens, Alain

    2007-01-01

    Spent ion exchanger media are considered to be problematic waste that, in many cases, requires special approaches and precautions during its immobilization to meet the acceptance criteria for disposal. The waste acceptance criteria define, among others, the quality of waste forms for disposal, and therefore will sometimes define appropriate treatment options. The selection of treatment options for spent ion exchange materials must consider their physical and chemical characteristics. Basically, the main methods for the treatment of spent organic ion exchange materials, following to pretreatment methods are: - Direct immobilization, producing a stable end product by using Cement, Bitumen, Polymer or High Integrity Containers, - The destruction of the organic compounds by using Thermochemical processes or Oxidation to produce an inorganic intermediate product that may or may not be further conditioned for storage and/or disposal, - The complete removal of the resin inner structural water by a thermal process. After a thorough technical economical analysis, Tractebel Engineering selected the Resin Hot Compaction Process to be installed at Tihange Nuclear Power Plant. The Resin Hot Compaction Process is used to make dense homogenous organic blocks from a wide range of particulate waste. In this process spent resins are first dewatered and dried to remove the inner structural water content. The drying takes place in a drying vessel that holds the contents of two 200 L drums (Figure). In the oil heated drying and mixing unit, the resins are heated to the necessary process temperature for the hot pressing step and then placed into special metal drums, which are automatically lidded and immediately transferred to a high force compactor. After high force compaction the pellets are transferred to a measuring unit, where the dose rate, height and weight are automatically measured and recorded. A volume reduction factor of approximately up to four (depending on the type of

  12. Diffusion in composite materials made of thermosetting resins

    International Nuclear Information System (INIS)

    Morin, Bruno.

    1981-03-01

    The embedding process of low and medium level radioactive wastes in thermosetting resins allows their containment in a solid matrix. During storage the risk of circulation of water is possible. The aim of this containment process is to prevent radionuclide migration in environment. Ion migration through membranes of thermosetting resins alone or filler added were measured to evaluate released radioactivity by embedded blocks with time and to compare the different embedding formulas. Water influence on diffusion was taken into account considering that radioactive wastes dispersion is faster in a wet medium than in a dry one [fr

  13. Merger of waste of lead in polyester resin for application of protection barriers

    International Nuclear Information System (INIS)

    Barros, Frieda Saicla; Paredes, Ramon Siguifredo Cortes

    2010-01-01

    This paper's main objective is the use of powdered lead waste obtained from the recycling of batteries embedded in polyester resin to be used as a protective barrier in environments subject to radiation. The aim is to enable a more economical procedure which is faster and more practical for use in walls where it is necessary to use protective barrier. This justification is reinforced having in sight the use of byproducts generated by industries, recycling them and using them as components in the development of barriers against ionizing radiation. In this study, we observed the morphological and physical-chemical properties of isolated and associated materials and performance analysis of the composite with respect to the attenuation properties for gamma rays, by means of experimental tests. For mixtures with 40% of waste lead, value referenced in mass, we obtained satisfactory results on the screen. Thus, we were able to combine the good performance of the composite with the reduction of environmental liabilities, in view of the recycling process of lead. (author)

  14. EPICOR-II resin degradation results from first resin samples of PF-8 and PF-20

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Sanders, R.D. Sr.

    1985-12-01

    The 28 March 1979 accident at Three Mile Island Unit 2 released approximately 560,000 gallons of contaminated water to the Auxiliary and Fuel Handling Buildings. The water was decontaminated using a demineralization system called EPICOR-II developed by Epicor, Inc. The Low-Level Waste Data Base Development - EPICOR-II Resin/Liner Investigation Project is studying the chemical and physical conditions of the synthetic ion exchange resins found in several EPICOR-II prefilters. This report summarizes results and analyses of the first sampling of ion exchange resins from EPICOR-II prefilters PE-8 and -20. Results are compared with baseline data from tests performed on unirradiated Epicor, Inc. resins to determine if degradation has occurred due to the high internal radiation dose received by the EPICOR-II resins. Results also are compared with recent findings on resin degradation by Battelle Columbus Laboratories and Brookhaven National Laboratory. Analyses comparing test results of resins from EPICOR-II prefilters PF-8 and -20 with unirradiated resins obtained from Epicor, Inc. show resin degradation has occurred in some of the EPICOR-II resins examined. The mechanism of degradation is compared with work of other researchers and is consistent with their findings. The strong acid cation resins (divinylbenzene, styrene base structure) are losing effective cross-linking along with scission of functional groups and are experiencing first an increase and eventually a decrease in total exchange capacity as the absorbed radiation dose increases. The phenolic cation resins (phenol-formaldehyde base structure) show a loss of effective cross-linking and oxidation of the polymer chain. Analyses of resins removed from EPICOR-II prefilters PF-8 and -20 over the next several years should show a further increase in degradation

  15. The efficiency of macroporous polystyrene ion-exchange resins in natural organic matter removal from surface water

    Directory of Open Access Journals (Sweden)

    Urbanowska Agnieszka

    2017-01-01

    Full Text Available Natural water sources used for water treatment contains various organic and inorganic compounds. Surface waters are commonly contaminated with natural organic matter (NOM. NOM removal from water is important e.g. due to lowering the risk of disinfection by-product formation during chlorination. Ion exchange with the use of synthetic ion-exchange resins is an alternative process to typical NOM removal approach (e.g. coagulation, adsorption or oxidation as most NOM compounds have anionic character. Moreover, neutral fraction could be removed from water due to its adsorption on resin surface. In this study, applicability of two macroporous, polystyrene ion exchange resins (BD400FD and A100 in NOM removal from water was assessed including comparison of treatment efficiency in various process set-ups and conditions. Moreover, resin regeneration effectivity was determined. Obtained results shown that examined resins could be applied in NOM removal and it should be noticed that column set-up yielded better results (contrary to batch set-up. Among the examined resins A100 one possessed better properties. It was determined that increase of solution pH resulted in a slight decrease in treatment efficiency while higher temperature improved it. It was also observed that regeneration efficiency was comparable in both tested methods but batch set-up required less reagents.

  16. Structure, mechanical and thermal behaviour of mixtures of polyester resin and dental ceramic waste

    Science.gov (United States)

    Peña Rodríguez, G.; Martínez Maldonado, L.; Dulce Moreno, H. J.

    2016-02-01

    The tensile strength and bending strength, structure and thermal behaviour of mixtures of polyester resin (P-2000) and powders (ASTM sieve 200, dental ceramic wastes (dentals impressions, alginate and gypsum) was reported. The samples consisted of mixtures with percentage weights of 50-50%, 60-40%, 70-30%, 80-20%, 90-10%, where the resin was the majority phase, the Mekc (4% wt) was used as catalyst. The structure was studied using SEM and XRD, the thermal behaviour using DSC, TGA and DMA, while the mechanical strength was tested using standards ASTM D790 and D638. Irregular morphology and presence of small agglomerations was observed, with particle sizes between 29.63 and 38.67μm, the presence of different phases of calcium sulphate was found, and that to the increasing the concentration of the powder, the materials becomes more crystalline, increasing its density. An average service temperature of 69.15±4.60°C was found. Vickers hardness values are reported in ranges from 18.65 to 27.96. Considering the elastic modules was established that the materials become more rigid by having more powder concentration.

  17. Synthesis, structural characterization, and performance evaluation of resorcinol-formaldehyde (R-F) ion-exchange resin

    International Nuclear Information System (INIS)

    Hubler, T.L.; Franz, J.A.; Shaw, W.J.; Bryan, S.A.; Hallen, R.T.; Brown, G.N.; Bray, L.A.; Linehan, J.C.

    1995-08-01

    The 177 underground storage tanks at the DOE's Hanford Site contain an estimated 180 million tons of high-level radioactive wastes. It is desirable to remove and concentrate the highly radioactive fraction of the tank wastes for vitrification. Resorcinol-formaldehyde (R-F) resin, an organic ion-exchange resin with high selectivity and capacity for the cesium ion, which is a candidate ion-exchange material for use in remediation of tank wastes. The report includes information on the structure/function analysis of R-F resin and the synthetic factors that affect performance of the resin. CS-100, a commercially available phenol-formaldehyde (P-F) resin, and currently the baseline ion-exchanger for removal of cesium ion at Hanford, is compared with the R-F resin. The primary structural unit of the R-F resin was determined to consist of a 1,2,3,4-tetrasubstituted resorcinol ring unit while CS-100, was composed mainly of a 1,2,4-trisubstituted ring. CS-100 shows the presence of phenoxy-ether groups, and this may account for the much lower decontamination factor of CS-100 for cesium ion. Curing temperatures for the R-F resin were found to be optimal at 105--130C. At lower temperatures, insufficient curing, hence crosslinking, of the polymer resin occurs and selectivity for cesium drops. Curing at elevated temperatures leads to chemical degradation. Optimal particle size for R-F resin is in the range of 20--50 mesh-sized particles. R-F resin undergoes chemical degradation or oxidation which destroys ion-exchange sites. The ion-exchange sites (hydroxyl groups) are converted to quinones and ketones. CS-100, though it has much lower performance for cesium ion-exchange, is significantly more chemically stable than R-F resin. To gamma radiation, CS-100 is more radiolytically stable than R-F resin

  18. Synthesis, structural characterization, and performance evaluation of resorcinol-formaldehyde (R-F) ion-exchange resin

    Energy Technology Data Exchange (ETDEWEB)

    Hubler, T.L.; Franz, J.A.; Shaw, W.J.; Bryan, S.A.; Hallen, R.T.; Brown, G.N.; Bray, L.A.; Linehan, J.C.

    1995-08-01

    The 177 underground storage tanks at the DOE`s Hanford Site contain an estimated 180 million tons of high-level radioactive wastes. It is desirable to remove and concentrate the highly radioactive fraction of the tank wastes for vitrification. Resorcinol-formaldehyde (R-F) resin, an organic ion-exchange resin with high selectivity and capacity for the cesium ion, which is a candidate ion-exchange material for use in remediation of tank wastes. The report includes information on the structure/function analysis of R-F resin and the synthetic factors that affect performance of the resin. CS-100, a commercially available phenol-formaldehyde (P-F) resin, and currently the baseline ion-exchanger for removal of cesium ion at Hanford, is compared with the R-F resin. The primary structural unit of the R-F resin was determined to consist of a 1,2,3,4-tetrasubstituted resorcinol ring unit while CS-100, was composed mainly of a 1,2,4-trisubstituted ring. CS-100 shows the presence of phenoxy-ether groups, and this may account for the much lower decontamination factor of CS-100 for cesium ion. Curing temperatures for the R-F resin were found to be optimal at 105--130C. At lower temperatures, insufficient curing, hence crosslinking, of the polymer resin occurs and selectivity for cesium drops. Curing at elevated temperatures leads to chemical degradation. Optimal particle size for R-F resin is in the range of 20--50 mesh-sized particles. R-F resin undergoes chemical degradation or oxidation which destroys ion-exchange sites. The ion-exchange sites (hydroxyl groups) are converted to quinones and ketones. CS-100, though it has much lower performance for cesium ion-exchange, is significantly more chemically stable than R-F resin. To gamma radiation, CS-100 is more radiolytically stable than R-F resin.

  19. Tests of the use of cation exchange organic resins for the decontamination of radioactive aqueous effluents; Essais d'emploi des resines organiques echangeuses de cations pour la decontamination des effluents aqueux radioactifs

    Energy Technology Data Exchange (ETDEWEB)

    Bourdrez, Jean; Girault, Jacques; Wormser, Gerald

    1962-12-14

    The authors report tests performed in laboratory and results obtained during an investigation of the use of synthetic ion exchangers for the decontamination of radioactive effluents of moderate activity level and with a non neglectable salt loading. Resins are used under sodium form and regenerated after each fixing operation. Once decontaminated and free of its disturbing ions, the regenerating agent (NaCl) is used for several operations. The authors present the used resins, the treated effluents, describe the tests, and discuss the obtained results [French] Dans ce rapport nous faisons le point des essais et resultats obtenus au laboratoire au cours d'une etude concernant l'utilisation des echangeurs d'ions synthetiques pour la decontamination des effluents radioactifs de niveau d'activite moyenne et de charge en sels non negligeable. Les resines sont employees sous forme sodique et regenerees apres chaque operation de fixation. Le regenerant decontamine et debarrasse de ses ions genants est utilise pour plusieurs operations d'elution. Les seuls residus a stocker proviennent d'une part eventuellement d'un pretraitement de l'effluent, d'autre part, dans tous les cas des precipites consecutifs a la purification de l'eluant dont le traitement chimique est plus aise et donne lieu a des boues beaucoup moins volumineuses qu'une coprecipitation effectuee sur la totalite de l'effluent. (auteurs)

  20. Effect of pore structure on the removal of clofibric acid by magnetic anion exchange resin.

    Science.gov (United States)

    Tan, Liang; Shuang, Chendong; Wang, Yunshu; Wang, Jun; Su, Yihong; Li, Aimin

    2018-01-01

    The effect of pore structure of resin on clofibric acid (CA) adsorption behavior was investigated by using magnetic anion exchange resins (ND-1, ND-2, ND-3) with increasing pore diameter by 11.68, 15.37, 24.94 nm. Resin with larger pores showed faster adsorption rates and a higher adsorption capacity because the more opened tunnels provided by larger pores benefit the CA diffusion into the resin matrix. The ion exchange by the electrostatic interactions between Cl-type resin and CA resulted in chloride releasing to the solution, and the ratio of released chloride to CA adsorption amount decreased from 0.90 to 0.65 for ND-1, ND-2 and ND-3, indicating that non-electrostatic interactions obtain a larger proportional part of the adsorption into the pores. Co-existing inorganic anions and organic acids reduced the CA adsorption amounts by the competition effect of electrostatic interaction, whereas resins with more opened pore structures weakened the negative influence on CA adsorption because of the existence of non-electrostatic interactions. 85.2% and 65.1% adsorption amounts decrease are calculated for resin ND-1 and ND-3 by the negative influence of 1 mmol L -1 NaCl. This weaken effect of organic acid is generally depends on its hydrophobicity (Log Kow) for carboxylic acid and its ionization degree (pKb) for sulfonic acid. The resins could be reused with the slightly decreases by 1.9%, 3.2% and 5.4% after 7 cycles of regeneration, respectively for ND-1, ND-2 and ND-3, suggesting the ion exchange resin with larger pores are against its reuse by the brine solution regeneration. Copyright © 2017 Elsevier Ltd. All rights reserved.

  1. Catalytic oxidative pyrolysis of spent organic ion exchange resins from nuclear power plants

    International Nuclear Information System (INIS)

    Sathi Sasidharan, N.; Deshingkar, D.S.; Wattal, P.K.; Shirsat, A.N.; Bharadwaj, S.R.

    2005-08-01

    The spent IX resins from nuclear power reactors are highly active solid wastes generated during operations of nuclear reactors. Catalytic oxidative pyrolysis of these resins can lead to high volume reduction of these wastes. Low temperature pyrolysis of transition metal ion loaded IX resins in presence of nitrogen was carried out in order to optimize catalyst composition to achieve maximum weight reduction. Thermo gravimetric analysis of the pyrolysis residues was carried out in presence of air in order to compare the oxidative characteristics of transition metal oxide catalysts. Copper along with iron, chromium and nickel present in the spent IX resins gave the most efficient catalyst combination for catalytic and oxidative pyrolysis of the residues. During low temperature catalytic pyrolysis, 137 Cesium volatility was estimated to be around 0.01% from cationic resins and around 0.1% from anionic resins. During oxidative pyrolysis at 700 degC, nearly 10 to 40% of 137 Cesium was found to be released to off gases depending upon type of resin and catalyst loaded on to it. The oxidation of pyrolytic residues at 700 degC gave weight reduction of 15% for cationic resins and 93% for anionic resins. Catalytic oxidative pyrolysis is attractive for reducing weight and volume of spent cationic resins from PHWRs and VVERs. (author)

  2. Characterization of monolith block of spent resin cementation

    International Nuclear Information System (INIS)

    Prayitno; Endro-Kismolo; Isman MT

    1996-01-01

    Spent resin immobilization process with cement was done to prevent release of radionuclide in the ultimate storage or disposal. The varied Composition of water/cement ratio in the cementation process were 0.3; 0.4; 0.5 and the various weight of resin waste are 25 g, 37.5 g and 50 gram. The compressive strength of the various water/cement ratio without spent resin was bigger than 0.3. This investigation proved that the compressive strength of Tiga Roda cement was bigger than those of Gresik cement or Nusantara cement. The compressive of the cement block of were the spent resin cementation was influenced by the water/cement ratio and the total spent resin addition. The best condition reached at the water/cement ratio of 0.3 and 25 gram spent resin, was compressive strength of 17.86 N/mm 2 . Leaching rate of the various weight composition of spent resin cementation for 91 days were between 10 -2 - 10 -4 gram.cm -2 .day -1

  3. Characterization and Regeneration of Pt-Catalysts Deactivated in Municipal Waste Flue Gas

    DEFF Research Database (Denmark)

    Rasmussen, Søren Birk; Kustov, Arkadii; Due-Hansen, Johannes

    2006-01-01

    Severe deactivation was observed for industrially aged catalysts used in waste incineration plants and tested in lab-scale. Possible compounds that cause deactivation of these Pt-based CO oxidation catalysts have been studied. Kinetic observations of industrial and model catalysts showed...... that siloxanes were the most severe catalyst poisons, although acidic sulfur compounds also caused deactivation. Furthermore, a method for on-site regeneration without shutdown of the catalytic flue gas cleaning system has been developed, i.e. an addition of H-2/N-2 gas to the off-gas can completely restore...... the activity of the deactivated catalysts. (c) 2006 Elsevier B.V. All rights reserved....

  4. Sorption of Uranium Ions from Their Aqueous Solution by Resins Containing Nanomagnetite Particles

    Directory of Open Access Journals (Sweden)

    Mahmoud O. Abd El-Magied

    2016-01-01

    Full Text Available Magnetic amine resins composed of nanomagnetite (Fe3O4 core and glycidyl methacrylate (GMA/N,N′-methylenebisacrylamide (MBA shell were prepared by suspension polymerization of glycidyl methacrylate with N,N′-methylenebisacrylamide in the presence of nanomagnetite particles and immobilized with different amine ligands. These resins showed good magnetic properties and could be easily retrieved from their suspensions using an external magnetic field. Adsorption behaviors of uranium ions on the prepared resins were studied. Maximum sorption capacities of uranium ions on R-1 and R-2 were found to be 92 and 158 mg/g. Uranium was extracted successfully from three granite samples collected from Gabal Gattar pluton, North Eastern Desert, Egypt. The studied resins showed good durability and regeneration using HNO3.

  5. Storage process of large solid radioactive wastes

    International Nuclear Information System (INIS)

    Morin, Bruno; Thiery, Daniel.

    1976-01-01

    Process for the storage of large size solid radioactive waste, consisting of contaminated objects such as cartridge filters, metal swarf, tools, etc, whereby such waste is incorporated in a thermohardening resin at room temperature, after prior addition of at least one inert charge to the resin. Cross-linking of the resin is then brought about [fr

  6. Encapsulation of nuclear wastes

    International Nuclear Information System (INIS)

    Arnold, J.L.; Boyle, R.W.

    1978-01-01

    Toxic waste materials are encapsulated by the method wherein the waste material in liquid or finely divided solid form is uniformly dispersed in a vinyl ester resin or an unsaturated polyester and the resin cured under conditions that the exotherm does not rise above the temperature at which the integrity of the encapsulating material is destroyed

  7. EDF specifications on nuclear grade resins

    International Nuclear Information System (INIS)

    Mascarenhas, Darren; Gressier, Frederic; Taunier, Stephane; Le-Calvar, Marc; Ranchoux, Gilles; Marteau, Herve; Labed, Veronique

    2012-09-01

    Ion exchange resins are widely used across EDF, especially within the nuclear division for the purification of water. Important applications include primary circuit, secondary circuit and effluent treatment, which require high quality nuclear grade resins to retain the dissolved species, some of which may be radioactive. There is a need for more and more efficient purification in order to decrease worker dose during maintenance but also to decrease volumes of radioactive resin waste. Resin performance is subject to several forms of degradation, including physical, chemical, thermal and radioactive, therefore appropriate resin properties have to be selected to reduce such effects. Work has been done with research institutes, manufacturers and on EDF sites to select these properties, create specifications and to continuously improve on these specifications. An interesting example of research regarding resin performance is the resin degradation under irradiation. Resins used in the CVCS circuit of EDF nuclear power plants are subject to irradiation over their lifetime. A study was carried out on the effects of total integrated doses of 0.1, 1 and 10 MGy on typically used EDF mixed bed resins in a 'mini-CVCS' apparatus to simultaneously test actual primary circuit fluid. The tests confirmed that the resins still perform efficiently after a typical CVCS radiation dose. Certain resins also need additional specifications in order to maintain the integrity of the particular circuits they are used in. Recently, EDF has updated its requirements on these high purity nuclear grade resins, produced generic doctrines for all products and materials used on site which include resins of all grades, and as a result have also updated a guide on recommended resin usage for the French fleet of reactors. An overview of the evolutions will be presented. (authors)

  8. Future perspectives of resin-based dental materials.

    Science.gov (United States)

    Jandt, Klaus D; Sigusch, Bernd W

    2009-08-01

    This concise review and outlook paper gives a view of selected potential future developments in the area of resin-based biomaterials with an emphasis on dental composites. A selection of key publications (1 book, 35 scientific original publications and 1 website source) covering the areas nanotechnology, antimicrobial materials, stimuli responsive materials, self-repairing materials and materials for tissue engineering with direct or indirect relations and/or implications to resin-based dental materials is critically reviewed and discussed. Connections between these fields and their potential for resin-based dental materials are highlighted and put in perspective. The need to improve shrinkage properties and wear resistance is obvious for dental composites, and a vast number of attempts have been made to accomplish these aims. Future resin-based materials may be further improved in this respect if, for example nanotechnology is applied. Dental composites may, however, reach a completely new quality by utilizing new trends from materials science, such as introducing nanostructures, antimicrobial properties, stimuli responsive capabilities, the ability to promote tissue regeneration or repair of dental tissues if the composites were able to repair themselves. This paper shows selected potential future developments in the area of resin-based dental materials, gives basic and industrial researchers in dental materials science, and dental practitioners a glance into the potential future of these materials, and should stimulate discussion about needs and future developments in the area.

  9. Necessity and opportunity of building a mobile facility for conditioning the spent ionic resins

    International Nuclear Information System (INIS)

    Barariu, Gheorghe

    2002-01-01

    The ionic resins utilized in nuclear industry for purification of different working fluids become contaminated by important amounts of radionuclides incorporated in their structure. Thus, these resins become intermediate-level radioactive wastes, containing radioisotopes of rather long lifetimes, and so their conditioning is required to ensure binding and isolation of the radionuclides over a period long enough, until their decay. Binding the radionuclides in stable matrices and their isolation by means of engineered barriers ensures protection of environment against contamination. In the frame of Radioactive Waste Management Program related to Cernavoda NPP, optimal technologies for processing the radioactive waste generated during the NPP operation were developed and also a site for final disposal of the low and intermediate level waste was selected. The last one is a modern, ground repository with multiple engineered barriers based on the natural properties of the host geological formation. The final disposal concept implies building a conditioning facility to ensure: a super-compacting stage of the already compacted wastes; cementing the uncompactable wastes; cementing spent filters; cementing the pellets resulted from the super-compacting stage and drums with conditioned wastes in concrete modules; cementing the organic wastes; cementing the spent ionic resins by means of a mobile facility. The spent ionic resins removed from the technologic purification systems of the station are discharged by means of demineralized water and transferred, corresponding to source, to one of the 3 storage 200 m 3 basins from the basement of the CANDU 6 NPP servicing building. To avoid the danger of spent resins being cemented at basin's bottom the basins should be emptied at every 10-15 years. To comply with IAEA requirements concerning the radioactive waste final disposal, a procedure for establishing the quantitative concentration of solidification agents and additives

  10. Monitoring for 99Tc in borehole waters using an extraction chromatographic resin

    International Nuclear Information System (INIS)

    Davis, T.M.; Nelson, D.M.; Thompson, E.G.

    1993-01-01

    A technique using an extraction chromatographic resin for the analysis of 99 Tc in borehole waters is presented. The method involves one pass of the sample through two resins, a rinse of the resins, then counting of one of the resins by liquid scintillation counting. The technique is intended to be used as a screening method of 99 Tc. The procedure employs commercially available materials, generates little waste and requires very little operator time

  11. Pyrolysis of Spent Ion Exchange Resins

    International Nuclear Information System (INIS)

    Braehler, Georg; Slametschka, Rainer

    2012-09-01

    Ion exchangers (IEX in international language) are used to remove radionuclides from the primary coolant in all nuclear power stations with a water cooling circuit. This is done by continuously removing a volume of coolant from the primary circuit and passing it through coolers, filters and the ion exchange beds. Cation and anion exchangers, in the form of coarse-grained resin beads in pressurized-water reactors and as finely ground powdered resins in boiling water reactors, are used. The trend for new power stations is to exploit all the possibilities for avoiding the generation of contaminated liquids and then to clean, as far as possible, the solutions that are nevertheless generated using ion exchange for it to be possible to dispose of them as non-radioactive waste. This relieves the burden on evaporator facilities, or means that these can even be dispensed with entirely. Regeneration is possible in principle, but little use is made of it. As the regeneration usual in conventional technologies is not employed in nuclear power stations, it is necessary to dispose of this material as radioactive waste. On the international level, a great number of processes are offered that are intended to meet the relevant national regulations, and these will be discussed in brief with their advantages and disadvantages. The aim is then to find a process which reduces the volume, yields an inert or mineralized product, works at temperatures of no more than approximately 600 deg. C and can be run in a simple facility. Originally, the pyrolysis process was developed to treat liquid organic waste from reprocessing. A typical application is the decomposition of spent solvent (TBP, tributyl phosphate, mixed with kerosene). In this process TBP is pyrolyzed together with calcium hydroxide in a fluidized bed facility at temperatures of around 500 deg. C, the calcium hydroxide reacts with the phosphate groups directly to form calcium pyrophosphate which contains all the radioactivity

  12. Cost effectiveness of detritiating water with resin columns

    International Nuclear Information System (INIS)

    Drake, R.H.; Williams, D.S.

    1997-10-01

    There are technologies in use for cleaning up concentrated tritiated process water. These are not cost effective for tritiated water with low concentrations of tritium. There are currently no cost-effective technologies for cleaning up low-tritium-concentration tritiated water, such as most tritiated groundwater, spent fuel storage basin water, or underground storage tank water. Resin removal of tritium from tritiated water at low concentrations (near the order of magnitude of drinking water standard maximums) is being tested on TA-SO (Los Alamos National Laboratory's Liquid Radioactive Waste Treatment Facility) waste streams. There are good theoretical and test indications that this may be a technologically effective means of removing tritium from tritiated water. Because of likely engineering design similarity, it is reasonable to anticipate that a resin column system's costs will be similar to some common commercial water treatment systems. Thus, the potential cost effectiveness of a resin treatment system offers hope for treating tritiated water at affordable costs. The TA-50 resin treatment cost projection of $18 per 1,000 gallons is within the same order of magnitude as cost data for typical commercial groundwater cleanup projects. The prospective Los Alamos National Laboratory (LANL) resin treatment system at $18 per 1,000 gallons appears to have a likely cost advantage of at least an order of magnitude over the competing, developmental, water detritiation technologies

  13. Volume reduction of ion exchange resin by a pyrolysis technique

    International Nuclear Information System (INIS)

    Matsuda, M.; Funabashi, K.; Uchida, S.; Kikuchi, M.

    1985-01-01

    Volume reduction techniques, such as incineration and acid digestion, of spent ion exchange resins from nuclear power plants are being developed with a view toward reducing radioactive waste volume and also making the final waste form more stable. The authors chose pyrolysis as a technique that can be done at low operating temperatures and low gas flow rates in a reactor vessel. Fundamental experiments were performed to clarify the resin pyrolysis characteristics, and the optimum pyrolysis temperature was determined. Consequently, a pilot plant with a treatment capacity of approx. 50 kg/batch was constructed based on the results. Using the pilot plant, the authors are now performing pyrolysis of the resins and solidification of their residues. This report will give the results of fundamental experiments and pilot plant tests

  14. Extraction of Co ions from ion-exchange resin by supercritical carbon dioxide

    International Nuclear Information System (INIS)

    Ju, Min Su; Koh, Moon Sung; Yang, Sung Woo; Park, Kwang Heon; Kim, Hak Won; Kim, Hong Doo

    2005-01-01

    There are a number of liquid treatment processes for eliminating radioactive ionic contaminants in nuclear facilities. One of the most common treatment methods for aqueous streams is the use of ion exchange, which is a well-developed technique that has been employed for many years in the nuclear industry. More specifically speaking, systems that ion exchange method is applied to in nuclear power plants are liquid radioactive waste treatment system, chemical and volume control system, steam generator blowdown treatment system, and service water supply system. During the operation of nuclear power plants, radioactive contaminants such as Co-60, Mn-54, Fe-59 and Cs-137 are contained in liquid radioactive wastes. And the wastes containing small amount of uranium are generated in nuclear fuel cycle facilities. To treat the liquid radioactive waste, we usually install ion exchangers rather than evaporators due to their simplicity and effectiveness, and this trend is increasing. However, the ion exchange process produces large volume of spent organic resin, and has some problems of radiation damage and thermal instability. And the reuse of the resin is limited due to the degradation of ion-exchanging ability. For this reason, were should consider a better method to expand the lifetime of the resin or to reduce the volume of radioactive resin wastes by extracting radioactive contaminants located in the resin. Supercritical fluid CO 2 has many good points as a process solvent that include low viscosity, negligible surface tension, and variable selectivity. And supercritical fluids have physical properties of both liquid and gas such as good penetration with a high dissolution capability. Supercritical fluids have been widely used in extraction, purification, and recovery processes. A number of workers applied supercritical CO 2 solvent for cleaning of precision devices and waste treatments. Since supercritical CO 2 has its mild critical point at 31 and 73.8bar as .deg. C

  15. Study PWA8 resin for chromatographic uranium concentration

    International Nuclear Information System (INIS)

    Coceancigh, Herman; Ramella, J. L.; Marrero, Julieta; Jiménez Rebagliati, Raúl

    2013-01-01

    For many years nuclear industry have been using resins as filler of chromatographic columns. These methods are specific and give reliable results in different applications, for those reasons are extremely useful as separation process. Currently the nuclear industry is growing and this brings new issues such as the need of reduction of the amount of waste, the optimization of the production process and others that the chromatography could solve with great results. AMBERLITETM PWA8 resin is an anion exchange resin which can be used for the removal of uranium from drinking water. In addition to high exchange capacity, this resin has excellent physical stability and a wide range of pH in which is operational. With the idea of concentrating uranium from wastes solution as main goal we made different experiments to understand the AMBERLITETM PWA8 and obtain the most important characteristics like; pH working range; capacity; activation and elution procedures. These procedures were developed and optimized the capacity was determined using a batch experiment and we obtain that the maximum capacity is 882,5 U ug /resin gr at a pH of 4,2. Following on from these results chromatographic experiments were performed in which both were obtained the percentage of recovery and the concentration factor. The percent recovery (% R) calculated as the percentage ratio between the total mass and the load mass eluted (% R = eluted mass / total mass * 100) was 94% with a concentration factor of 5 times From these results it is intended to concentrate wastes solutions from the fuel cycle processes with two main goals: decreasing volume for storage and for future reusing of the uranium coming from production. (author)

  16. Method of solidifying radioactive wastes

    International Nuclear Information System (INIS)

    Tomita, Toshihide; Minami, Yuji; Matsuura, Hiroyuki; Kageyama, Hisashi; Kobori, Junzo.

    1984-01-01

    Purpose: To perform the curing sufficiently even when copper hydroxide that interferes the curing reaction is contained in radioactive wastes. Method: Solidification of radioactive wastes containing copper hydroxide using thermoset resins is carried out under the presence of an alkaline material. The thermoset resin used herein is an polyester resin comprising unsaturated polyester and a polymerizable monomer. The alkaline substance usable herein can include powder or an aqueous solution of hydroxides or oxides of sodium, magnesium, calcium or the like. (Yoshino, Y.)

  17. Electrodialytic decontamination of spent ion exchange resins

    International Nuclear Information System (INIS)

    Nott, B.R.

    1982-01-01

    Development of a novel electrodialytic decontamination process for the selective removal of radioactive Cs from spent ion exchange resins containing large amounts of Li is described. The process involves passage of a dc electric current through a bed of the spent ion exchange resin in a specially designed electrodialytic cell. The radiocesium so removed from a volume of the spent resin is concentrated onto a much smaller volume of a Cs selective sorbent to achieve a significant radioactive waste volume reduction. Technical feasibility of the electrodialytic resin decontamination process has been demonstrated on a bench scale with a batch of simulated spent ion exchange resin and using potassium cobalt ferrocyanide as the Cs selective sorbent. A volume reduction factor between 10 and 17 has been estimated. The process appears to be economically attractive. Improvements in process economics can be expected from optimization of the process. Other possible applications of the EDRD process have been identified

  18. MECHANICAL REGENERATION OF SAND WASTE

    Directory of Open Access Journals (Sweden)

    D. I. Gnir

    2005-01-01

    Full Text Available The experimental activation of the sand regenerator of the firm SINTO is carried out at ОАО “MZOO". It is shown that sand grains are cleared from films of binding agents, that allows to use the treated sand for preparation of agglutinant and core sands.

  19. Development Of An Approach To Modeling Loading And Elution Of Spherical Resorcinol Formaldehyde Ion-Exchange Resin

    International Nuclear Information System (INIS)

    Aleman, S.; Hamm, L.; Smith, F.

    2011-01-01

    The current strategy for removal of cesium from the Hanford waste stream is ion-exchange using spherical Resorcinol-Formaldehyde (sRF) resin. The original resin of choice was granular SuperLig 644 resin and during testing of this resin several operational issues were identified. For example, the granular material had a high angle of internal friction resulting in fragmentation of resin particles along its edges during cycling and adverse hydraulic performance. Efforts to replace SuperLig 644 were undertaken and one candidate was the granular Resorcinol-Formaldehyde (RF) resin where experience with this cation exchanger dates back to the late 1940's. To minimize hydraulic concerns a spherical version of RF was developed and several different chemically produced batches were created. The 5E-370/641 batch of sRF was selected and for the last decade numerous studies have been performed (e.g., batch contact tests, column loading and elution tests). The Waste Treatment Plant (WTP) flowsheet shows that the aqueous phase waste stream will have a wide range of ionic concentrations (e.g., during the loading step 0-3 M free OH, 5+ M Na, 0-1 M K, 0-3 M NO 3 ). Several steps are required in the ion-exchange process to achieve the required Cs separation factors: loading, displacement, washing, elution, and regeneration. The sRF resin will be operated over a wide range in pH (i.e., pH of 12-14 during the loading step and pH of 0.01-1 during the elution step). During some of these steps very high levels of counter-ions and co-ions will be present within the aqueous phase. Alternative process feeds are under consideration as well (e.g., sodium levels as high as 8 M and column operation up to 45 C during loading, reduced and recycled HNO 3 during elution). In order to model the performance of sRF resin through an entire ion-exchange cycle, a more robust isotherm model is required. To achieve this more robust isotherm model requires knowledge of the numbers and kinds of fixed

  20. DEVELOPMENT OF AN APPROACH TO MODELING LOADING AND ELUTION OF SPHERICAL RESORCINOL FORMALDEHYDE ION-EXCHANGE RESIN

    Energy Technology Data Exchange (ETDEWEB)

    Aleman, S.; Hamm, L.; Smith, F.

    2011-10-03

    The current strategy for removal of cesium from the Hanford waste stream is ion-exchange using spherical Resorcinol-Formaldehyde (sRF) resin. The original resin of choice was granular SuperLig 644 resin and during testing of this resin several operational issues were identified. For example, the granular material had a high angle of internal friction resulting in fragmentation of resin particles along its edges during cycling and adverse hydraulic performance. Efforts to replace SuperLig 644 were undertaken and one candidate was the granular Resorcinol-Formaldehyde (RF) resin where experience with this cation exchanger dates back to the late 1940's. To minimize hydraulic concerns a spherical version of RF was developed and several different chemically produced batches were created. The 5E-370/641 batch of sRF was selected and for the last decade numerous studies have been performed (e.g., batch contact tests, column loading and elution tests). The Waste Treatment Plant (WTP) flowsheet shows that the aqueous phase waste stream will have a wide range of ionic concentrations (e.g., during the loading step 0-3 M free OH, 5+ M Na, 0-1 M K, 0-3 M NO{sub 3}). Several steps are required in the ion-exchange process to achieve the required Cs separation factors: loading, displacement, washing, elution, and regeneration. The sRF resin will be operated over a wide range in pH (i.e., pH of 12-14 during the loading step and pH of 0.01-1 during the elution step). During some of these steps very high levels of counter-ions and co-ions will be present within the aqueous phase. Alternative process feeds are under consideration as well (e.g., sodium levels as high as 8 M and column operation up to 45 C during loading, reduced and recycled HNO{sub 3} during elution). In order to model the performance of sRF resin through an entire ion-exchange cycle, a more robust isotherm model is required. To achieve this more robust isotherm model requires knowledge of the numbers and kinds of

  1. Characterization and disposal of ion exchange resins used in nuclear installations

    International Nuclear Information System (INIS)

    Flores E, R.M.; Ortiz O, H.B.; Olguin G, M.T.; Emeterio H, M.; Garcia M, H.

    2006-01-01

    To dispose of an appropriate way the used ion exchange resins so much in the pool water purification systems of the TRIGA Mark III reactor like in the JS6500 gamma irradiator, of the National Institute of Nuclear Research, were carried out a series of analytic nuclear techniques and complementary conventional to those recommended by the ASTM, with the object of to control and to manage 14 lots of worn out resins appropriately. For its were identified the radioactive isotopes, the resins type, the grade of chemical pollution and the physicochemical degradation of the same ones. The lots of resins that didn't contain radioactive isotopes its were regenerated in an usual way, as long as those that if they controlled them they selected options for its final disposition. The first selected option was the extraction method of ion radioactive isotopes, concentrating the elution product by evaporation. As second option it was carried out the resins stabilization damaged by micro-encapsulation by forged to ambient temperature, using an organic polymer. Previous to the immobilization the resins were pretreated by vacuum drying, pulverization and thermal drying, however before carrying out this last, it was carried out a thermal gravimetric analysis to determine the drying conditions of the resins avoiding its chemical decomposition. (Author)

  2. Corrosion of steel drums containing immobilized ion exchange-resins and incineration ashes

    International Nuclear Information System (INIS)

    Marotta, F.; Schulz Rodriguez, F.M.; Farina, Silvia B.; Duffo, Gustavo S.

    2009-01-01

    The Argentine Atomic Energy Commission (CNEA) is responsible for developing the management nuclear waste disposal programme. This programme contemplates the design and construction of a facility for the final disposal of intermediate-level radioactive wastes. The proposed model is a near-surface monolithic repository similar to those in operation in El Cabril, Spain. The design of this type of repository is based on the use of multiple, independent and redundant barriers. The intermediate radioactive waste consists mostly in spent ionic exchange resins and filters from the nuclear power plants, research reactors and radioisotopes production facilities. The spent resins, as well as the incineration ashes, have to be immobilized before being stored to improve leach resistance of waste matrix and to maintain mechanical stability for safety requirements. Generally, cementation processes have been used as immobilization techniques for economical reasons as well as for being a simple operation. The immobilized resins and incineration ashes are thus contained in steel drums that, in turn, can undergo corrosion depending on the ionic content of the matrix. This work is a part of a systematic study of the corrosion susceptibility of steel drums in contact with immobilized cemented exchange-resins with different types and contents of aggressive species and incineration ashes. To this purpose, a special type of specimen was manufactured to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix are being monitored along time. The aggressive species studied were chloride ions (the main ionic species present in nature) and sulphate ions (produced during the radiolysis process of the cationic exchange-resins after cementation). Preliminary results show the strong effect of chloride on the corrosion susceptibility of the steel. Monitoring will continue for

  3. Purification of degraded TBP solvent using macroreticular anion exchange resin

    International Nuclear Information System (INIS)

    Kartha, P.K.S.; Kutty, P.V.E.; Janaradanan, C.; Ramanujam, A.; Dhumwad, R.K.

    1989-01-01

    Tri-n-butyl phosphate (TBP) diluted with a suitable diluent is commonly used for solvent extraction in Purex process for the recovery of uranium and plutonium from irradiated nuclear fuels. This solvent gets degraded due to various factors, the main degradation product being dibutyl phosphoric acid (HDBP). A solvent cleanup step is generally incorporated in the process for removing the degradation products from the used solvent. A liquid-liquid cleanup system using sodium carbonate or sodium hydroxide solution is routinely used. Considering certain advantages, like the possibility of loading the resin almost to saturation capacity and the subsequent disposal of the spent resin by incineration and the feasibility of adopting it to the process, a liquid-solid system has been tried as an alternate method, employing various available macroreticular anion exchange resins in OH - form for the sorption of HDBP from TBP. After standardizing the various conditions for the satisfactory removal of HDBP from TBP using synthetic mixtures, resins were tested with process solvent in batch contacts. The parameters studied were (1) capacity of different resins for HDBP sorption (2) influence of acidity, uranium and HDBP on the sorption behaviour of the latter (3) removal of fission products from the solvent by the resin and (4) regeneration and recycling of the resin. (author). 2 figs., 13 tabs., 17 refs

  4. Bituminous solidification, disposal, transport and burial of spent ion-exchange resins. Part of a coordinated programme on treatment of spent ion exchange resins

    International Nuclear Information System (INIS)

    Mozes, G.; Kristof, M.

    1983-07-01

    The project dealing with the incorporation of spent ion-exchange resins into bitumen was performed within the Agency coordinated research programme on treatment of spent ion-exchange resins. Physical and chemical properties of commercial ion-exchange resins, bitumens and bituminized resins were studied. It was shown that bitumen with low oil content and with a softening point of 60-70 deg. C are applicable for the incorporation of resins. The final waste form is allowed to contain maximum 50% resin. The comprehensive study of the biological resistance of B-30 bitumen was performed. That showed that any bacteriological attack can be regarded as generally insignificant. A continuously operating technology was realized on a semi-plant scale. The best operating conditions of this technology were determined. On the basis of the experience gained from the experiments a design of the bituminization plant of 50m 3 dry resin/year treatment capacity was proposed

  5. Solvent impregnated resin for isolation of U(VI) from industrial wastes

    International Nuclear Information System (INIS)

    Karve, M.; Rajgor, R.V.

    2008-01-01

    A solid-phase extraction method based upon impregnation of Cyanex 302 (bis(2,4,4- trimethylpentyl)mono-thio-phosphinic acid) on Amberlite XAD-2 resin is proposed for isolation of U(VI) from uranmicrolite ore tailing samples and industrial effluent samples. U(VI) was sorbed from nitric acid media on the solvent-impregnated resin (SIR) and was recovered completely with 1.0 M HCl. Based upon sorption behavior of U(VI) with Cyanex 302, it was quantitatively sorbed on the SIR in a dynamic method, while the other metal ions were not sorbed by the modified resin. The preparation of impregnated resin is simple, based upon physical interaction of the extractant and solid support, has good sorption capacity for U(VI), and is also reliable for detection of traces of U(VI). (authors)

  6. Treatment of solid waste highly contaminated by alpha emitters: Recent developments of leaching process with continuous electrolyte regeneration

    International Nuclear Information System (INIS)

    Madic, C.; Lecomte, M.; Vigreux, B.

    1990-01-01

    Development of processes for leaching solid waste contaminated by alpha or alphaβgamma emitters has been pursued at the Nuclear Research Center in Fontenay-aux-Roses, France with the recent active commissioning of two pilot facilities: the Elise glove box system in February 1987 and the Prolixe shielded hot cell in March 1988. The Elise facility is designed to handle alpha waste and the Prolixe facility is designed to handle alphaβgamma waste. The common goal of the studies conducted in these facilities is to define the operating conditions for declassification of solid waste, i.e. to ensure that the alpha concentration of this waste will be less than 3.7 x 10 6 Bq/kg after treatment, packaging and decay prior to storage in surface repositories. The leaching process developed is mainly based on the continuous electrolytic regeneration of an aggressive agent, AgII, which can induce the dissolution of PuO 2 , the most difficult compound to remove from the solid waste. This paper summarizes recent achievements in the development of this process. 11 refs., 8 figs., 6 tabs

  7. Treatment of solid waste highly contaminated by alpha emitters: recent developments of leaching process with continuous electrolyte regeneration

    International Nuclear Information System (INIS)

    Madic, C.; Lecomte, M.

    1990-01-01

    Development of processes for leaching solid waste contaminated by alpha or alpha/beta/gamma emitters has been pursued at the Nuclear Research Center in Fontenay-aux-Roses, France with the recent active commissioning of two pilot facilities: the Elise glove box system in February 1987 and the Prolixe shielded hot cell in March 1988. The Elise facility is designed to handle alpha waste and the Prolixe facility is designed to handle alpha/beta/gamma waste. The common goal of the studies conducted in these facilities is to define the operating conditions for declassification of solid waste, i.e. to ensure that the alpha concentration of this waste will be less than 3.7 x 10 6 Bq/kg after treatment, packaging and decay prior to storage in surface repositories. The leaching process developed is mainly based on the continuous electrolytic regeneration of an aggressive agent, AgII, which can induce the dissolution of PuO 2 , the most difficult compound to remove from the solid waste. This paper summarizes recent achievements in the development of this process

  8. Laboratory studies on enhancing the throughput of demineraliser units by proper selection of ion exchange resins

    Energy Technology Data Exchange (ETDEWEB)

    Rao, S V; Unny, V K.P.; Shetiya, R S [Reactor Services and Maintenance Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    In water treatment plants of nuclear reactors, type I strong base anion resins are extensively used in anion units along with strong acid cation resins in cation units for demineralisation of water. There is another kind of strong base anion resin called type II resin which is chemically slightly different from type I resin. Type II resins differ from the type I in that one of the methyl groups in the quaternary ammonium functional group of the latter, is replaced by an ethanol group. This results in lower basicity of the resin and hence, higher regeneration efficiency. However, the resin is not like a weak base resin in its properties since it is capable of removing silica from water, though not with the same efficiency as that of the type I strong base resin. Studies were carried out to compare the performance of presently used type I resins with that of type II resins supplied by local manufacturers to assess the suitability of the latter for use in the DM plants of Trombay reactors. This paper discusses the results of the studies. (author). 2 figs., 6 tabs.

  9. Immobilisation Of Spent Ion Exchange Resins Using Portland Cement Blending With Organic Material

    International Nuclear Information System (INIS)

    Zalina Laili; Mohd Abdul Wahab; Nur Azna Mahmud

    2014-01-01

    Immobilisation of spent ion exchange resins (spent resins) using Portland cement blending with organic material for example bio char was investigated. The performance of cement-bio char matrix for immobilisation of spent ion exchange resins was evaluated based on their compression strength and leachability under different experimental conditions. The results showed that the amount of bio char and spent resins loading effect the compressive strength of the waste form. Several factors affecting the leaching behaviour of immobilised spent resins in cement-bio char matrix. (author)

  10. Engineering study for the treatment of spent ion exchange resin resulting from nuclear process applications

    International Nuclear Information System (INIS)

    Place, B.G.

    1990-09-01

    This document is an engineering study of spent ion exchange resin treatment processes with the purpose of identifying one or more suitable treatment technologies. Classifications of waste considered include all classes of low-level waste (LLW), mixed LLW, transuranic (TRU) waste, and mixed TRU waste. A total of 29 process alternatives have been evaluated. Evaluation parameters have included economic parameters (both total life-cycle costs and capital costs), demonstrated operability, environmental permitting, operational availability, waste volume reduction, programmatic consistency, and multiple utilization. The results of this study suggest that there are a number of alternative process configurations that are suitable for the treatment of spent ion exchange resin. The determinative evaluation parameters were economic variables (total life-cycle cost or capital cost) and waste volume reduction. Immobilization processes are generally poor in volume reduction. Thermal volume reduction processes tend to have high capital costs. There are immobilization processes and thermal volume reduction processes that can treat all classifications of spent ion exchange resin likely to be encountered. 40 refs., 19 figs., 17 tabs

  11. Use of strong anion exchange resins for the removal of perfluoroalkylated substances from contaminated drinking water in batch and continuous pilot plants.

    Science.gov (United States)

    Zaggia, Alessandro; Conte, Lino; Falletti, Luigi; Fant, Massimo; Chiorboli, Andrea

    2016-03-15

    In recent years abnormally high levels of perfluoroalkylated substances (PFAS) have been detected both in surface and underground water sampled in an area covering approximately 150 square kilometers in the Veneto region (Italy) indicating the presence of a pollution point source (fluorochemicals production plant). Adsorption on granular activated carbon is an emergency measure which is poorly effective requiring frequent replacement. This work focuses on the application of three strong anion exchange resins (Purolite® A520E, A600E and A532E) for the removal of traces of PFOA, PFOS, PFBA and PFBS (concentration of hundreds of ng L(-1)) from drinking water. This technology is attractive for the possibility of reusing resins after an in-situ regeneration step. A strong relationship between the hydrophobicity of the exchange functional group of the resin and its capacity in removing PFAS exists. A600E (non hydrophobic) and A520E (fairly hydrophobic) show a reduced sorption capacity compared to A532E (highly hydrophobic). While A600E and A520E can be regenerated with solvent-less dilute solutions of non-toxic NH4Cl and NH4OH, A532E requires concentrated solutions of methanol or ethanol and 1% NH4Cl and for the sake of this work it was regarded as non-regenerable. The volume of regeneration effluents requiring incineration can be efficiently reduced by more than 96.5% by using reverse osmosis coupled with under-vacuum evaporation. Transmission electron analysis on saturated resins showed that large molecular macro-aggregates of PFAS can form in the intraparticle pores of resin indicating that ion exchange is not the only mechanism involved in PFAS removal. Copyright © 2015 Elsevier Ltd. All rights reserved.

  12. Cesium Isotherm Testing with Spherical Resorcinol-Formaldehyde Resin at High Sodium Concentrations

    Energy Technology Data Exchange (ETDEWEB)

    Russell, Renee L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fiskum, Sandra K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smoot, Margaret R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rinehart, Donald E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-04-01

    Washington River Protection Solutions (WRPS) is developing a Low-Activity Waste Pretreatment System (LAWPS) to provide low-activity waste (LAW) directly to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste Facility for immobilization. The pretreatment that will be conducted on tank waste supernate at the LAWPS facility entails filtration to remove entrained solids and cesium (Cs) ion exchange to remove Cs from the product sent to the WTP. Currently, spherical resorcinol-formaldehyde (sRF) resin (Microbeads AS, Skedsmokorset, Norway) is the Cs ion exchange resin of choice. Most work on Cs ion exchange efficacy in Hanford tank waste has been conducted at nominally 5 M sodium (Na). WRPS is examining the possibility of processing supernatant at high Na concentrations—up to 8 M Na—to maximize processing efficiency through the LAWPS. Minimal Cs ion exchange work has been conducted at 6 M and 8 M Na concentrations..

  13. Method of solidifying radioactive laundry wastes

    International Nuclear Information System (INIS)

    Yasumura, Keijiro

    1984-01-01

    Purpose: To enable to solidify radioactive laundry wastes containing non-ionic liquid detergents less solidifiable by plastic solidification process in liquid laundry wastes for cloths or the likes discharged from a nuclear power plant. Method: Radioactive laundry wastes are solidified by using plastic solidifying agent comprising, as a main ingredient, unsaturated polyester resins and methylmethacrylate monomers. The plastic solidifying agents usable herein include, for example, unsaturated polyester resins prepared by condensating maleic anhydride and phthalic anhydride with propylene glycol and incorporated with methylmethacrylate monomers. The mixing ratio of the methylmethacrylate monomers is preferably 30 % by weight based on the unsaturated polyester resins. (Aizawa, K.)

  14. 77 FR 58416 - Comparative Environmental Evaluation of Alternatives for Handling Low-Level Radioactive Waste...

    Science.gov (United States)

    2012-09-20

    ... for Handling Low-Level Radioactive Waste Spent Ion Exchange Resins From Commercial Nuclear Power... Radioactive Waste Spent Ion Exchange Resins from Commercial Nuclear Power Reactors. DATES: Please submit... Evaluation of Alternatives for Handling Low-Level Radioactive Waste Spent Ion Exchange Resins from Commercial...

  15. Testing waste forms containing high radionuclide loadings

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  16. Application of a weak base anion exchange resin for recovery of uranium at Uravan, Colorado, U.S.A

    International Nuclear Information System (INIS)

    Gardner, N.E.; Kunin, R.

    1976-01-01

    Resin ion-exchange technology has been used to recover uranium at the Uravan, Colorado plant for over 18 years; however, since the end of U.S. Atomic Energy Commission purchase of U 3 O 8 concentrate in 1970, it has become necessary to develop techniques for upgrading the product to meet the more stringent specifications of private sales. The standard gel type quaternary ammonium anion exchange resin had been used previously. The development of the tertiary amine anion exchange resin, Amberlite XE-299, led to an experimental program of laboratory and pilot plant work to evaluate the resin on actual plant solutions. General information on ion-exchange resin structure and chemistry is discussed. Summary data of specific test work on loading the resin, various elution schemes, resin regeneration and product purity from the pilot plant tests and comments on actual plant operation using Amberlite XE-299 resin are presented. (author)

  17. Electrodialysis Treatment of Industrial Waste Waters That Generaters Little Sludge; Tratamiento de aguas residuales industriales por electrodialisis, tecnica que genera pocos lodos

    Energy Technology Data Exchange (ETDEWEB)

    Savastoma Mariel, C. [Universidad de Cantabria. Santander (Spain)

    2002-07-01

    Electrodialysis with bipolar membranes is reviewed as a separation technique and details are shown. The method is used in several process applications but its adoption looks particularly attractive in the treatment of waste water streams consisting primarily of diluted residual salts and electrolytes. It is argued that the main advantage shown by the technology compared to the classical treatment consisting of adding lime to neutralize the waste waters resides in tis avoiding the generation of voluminous sludges thus permitting savings in the management and disposal of the mud. Other interesting applications are those operations where generating acidity and/or alkalinity with addition of no extra chemicals is desirable, such as the regeneration of ion exchange resins. (Author) 34 refs.

  18. Immobilization of Ion Exchange radioactive resins of the TRIGA Mark III Nuclear Reactor

    International Nuclear Information System (INIS)

    Garcia Martinez, H.

    1999-01-01

    In the last decades many countries in the world have taken interest in the use, availability, and final disposal of dangerous wastes in the environment, within these, those dangerous wastes that contain radioactive material. That is why studies have been made on materials used as immobilization agent of radioactive waste that may guarantee its storage for long periods of time under drastic conditions of humidity, temperature change and biodegradation. In mexico, the development of different applications of radioactive material in the industry, medicine and investigation, have generated radioactive waste, sealed and open sources, whose require a special technological development for its management and final disposal. The present work has as a finality to develop the process and define the agglutinating material, bitumen, cement and polyester resin that permits immobilization of resins of Ionic Exchange contaminated by Barium 153, Cesium 137, Europium 152, Cobalt 60 and Manganese 54 generated from the nuclear reactor TRIGA Mark III. Ionic interchange contaminated resin must be immobilized and is analysed under different established tests by the Mexican Official Standard NOM-019-NUCL-1995 L ow level radioactive wastes package requirements for its near-surface final disposal. Immobilization of ionic interchange contaminated resins must count with the International Standards applicable in this process; in these standards, the following test must be taken in prototype examples: Free-standing water, leachability, compressive strength, biodegradation, radiation stability, thermal stability and burning rate. (Author)

  19. Method of processing radioactive wastes

    International Nuclear Information System (INIS)

    Takahashi, Toshihiko; Maruko, Morihisa; Takamura, Yoshiyuki.

    1981-01-01

    Purpose: To effectively separate radioactive claddings from the slurry of wasted ion exchange resins containing radioactive claddings. Method: Wasted ion exchange resins having radioactive claddings (fine particles of iron oxides or hydroxide adhered with radioactive cobalt) are introduced into a clad separation tank. Sulfuric acid or sodium hydroxide is introduced to the separation tank to adjust the pH value to 3 - 6. Then, sodium lauryl sulfate is added for capturing claddings and airs are blown from an air supply nozzle to generate air bubbles. The claddings are detached from the ion exchange resins and adhered to the air bubbles. The air bubbles adhered with the claddings float up to the surface of the liquid wastes and then forced out of the separation tank. (Ikeda, J.)

  20. Cement solidification of spent ion exchange resins produced by the nuclear industry

    International Nuclear Information System (INIS)

    Jaouen, C.; Vigreux, B.

    1988-01-01

    Cement solidification technology has been applied to spent ion exchange resins for many years in countries throughout the world (at reactors, research centers and spent fuel reprocessing plants). Changing specifications for storage of radioactive waste have, however, confronted the operators of such facilities with a number of problems. Problems related both to the cement solidification process (water/cement/resin interactions and chemical interactions) and to its utilization (mixing, process control, variable feed composition, etc.) have often led waste producers to prefer other, polymer-based processes, which are very expensive and virtually incompatible with water. This paper discusses research on cement solidification of ion exchange resins since 1983 and the development of application technologies adapted to nuclear service conditions and stringent finished product quality requirements

  1. Regeneration of Waste Edible Oil by the Use of Virgin and Calcined Magnesium Hydroxide as Adsorbents.

    Science.gov (United States)

    Ogata, Fumihiko; Kawasaki, Naohito

    2016-01-01

    In this study, we prepared virgin (S, L) and calcined (S-380, S-1000, L-380, L-1000) magnesium hydroxide for regeneration of waste edible oil. Deterioration of soybean oil, rapeseed oil, and olive oil was achieved by heat and aeration treatment. The properties of the different adsorbents were investigated using specific surface area measurements, scanning electron microscopy, X-ray diffraction analysis, thermogravimetric-differential thermal analysis, and surface pH measurement. Moreover, the relationship between the changes in acid value (AV) and carbonyl value (CV) and the adsorbent properties were evaluated. The specific surface areas of S-380 and L-380 were greater than that of other adsorbents. In addition, the XRD results show that S-380 and L-380 contain both magnesium hydroxide and magnesium oxide structures. The decreases in AV and CV using S-380 and L-380 were greater than achieved using other adsorbents. The correlation coefficients between the decrease in AV and CV and specific surface area were 0.947 for soybean oil, 0.649 for rapeseed oil, and 0.773 for olive oil, respectively. The results obtained in this study suggest that a physical property of the adsorbent, namely specific surface area, was primarily responsible for the observed decreases in AV and CV. Overall, the results suggest that S-380 and L-380 are useful for the regeneration of waste edible oil.

  2. Recovery of plutonium from nitric acid containing oxalate and fluoride by a macroporous bifunctional phosphinic acid resin (MPBPA)

    International Nuclear Information System (INIS)

    Venugopal Chetty, K.; Godbole, A.G.; Swarup, R.; Vaidya, V.N.; Venugopal, V.; Vasudeva Rao, P.R.

    2006-01-01

    The sorption of Pu from nitric acid solutions containing oxalate/fluoride was studied using an indigenously available macroporous bifunctional phosphinic acid (MPBPA) resin. Batch experiments were carried out to obtain the distribution data of Pu(IV) with a view to optimize conditions for its recovery from nitric acid waste solutions containing oxalate or fluoride ions. The measurements showed high distribution ratio (D) values even in the presence of strong complexing ions, like oxalate and fluoride, indicating the possibility of recovery of Pu from these types of waste solution. Column studies were carried out using this resin to recover Pu from the oxalate supernatant waste solution, which showed that up to 99% of Pu could be adsorbed on the resin. Elution of Pu loaded on the resin was studied using different eluting agents. (author)

  3. Embedding in thermosetting resins

    International Nuclear Information System (INIS)

    Buzonniere, A. de

    1985-01-01

    Medium activity waste coming either from nuclear power plants in operation such as evaporator concentrates, spent resins, filter cartridges or the dismantling of installations are embedded in order to obtain a product suitable for long term disposal. Embedding in thermosetting resins (polyester or epoxy) is one among currently used techniques; it is being developed by the CEA (Commissariat a l'Energie Atomique) and Technicatome (subsidiary of CEA and EDF). The process is easy to operate and yields excellent results particularly as far as volume reduction and radioelement containment (cesium particularly) are concerned. The process has already been in operation in four stationary plants for several years. Extension of the process to mobile units has been completed by Technicatome in collaboration with the CEA [fr

  4. Preparation of novel polyamine-type chelating resin with hyperbranched structures and its adsorption performance

    Science.gov (United States)

    Zhao, Wei; Wang, Huan; Li, Yuhong; Li, Chenxi

    2018-01-01

    This paper explored the method of combining atom transfer radical polymerization (ATRP) technology and hyperbranched polymer principle to prepare the high capacity chelating resin. First, surface-initiated atom transfer radical polymerization (SI-ATRP) method was used to graft glycidyl methacrylate (GMA) on chloromethylated cross-linked styrene-divinylbenzene resin, and then the novel polyamine chelating resin with a kind of hyperbranched structure was prepared through the amination reaction between amino group of (2-aminoethyl) triamine and epoxy group in GMA. This resin had a selective effect on As(V) and Cr(VI) at a relatively low pH and can be used for the disposal of waste water containing As(V) and Cr(VI). It had a relatively strong adsorption effect on Cu(II), Pb(II), Cd(II) and Cr(III) and can be used for the disposal of heavy metal ion waste water. The finding was that, the adsorption capacity of resin on the studied heavy metal ions was higher than that of the chelating resin synthesized by traditional technology and also higher than that of the resin modified by ATRP technology and bifunctional chelator, indicating that the combination of ATRP and hyperbranched polymer concept is an effective method to prepare chelating resin with high capacity. PMID:29515875

  5. Preparation of novel polyamine-type chelating resin with hyperbranched structures and its adsorption performance

    Science.gov (United States)

    Chen, Youning; Zhao, Wei; Wang, Huan; Li, Yuhong; Li, Chenxi

    2018-02-01

    This paper explored the method of combining atom transfer radical polymerization (ATRP) technology and hyperbranched polymer principle to prepare the high capacity chelating resin. First, surface-initiated atom transfer radical polymerization (SI-ATRP) method was used to graft glycidyl methacrylate (GMA) on chloromethylated cross-linked styrene-divinylbenzene resin, and then the novel polyamine chelating resin with a kind of hyperbranched structure was prepared through the amination reaction between amino group of (2-aminoethyl) triamine and epoxy group in GMA. This resin had a selective effect on As(V) and Cr(VI) at a relatively low pH and can be used for the disposal of waste water containing As(V) and Cr(VI). It had a relatively strong adsorption effect on Cu(II), Pb(II), Cd(II) and Cr(III) and can be used for the disposal of heavy metal ion waste water. The finding was that, the adsorption capacity of resin on the studied heavy metal ions was higher than that of the chelating resin synthesized by traditional technology and also higher than that of the resin modified by ATRP technology and bifunctional chelator, indicating that the combination of ATRP and hyperbranched polymer concept is an effective method to prepare chelating resin with high capacity.

  6. Treatment of radioactive waste - Routine or challenge? Proceedings

    International Nuclear Information System (INIS)

    2003-01-01

    The seminar had the following topics: Proposal for new legislation covering radioactive waste management in the EU, new requirements preparations for the later repository, efficient and cost effective treatment of radioactive waste water, intermediate level waste cementation, incineration of spent ion exchange resins in a triphasic mixture, application of THOR-technology on resins, new development for transportation and storage of reactor vessel parts, and conditioning of nuclear fuel containing wastes. (uke)

  7. Leaching of nuclear power reactor wastes forms

    International Nuclear Information System (INIS)

    Endo, L.S.; Villalobos, J.P.; Miyamoto, H.

    1986-01-01

    The leaching tests for power reactor wastes carried out at IPEN/CNEN-SP are described. These waste forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. 3 years leaching results are reported, determining cesium and strontium diffusivity coefficients for boric acid waste form and ion-exchange resins. (Author) [pt

  8. Radiation effects on resins and zeolites at Three Mile Island Unit II

    International Nuclear Information System (INIS)

    Reilly, J.K.; Grant, P.J.; Hofstetter, K.J.; Quinn, G.J.; Runion, T.C.

    1984-01-01

    The March 1979 accident at Three Mile Island created certain waste forms that are not routinely encountered in normal light water reactor plants and were the subject of a research and development program. Fifty EPICOR II organic resin containers were loaded to as much as 2200 Ci (of predominantly cesium plus strontium) in the course of processing contaminated water. Resin samples were taken from two of the EPICOR II vessels and examined. Nineteen zeolite containers loaded with radioactive cesium plus strontium to as much as 55,000 Ci have been analyzed. The makeup and purification system demineralizer resins, highly contaminated by the letdown of reactor coolant, have been radiation surveyed, and gas, liquid, and resin samples have been obtained. On-site sampling and analysis of these wastes indicated that combustible gases (hydrogen and oxygen) were being generated as a result of radiolysis. The results of this work are discussed, compared, and presented

  9. Studies on the incorporation of spent ion exchange resins from nuclear power plants into bitumen and cement

    International Nuclear Information System (INIS)

    Bonnevie-Svendsen, M.; Tallberg, K.; Aittola, P.; Tollbaeck, H.

    1976-01-01

    The joint Nordic incorporation experiments should provide technical data needed for the assessment of solidification techniques for wastes from nuclear reactors in the Nordic countries. Spent ion exchange resins are a main fraction of such wastes, and more knowledge about their incorporation is wanted. The effects of simulated and real ion exchange wastes on the quality of bitumen and cement incorporation products were studied. Blown and distilled bitumen and three Portland cement qualities were used. Product characterizations were based on properties relevant for safe waste management, storage, transport and disposal. The applicability and relevance of established and suggested tests is discussed. Up to 40-60% dry resin could be incorporated into bitumen without impairing product qualities. Products with higher resin contents were found to swell in contact with water. The products had a high leach resistance. Their form stability was improved by incorporated resins. Product qualities appeared to be less affected by physico-chemical variables than by mechanical process parameters. Pure resin-cement products tend to decompose in water. Product qualities were strongly affected by a variety of physico-chemical process parameters, and integer products were only obtained within narrow tolerance limits. Caesium was rapidly leached out. To attain integer products and improved leach resistance within technically acceptable tolerance limits it was necessary to utilize stabilizing and caesium-retaining additives such as Silix and vermiculite. Under the present conditions the water content of the resins limited the amounts that could be incorporated in 40-50wt% or about 70vol.% water-saturated (containing 20-40% dry) resin. (author)

  10. Phosphorus recovery as struvite from eutropic waters by XDA-7 resin.

    Science.gov (United States)

    Li, Huanwen; Ye, Zhiping; Lin, Ying; Wang, Fengying

    2012-01-01

    Phosphorus releases into aquatic environment and its subsequent contribution to eutrophication have resulted in a widespread global pollution issue. However, phosphorus is a non-renewable source. The potential supplies of phosphorus are decreasing worldwide. Therefore, removal and recovery of phosphorus from the eutropic waters is important, emergent and necessary. In this research, experiments for recovering phosphate from eutropic waters by anion exchange combined with struvite precipitation were conducted. The results indicated that the prepared XDA-7 resin was an effective adsorbent for phosphate. The adsorption isotherm of XDA-7 resin was found to be a modified Freundlich type. The maximum phosphate adsorption (20.9 mg/g) occurred in the pH range of 6.0-8.0. Phosphate adsorbed on the XDA-7 resin was effectively desorbed with 8% NaCl solution, and the resin was able to be regenerated with 3% NaClO and 4% NaOH solutions. Phosphate desorbed from the resin was recovered as magnesium ammonium phosphate (struvite). The obtained struvite was analyzed by acid dissolution method, scanning electron microscope (SEM) and Fourier transform infrared spectroscopy (FTIR). The struvite precipitate was found to be 75.8% in purity, a high-value fertilizer.

  11. Development of an ion-exchange process for removing cesium from high-level radioactive liquid wastes

    International Nuclear Information System (INIS)

    Baumgarten, P.K.; Wallace, R.M.; Whitehurst, D.A.; Steed, J.M.

    1979-11-01

    Methods to determine resin characteristics, i.e., cesium equilibria and diffusion rates, were developed. These parameters can now guide resin selection and aid in interpreting column performance. The K/sub D/ cesium ion concentration relation gives evidence of three different types of ion exchange sites. The countercurrent load/elution/regeneration cycle for the removal of cesium by ion exchange repeatedly reached the goal decontamination factor (DF) of 10,000 at throughputs up to 60 column volumes. Resin backwashing appears feasible, but further development of column geometry will be required. The proposed ammonium elutriant is satisfactory. Regeneration end-point can be controlled by electrical conductivity monitoring

  12. Pyrolysis and oxidative pyrolysis experiments with organization exchange resin

    International Nuclear Information System (INIS)

    Chun, Ung Kyung

    1997-01-01

    Pyrolysis may be an important pretreatment step before vitrification in a cold crucible melter (CCM). During vitrification of organic resin the carbon or other remaining residues may harm the performance of the cold crucible melter of the eventual stability of the final glass product. Hence, it is important to reduce or prevent such harmful waste from entry into the cold crucible melter. Pretreatment with pyrolysis will generally provide volume reduction resulting in less amount of solid waste that needs to be handled by the CCM; in addition, the pyrolytic processes may breakdown much of the complex organics causing release through volatilization resulting in less carbon and other harmful substances. Hence, KEPRI has undertaken studies on the pyrolysis and oxidative pyrolysis of organic ion exchange resin. Pyrolysis and oxidative pyrolysis were examined with TGA and a tube furnace. TGA results for pyrolysis with the flow of nitrogen indicate that even after pyrolyzing from room temperature to about 900 deg C, a significant mass fraction of the original cationic resin remains, approximately 46 %. The anionic resin when pyrolytically heated in a flow of nitrogen only, from room temperature to about 900 deg C, produced a final residue mass fraction of about 8 percent. Oxidation at a ratio of air to nitrogen, 1:2, reduced the cationic resin to 5.3% when heated at 5 C/min. Oxidation of anionic resin at the same ratio and same heating rate left almost no solid residue. Pyrolysis (e.g. nitrogen-only environment) in the tube furnace of larger samples relative to the TGA produced very similar results to the TGA. The differences may be attributed to the scale effects such as surface area exposure to the gas stream, temperature distributions throughout the resin, etc. (author) 7 refs., 7 figs

  13. Pyrolysis and oxidative pyrolysis experiments with organization exchange resin

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Ung Kyung [Korea Electric Power Research Insititute, Taejon (Korea, Republic of)

    1997-12-31

    Pyrolysis may be an important pretreatment step before vitrification in a cold crucible melter (CCM). During vitrification of organic resin the carbon or other remaining residues may harm the performance of the cold crucible melter of the eventual stability of the final glass product. Hence, it is important to reduce or prevent such harmful waste from entry into the cold crucible melter. Pretreatment with pyrolysis will generally provide volume reduction resulting in less amount of solid waste that needs to be handled by the CCM; in addition, the pyrolytic processes may breakdown much of the complex organics causing release through volatilization resulting in less carbon and other harmful substances. Hence, KEPRI has undertaken studies on the pyrolysis and oxidative pyrolysis of organic ion exchange resin. Pyrolysis and oxidative pyrolysis were examined with TGA and a tube furnace. TGA results for pyrolysis with the flow of nitrogen indicate that even after pyrolyzing from room temperature to about 900 deg C, a significant mass fraction of the original cationic resin remains, approximately 46 %. The anionic resin when pyrolytically heated in a flow of nitrogen only, from room temperature to about 900 deg C, produced a final residue mass fraction of about 8 percent. Oxidation at a ratio of air to nitrogen, 1:2, reduced the cationic resin to 5.3% when heated at 5 C/min. Oxidation of anionic resin at the same ratio and same heating rate left almost no solid residue. Pyrolysis (e.g. nitrogen-only environment) in the tube furnace of larger samples relative to the TGA produced very similar results to the TGA. The differences may be attributed to the scale effects such as surface area exposure to the gas stream, temperature distributions throughout the resin, etc. (author) 7 refs., 7 figs.

  14. Alternative solidification techniques for radioactive ion exchange resins and liquid concentrates

    International Nuclear Information System (INIS)

    Thegerstroem, C.

    1980-01-01

    Methods, that are used or are under development for solidification of radioactive ion exchange resins or liquid concentrates, utilize normally cement, bitumen or some polymere as matrix material. This report contains a review and a description of these solidification processes and their products, especially of relatively new techniques that are under development in different countries. It is possible that solidification in thermosetting resins will be more used in the future, especially when product quality requirements are high (for instance when solidifying medium level resins) or when special waste categories has to be solidified. However it is not probable that thermosetting resins will be extensively used in a broad application as matrix material. In that case the methods are to complicated and expensive compared to, for instance, solidification in concrete. Systems for incorporation in polyesteremulsions (Dow-process) have a potential as they are quite simple and can accept a large variation of liquid wastes. Some methods in an early stage of development (for instance Inert Carrier Radwaste Process) will have to be tested in active application before they can be further evaluated. (author)

  15. Electromagnetic mixed-waste processing system for asbestos decontamination

    International Nuclear Information System (INIS)

    1995-04-01

    The first phase of a program to develop and demonstrate a cost-effective, integrated process for remediation of asbestos-containing material that is contaminated with organics, heavy metals, and radioactive compounds was successfully completed. Laboratory scale tests were performed to demonstrate initial process viability for asbestos conversion, organics removal, and radionuclide and heavy metal removal. All success criteria for the laboratory tests were met. (1) Ohio DSI demonstrated greater than 99% asbestos conversion to amorphous solids using their commercial process. (2) KAI demonstrated 90% removal of organics from the asbestos suspension. (3) Westinghouse STC achieved the required metals removal criteria on a laboratory scale (e.g., 92% removal of uranium from solution, resin loadings of 0.6 equivalents per liter, and greater than 50% regeneration of resin in a batch test.) Using the information gained in the laboratory tests, the process was reconfigured to provide the basis for the mixed waste remediation system. An integrated process is conceptually developed, and a Phase 2 program plan is proposed to provide the bench-scale development needed in order to refine the design basis for a pilot processing system

  16. Vitrification of cesium-contaminated organic ion exchange resin

    International Nuclear Information System (INIS)

    Sargent, T.N. Jr.

    1994-08-01

    Vitrification has been declared by the Environmental Protection Agency (USEPA) as the Best Demonstrated Available Technology (BDAT) for the permanent disposal of high-level radioactive waste. Savannah River Site currently uses a sodium tetraphenylborate (NaTPB) precipitation process to remove Cs-137 from a wastewater solution created from the processing of nuclear fuel. This process has several disadvantages such as the formation of a benzene waste stream. It has been proposed to replace the precipitation process with an ion exchange process using a new resorcinol-formaldehyde resin developed by Savannah River Technical Center (SRTC). Preliminary tests, however, showed that problems such as crust formation and a reduced final glass wasteform exist when the resin is placed in the melter environment. The newly developed stirred melter could be capable of overcoming these problems. This research explored the operational feasibility of using the stirred tank melter to vitrify an organic ion exchange resin. Preliminary tests included crucible studies to determine the reducing potential of the resin and the extent of oxygen consuming reactions and oxygen transfer tests to approximate the extent of oxygen transfer into the molten glass using an impeller and a combination of the impeller and an external oxygen transfer system. These preliminary studies were used as a basis for the final test which was using the stirred tank melter to vitrify nonradioactive cesium loaded organic ion exchange resin. Results from this test included a cesium mass balance, a characterization of the semi-volatile organic compounds present in the off gas as products of incomplete combustion (PIC), a qualitative analysis of other volatile metals, and observations relating to the effect the resin had on the final redox state of the glass

  17. Treatment of radioactive laboratory waste for mercury removal

    International Nuclear Information System (INIS)

    Osteen, A.B.; Bibler, J.P.

    1990-01-01

    Routine analyses of Savannah River Laboratory wastes at the Savannah River Site occasionally reveal mercury concentrations in the waste in excess of the 0.200 μg/L RCRA limit. An ion exchange resin has been demonstrated to be effective for the removal of dissolved mercury from laboratory waste in a special permitted project. The ion exchange material is Duolite trademark GT-73, a polystyrene/divinylbenzene resin with thiol functional groups. As a result of the decontamination demonstration, the resin is in use or under consideration for use with several other SRS radwaste streams as a reliable medium for mercury removal

  18. Effective regeneration of anode material recycled from scrapped Li-ion batteries

    Science.gov (United States)

    Zhang, Jin; Li, Xuelei; Song, Dawei; Miao, Yanli; Song, Jishun; Zhang, Lianqi

    2018-06-01

    Recycling high-valuable metal elements (such as Li, Ni, Co, Al and Cu elements) from scrapped lithium ion batteries can bring significant economic benefits. However, recycling and reusing anode material has not yet attracted wide attention up to now, due to the lower added-value than the above valuable metal materials and the difficulties in regenerating process. In this paper, a novel regeneration process with significant green advance is proposed to regenerate anode material recycled from scrapped Li-ion batteries for the first time. After regenerated, most acetylene black (AB) and all the styrene butadiene rubber (SBR), carboxymethylcellulose sodium (CMC) in recycled anode material are removed, and the surface of anode material is coated with pyrolytic carbon from phenolic resin again. Finally, the regenerated anode material (graphite with coating layer, residual AB and a little CMC pyrolysis product) is obtained. As expected, all the technical indexs of regenerated anode material exceed that of a midrange graphite with the same type, and partial technical indexs are even closed to that of the unused graphite. The results indicate the effective regeneration of anode material recycled from scrapped Li-ion batteries is really achieved.

  19. Synthesis of hemicellulose-acrylic acid graft copolymer super water absorbent resin by ultrasonic irradiation technology

    Directory of Open Access Journals (Sweden)

    Fangfang LIU

    2015-12-01

    Full Text Available The hemicellulose super water absorbent resin is prepared by using ultrasonic irradiation technology, with the waste liquid produced during the preparation of viscose fiber which contains a large amount of hemicellulose as raw material, acrylic acid as graft monomer, N,N’-methylene bis acrylamide (NMBA as cross linking agent, and (NH42S2O8-NaHSO3 as the redox initiation system. The synthesis conditions, structure and water absorption ability of resin are discussed. The results indicate that water absorbency of the resin is 311 g/g, the tap water absorbency is 102 g/g, the normal saline absorbency is 55 g/g, and the artificial urine absorbency is 31 g/g under the optimal synthesis conditions, so the resin has great water absorption rate and water retaining capacity. The FT-IR and SEM analysis shows that the resin with honeycomb network structure is prepared. The successfully synthesized of the resin means that the hemicellulose waste liquid can be highly effectively recycled, and it provides a kind of new raw material for the synthesis of super water absorbent resin.

  20. Testing waste forms containing high radionuclide loadings

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  1. Comparison of solvent extraction and extraction chromatography resin techniques for uranium isotopic characterization in high-level radioactive waste and barrier materials.

    Science.gov (United States)

    Hurtado-Bermúdez, Santiago; Villa-Alfageme, María; Mas, José Luis; Alba, María Dolores

    2018-07-01

    The development of Deep Geological Repositories (DGP) to the storage of high-level radioactive waste (HLRW) is mainly focused in systems of multiple barriers based on the use of clays, and particularly bentonites, as natural and engineered barriers in nuclear waste isolation due to their remarkable properties. Due to the fact that uranium is the major component of HLRW, it is required to go in depth in the analysis of the chemistry of the reaction of this element within bentonites. The determination of uranium under the conditions of HLRW, including the analysis of silicate matrices before and after the uranium-bentonite reaction, was investigated. The performances of a state-of-the-art and widespread radiochemical method based on chromatographic UTEVA resins, and a well-known and traditional method based on solvent extraction with tri-n-butyl phosphate (TBP), for the analysis of uranium and thorium isotopes in solid matrices with high concentrations of uranium were analysed in detail. In the development of this comparison, both radiochemical approaches have an overall excellent performance in order to analyse uranium concentration in HLRW samples. However, due to the high uranium concentration in the samples, the chromatographic resin is not able to avoid completely the uranium contamination in the thorium fraction. Copyright © 2018 Elsevier Ltd. All rights reserved.

  2. Feasibility study of solidification for low-level liquid waste generated by sulfuric acid elution treatment of spent ion exchange resin

    International Nuclear Information System (INIS)

    Asano, Takashi; Kawasaki, Tooru; Higuchi, Natsuko; Horikawa, Yoshihiko

    2007-01-01

    Low-level liquid waste with relatively high levels of radioactivity is generated by the sulfuric acid elution treatment of spent ion exchange resin used in water purification systems of nuclear power plants. We studied cement-like solidification process for this type waste that contains a high concentration of sodium sulfate. For this type waste, it is important that the sulfate ion should not dissolve from the solid waste because it forms ettringite on reaction with minerals in the concrete, and this leads to cracking during repository storage. It is also preferable that the pH of pore water of the solid waste be low, because the bentonite of the repository changes in quality on exposure to alkaline solution. Our solidification process has two procedures: conversion into insoluble sulfate from sodium sulfate (CIS) and formation of low pH cement-like solid (FLS). In the CIS procedure, BaSO 4 precipitation occurs with addition of Ba(OH) 2 ·8H 2 O to the liquid waste when the Ba/SO 4 molar ratio > 1. In the FLS procedure, silica fume and blast furnace slag are added to the liquid wastes containing Ba S O 4 precipitate. The CIS reaction yield is over 98% and the pH of pore water of the solid waste is 11.5 or less. Therefore, we think that our solidification process is one of the best methods for treating liquid waste that contains a high concentration of sodium sulfate. (author)

  3. Hydraulic Permeability of Resorcinol-Formaldehyde Ion-Exchange Resin - Effects of Oxygen Uptake and Radiation

    International Nuclear Information System (INIS)

    Taylor, Paul Allen

    2009-01-01

    An ion-exchange process, using spherical resorcinol-formaldehyde (RF) resin is the baseline process for removing cesium from the dissolved salt solution in the high-level waste tanks at the U.S. Department of Energy's (DOE) Hanford site in Washington State. The RF resin is also being evaluated for use in the proposed Small Column Ion Exchange (SCIX) system, which is an alternative treatment option at DOE's Savannah River Site (SRS)in South Carolina. Testing at ORNL will determine the impact of radiation exposure and oxygen uptake by the RF resin on the hydraulic permeability of the resin. Samples of the resin will be removed periodically to measure physical properties (bead size and compressibility) and cesium capacity. The proposed full-scale treatment system at Hanford, the Waste Treatment Plant (WTP), will use an ion-exchange column containing nominally 680 gal of resin, which will treat 30 gpm of waste solution. The ion-exchange column is designed for a typical pressure drop of 6 psig, with a maximum of 9.7 psig. The lab-scale column is 3-in. clear PVC pipe and is prototypic of the proposed Hanford column. The fluid velocity in the lab-scale test will be much higher than for the full-scale column, in order to generate the maximum pressure drop expected in that column (9.7 psig). The frictional drag from this high velocity will produce similar forces on the resin in the lab-scale column as would be expected at the bottom of the full-scale column. The chemical changes in the resin caused by radiation exposure and oxygen uptake are expected to cause physical changes in the resin that could reduce the bed porosity and reduce the hydraulic permeability of the resin bed. These changes will be monitored by measuring the pressure drop through the lab-scale column and by measuring the physical properties of samples of the resin. The test loop with the lab-scale column is currently being fabricated, and operation will start by late May. Testing will be completed by the

  4. Simultaneous separation and detection of actinides in acidic solutions using an extractive scintillating resin.

    Science.gov (United States)

    Roane, J E; DeVol, T A

    2002-11-01

    An extractive scintillating resin was evaluated for the simultaneous separation and detection of actinides in acidic solutions. The transuranic extractive scintillating (TRU-ES) resin is composed of an inert macroporous polystyrene core impregnated with organic fluors (diphenyloxazole and 1,4-bis-(4-methyl-5-phenyl-2-oxazolyl)benzene) and an extractant (octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide in tributyl phosphate). The TRU-ES resin was packed into FEP Teflon tubing to produce a flow cell (0.2-mL free column volume), which is placed into a scintillation detection system to obtain pulse height spectra and time series data during loading and elution of actinides onto/from the resin. The alpha-particle absolute detection efficiencies ranged from 77% to 96.5%, depending on the alpha energy and quench. In addition to the on-line analyses, off-line analyses of the effluent can be conducted using conventional detection methods. The TRU-ES resin was applied to the quantification of a mixed radionuclide solution and two actual waste samples. The on-line characterization of the mixed radionuclide solution was within 10% of the reported activities whereas the agreement with the waste samples was not as good due to sorption onto the sample container walls and the oxidation state of plutonium. Agreement between the on-line and off-line analyses was within 35% of one another for both waste samples.

  5. Studies of Vitrification of Ion-Exchange Resins. A Joint USA-Argentina Collaborative Work

    International Nuclear Information System (INIS)

    Hutson, N.D.; Herman, C.A.; Zamecnik, J.R.; Sundaram, S.K.; Perez, J.M.; Hoeffner, S.L.; Russo, D.O.; Sterba, M.

    2003-01-01

    Under the Science and Technology Implementing Arrangement for Cooperation on Radioactive and Mixed Waste Management (JCCRM), the U.S.Department of Energy (DOE) is helping to transfer waste treatment technology to international atomic energy commissions.As part of the JCCRM, DOE has established a collaborative research agreement with the Comision Nacional de Energia Atomica (Cnea).The Cnea is investigating treatment and disposal options for organic ion exchange resins currently stored at two nuclear power plants in the Republic of Argentina.The major hazards of the ion exchange resins are their organic composition and the contaminants that are present on the resins after purification processes.The principal contaminants are usually the radioactive species that are removed.For these studies, actual non-radioactive resins from Argentina's Embalse and Atucha plants were tested.The glass produced during the runs was durable was measured by the Product Consistency Test (PCT).The product had a predictable, mostly amorphous composition throughout the demonstrations; though there was some evidence of the formation of clinopyroxene crystals.The immobilized product represented an approximately 70% volume reduction from the simulated Argentine ion exchange resin (i.e., a reduction from the volume of as-stored wet resin to the volume of the ultimate borosilicate glass product).For all runs, the radioactive surrogate retention was near 100%

  6. Sequestration Resins for Accelerating Removal of Radioactive Contaminants

    International Nuclear Information System (INIS)

    Frattini, Paul-L.; Wells, Daniel-M.; Garcia, Susan-E.; Richard, Kohlmann; Asay, Roger; Yengoyan, Leon

    2012-09-01

    The Electric Power Research Institute (EPRI) is developing sequestration resins that can be used in the treatment of nuclear plant water streams for the enhanced removal of ionic cobalt. EPRI is focusing on three key areas of success: 1. Plant safety. The resins that are synthesized must be fully tested to determine that no leachable species or decomposition products (in the event of a resin bed failure) would be introduced to the plant. 2. Acceptable system performance. The resins are currently being synthesized in a powdered form for use in the reactor water clean-up and fuel pool clean-up systems that utilize pre-coatable filter elements. The resins must have effective flocking behavior; uniform application over the underlay resin and efficient removal from the septa elements after use. Bead type resins are also under development. 3. Enhanced cobalt removal. The resins are expected to out-perform the currently used ion exchange resins in the removal of ionic cobalt. During nuclear plant maintenance or refueling outages, current ion exchange resins may require several days to reduce concentrations of cobalt (for example, radio-cobalt 60 Co and 58 Co) and other activated corrosion products to safe levels in reactor coolant streams. This performance limitation often delays key maintenance activities. EPRI's resins are expected to provide at least a three-fold increase in removal capacity in light water reactor coolants. These resins also offer the potential for higher overall removal efficiencies reducing occupational exposures and waste management costs. This paper addresses issues from the range of novel resin development for radio-cobalt removal from synthesis at the bench-top level through scale-up to demonstration of use in an actual operating nuclear power plant. (authors)

  7. Accurate determination of 129I concentrations and 129I/137Cs ratios in spent nuclear resins by Accelerator Mass Spectrometry

    International Nuclear Information System (INIS)

    Nottoli, Emmanuelle; Bienvenu, Philippe; Labet, Alexandre; Bourlès, Didier; Arnold, Maurice; Bertaux, Maité

    2014-01-01

    Determining long-lived radionuclide concentrations in radioactive waste has fundamental implications for the long-term management of storage sites. This paper focuses on the measurement of low 129 I contents in ion exchange resins used for primary fluid purification in Pressurised Water Reactors (PWR). Iodine-129 concentrations were successfully determined using Accelerator Mass Spectrometry (AMS) following a chemical procedure which included (1) acid digestion of resin samples in HNO 3 /HClO 4 , (2) radioactive decontamination by selective iodine extraction using a new chromatographic resin (CL Resin), and (3) AgI precipitation. Measured 129 I concentrations ranged from 4 to 12 ng/g, i.e. from 0.03 to 0.08 Bq/g. The calculation of 129 I/ 137 Cs activity ratios used for routine waste management produced values in agreement with the few available data for PWR resin samples. - Highlights: • In the context of radioactive waste management, this study aimed at measuring 129 I in spent resins using accelerator mass spectrometry. • The treatment procedure included microwave acid digestion of samples, iodine extraction by CL resins and AgI precipitation. • Developed first on synthetic matrices, the chemical treatment procedure was then successfully applied to real resin samples. • 129 I concentrations ranged from 4 to 12 ng/g of dry resin. • Results are in agreement with previous measurements and support reference values currently used for nuclear resin management

  8. Upgrade to Ion Exchange Modeling for Removal of Technetium from Hanford Waste Using SuperLig® 639 Resin

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L. [Savannah River Site (SRS), Aiken, SC (United States); Smith, F. [Savannah River Site (SRS), Aiken, SC (United States); Aleman, S. [Savannah River Site (SRS), Aiken, SC (United States); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States)

    2013-05-16

    This report documents the development and application of computer models to describe the sorption of pertechnetate [TcO₄⁻], and its surrogate perrhenate [ReO₄⁻], on SuperLig® 639 resin. Two models have been developed: 1) A thermodynamic isotherm model, based on experimental data, that predicts [TcO₄⁻] and [ReO₄⁻] sorption as a function of solution composition and temperature and 2) A column model that uses the isotherm calculated by the first model to simulate the performance of a full-scale sorption process. The isotherm model provides a synthesis of experimental data collected from many different sources to give a best estimate prediction of the behavior of the pertechnetate-SuperLig® 639 system and an estimate of the uncertainty in this prediction. The column model provides a prediction of the expected performance of the plant process by determining the volume of waste solution that can be processed based on process design parameters such as column size, flow rate and resin physical properties.

  9. Recycling the construction and demolition waste to produce polymer concrete

    Science.gov (United States)

    Hamza, Mohammad T.; Hameed, Awham M., Dr.

    2018-05-01

    The sustainable management for solid wastes of the construction and demolition waste stimulates searching for safety applications for these wastes. The aim of this research is recycling of construction and demolition waste with some different types of polymeric resins to be used in manufacturing process of polymer mortar or polymer concrete, and studying their mechanical and physical properties, and also Specify how the values of compressive strength and the density are affected via the different parameters. In this research two types of construction and demolition waste were used as aggregates replacement (i.e. waste cement/concrete debris, and the waste blocks) while the two types of polymer resins (i.e. Unsaturated polyester and Epoxy) as cement replacements. The used weight percentages of the resins were changed within (1°, 20, 25 and 30) % to manufacture this polymer concrete.

  10. Study and application on lean resin converting in uranium mill

    International Nuclear Information System (INIS)

    Zhao Shaoxi; Huang Qijin; Zhu Shuguang; Yi Faqing; Du Wenjie

    2012-01-01

    The field test about sulphuric acid used to convert lean resin was finished. The results indicated sulphuric acid could replace chlorin in lean resin and could be reclaimed to desorption procedure. The consumption of NaCl decreased, the chlorin concentration of tailing decreased too. Both of uranium loss and waste water volume were reduced. The uranium concentration of tailing decreased and energy saving and emission reduction can be achieved. (authors)

  11. Technical Task and Quality Assurance Plan in Support of BNFL Part B: Studies of Ion Exchange Resin Integrity under Flowsheet Extremes: Part II

    International Nuclear Information System (INIS)

    Nash, C.A.

    2000-01-01

    This task will address four items related to ion exchange stability: (1) process upset evaluation of resin in contact with 1 molar sodium permanganate at 25 and 40 degrees C, (2) accelerated aging with nitric acid solution used during normal regeneration operations, (3) prolonged contacting of SuperLig 644 resin with 5 molar nitric acid at room temperature, and (4) prolonged contacting of SuperLig 644 resin with deionized water at 60 plus/minus 5 degrees C

  12. Degradation of resins in EPICOR-II prefilters from Three Mile Island

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Sanders, R.D. Sr.

    1986-01-01

    The Low-Level Waste Data Base Development--EPICOR-II Resin/Liner Investigation Program funded by the U.S. Nuclear Regulatory Commission is investigating the chemical and physical conditions of the synthetic ion exchange resins contained in several EPICOR-II prefilters. Those prefilters were used during cleanup of contaminated water from the Three Mile Island Nuclear Power Station after the March 1979 accident. This paper summarizes results and analyses of the second sampling of resins from prefilters PF-8 and -20. Results are compared with baseline data from tests performed on unirradiated resins supplied by Epicor, Inc. to determine if degradation has occurred due to the high internal radiation dose. Results also are compared with results from tests performed on resins obtained from the first sampling of those two prefilters

  13. Method of continuously regenerating decontaminating electrolytic solution

    International Nuclear Information System (INIS)

    Sasaki, Takashi; Kobayashi, Toshio; Wada, Koichi.

    1985-01-01

    Purpose: To continuously recover radioactive metal ions from the electrolytic solution used for the electrolytic decontamination of radioactive equipment and increased with the radioactive dose, as well as regenerate the electrolytic solution to a high concentration acid. Method: A liquid in an auxiliary tank is recycled to a cathode chamber containing water of an electro depositing regeneration tank to render pH = 2 by way of a pH controller and a pH electrode. The electrolytic solution in an electrolytic decontaminating tank is introduced by way of an injection pump to an auxiliary tank and, interlocking therewith, a regenerating solution is introduced from a regenerating solution extracting pump by way of a extraction pipeway to an electrolytic decontaminating tank. Meanwhile, electric current is supplied to the electrode to deposit radioactive metal ions dissolved in the cathode chamber on the capturing electrode. While on the other hand, anions are transferred by way of a partition wall to an anode chamber to regenerate the electrolytic solution to high concentration acid solution. While on the other hand, water is supplied by way of an electromagnetic valve interlocking with the level meter to maintain the level meter constant. This can decrease the generation of the liquid wastes and also reduce the amount of the radioactive secondary wastes. (Horiuchi, T.)

  14. Method of processing radioactive wastes

    International Nuclear Information System (INIS)

    Nomura, Ichiro; Hashimoto, Yasuo.

    1984-01-01

    Purpose: To improve the volume-reduction effect, as well as enable simultaneous procession for the wastes such as burnable solid wastes, resin wastes or sludges, and further convert the processed materials into glass-solidified products which are much less burnable and stable chemically and thermally. Method: Auxiliaries mainly composed of SiO 2 such as clays, and wastes such as burnable solid wastes, waste resins and sludges are charged through a waste hopper into an incinerating melting furnace comprising an incinerating and a melting furnace, while radioactive concentrated liquid wastes are sprayed from a spray nozzle. The wastes are burnt by the heat from the melting furnace and combustion air, and the sprayed concentrated wastes are dried by the hot air after the combustion into solid components. The solid matters from the concentrated liquid wastes and the incinerating ashes of the wastes are melted together with the auxiliaries in the melting furnace and converted into glass-like matters. The glass-like matters thus formed are caused to flow into a vessel and gradually cooled to solidify. (Horiuchi, T.)

  15. Nitrate and Perchlorate removal from groundwater by ion exchange; TOPICAL

    International Nuclear Information System (INIS)

    Burge, S; Halden, R

    1999-01-01

    This study was conducted to evaluate the performance of a small scale ion exchange unit (Krudico, Inc of Auborn, IA) for removal of nitrate and perchlorate from groundwater at Lawrence Livermore National Laboratory's Site 300. The unit was able to treat 3,600 gallons of Site 300 groundwater, at an average influent concentration of 100 mg/L NO(sub 3)(sup -) before breakthrough occurred. The unit contained 2.5 ft(sup 3) of Sybron SR-7 resin. Seventy gallons of regeneration waste were generated (water treated to waste ratio of 51:1). The effluent concentration was about 20 mg/L NO(sub 3)(sup -), which is equivalent to a treatment efficiency of at least 80%. There are several options for implementing this technology at Site 300. A target well, in the 817 area, has been selected. It has a 3 to 4 gpm flow rate, and concentrations of 90 mg/L NO(sub 3)(sup -) and 40(micro)g/L perchlorate. The different treatment options include ion exchange treatment of nitrate only, nitrate and perchlorate, or perchlorate only. Option 1: For the treatment of nitrate only, this unit will be able to treat 3,700 gallons of water before regeneration is required. If both columns of the ion exchange unit are used, 7,400 gallons could be treated before the columns will need to be regenerated (producing 140 gallons of waste, per cycle or every 1.5 days). The effluent nitrate concentration is expected to be about 17 mg/L. Annual operation and maintenance costs are estimated to be$0.14 per gallon of water treated. Option 2: If only perchlorate is to be removed with ion exchange at the 817 area, a smaller unit should be considered. A 55 gallon canister filled with ion exchange resin should be able to reduce perchlorate concentrations in the groundwater from 40(micro)g/L to non-detect levels for three years before the resin would need to be replaced. The contaminant-laden resin would be disposed of as hazardous waste. It is not practical to regenerate the resin because of the extreme difficulty of

  16. Surface Engineering of PAMAM-SDB Chelating Resin with Diglycolamic Acid (DGA) Functional Group for Efficient Sorption of U(VI) and Th(IV) from Aqueous Medium

    Energy Technology Data Exchange (ETDEWEB)

    Ilaiyaraja, P.; Venkatraman, B., E-mail: chemila07@gmail.com [Radiological Safety Division, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Deb, A.K. Singha [Chemical Engineering Division, Bhabha Atomic Research Centre, Mumbai (India); Ponraju, D. [Safety Engineering Division, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Ali, Sk. Musharaf [Chemical Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

    2017-04-15

    Highlights: • A new DGA-PAMAM-SDB chelating resin has been synthesized for actinide sorption. • Maximum sorption capacities of resin are 682 and 544.2 mg g{sup −1}for U(VI) and Th(IV). • DGA-PAMAM-SDB chelating resin could be regenerated and reused. • DFT calculation of actinides interaction with resin corroborates the experimental. • Resin is effective for sorption of actinides from both aqueous and HNO{sub 3} medium. - Abstract: A novel chelating resin obtained via growth of PAMAM dendron on surface of styrene divinyl benzene resin beads, followed by diglycolamic acid functionalization of the dendrimer terminal. Batch experiments were conducted to study the effects of pH, nitric acid concentration, amount of adsorbent, shaking time, initial metal ion concentration and temperature on U(VI) and Th(IV) adsorption efficiency. Diglycolamic acid terminated PAMAM dendrimer functionalized styrene divinylbenzene chelating resin (DGA-PAMAM-SDB) is found to be an efficient candidate for the removal of U(VI) and Th(IV) ions from aqueous (pH > 4) and nitric acid media (> 3 M). The sorption equilibrium could be reached within 60 min, and the experimental data fits with pseudo-second-order model. Langmuir sorption isotherm model correlates well with sorption equilibrium data. The maximum U(VI) and Th(IV) sorption capacity onto DGA-PAMAMG{sub 5}-SDB was estimated to be about 682 and 544.2 mg g{sup −1} respectively at 25 °C. The interaction of actinides and chelating resin is reversible and hence, the resin can be regenerated and reused. DFT calculation on the interaction of U(VI) and Th(IV) ions with chelating resin validates the experimental findings.

  17. Ion Exchange Temperature Testing with SRF Resin - 12088

    Energy Technology Data Exchange (ETDEWEB)

    Russell, R.L.; Rinehart, D.E.; Brown, G.N.; Peterson, R.A. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States)

    2012-07-01

    Ion exchange using the Spherical Resorcinol-Formaldehyde (SRF) resin has been selected by the U.S. Department of Energy's Office of River Protection for use in the Pretreatment Facility of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) and for potential application in an at-tank deployment for removing Cs-137. Recent proposed changes to the WTP ion exchange process baseline indicate that higher temperatures (50 deg. C) to alleviate post-filtration precipitation issues prior to reaching the ion exchange columns may be required. Therefore, it is important to understand the behavior of SRF resin performance under the conditions expected with the new equipment and process changes. This research examined the impact of elevated temperature on resin loading and resin degradation during extended solution flow at elevated temperature (45 deg., 50 deg., 55 deg., 60 deg., 65 deg., 75 deg. C). Testing for extended times at elevated temperatures showed that the resin does degrade and loading capacity is reduced at and above 45 deg. C. Above 60 deg. C the resin appears to not load at all. It was observed that the resin disintegrated at 75 deg. C until not much was left and partially disintegrated at 65 deg. C, which caused the column to plug in both tests after ∼336 hours. The results indicate that WTP will lose resin loading capacity if the ion exchange process is performed above 25 deg. C, and the resin will disintegrate above 65 deg. C. Therefore, WTP will have a restricted operating range of temperatures to perform the ion exchange process with this resin. PNNL and WTP are currently evaluating the operating limits of the resin in further detail. Aging in 0.5 M HNO{sub 3} also caused the resin to lose capacity above 25 deg. C and to completely dissolve at 55 deg. C. Again, WTP will have a restricted operating range of temperatures when eluting the resin with nitric acid in order to maintain resin loading capacity and avoid disintegration of the resin

  18. Research and development for decontamination system of spent resin in Hanbit Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Gi Hong [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2015-12-15

    When reactor coolant leaks occur due to cracks of a steam generator tube, radioactive materials contained in the primary cooling water in nuclear power plant are forced out toward the secondary systems. At this time the secondary water purification resin in the ion exchange resin tower of the steam generator blowdown system is contaminated by the radioactivity of the leaked radioactive materials, so we pack this in special containers and store temporarily because we could not dispose it by ourselves. If steam generator tube leakage occurs, it produces contaminated spent resins annually about 5,000-7,000 liters. This may increase the amount of nuclear waste productions, a disposal working cost and a unit price of generating electricity in the plant. For this reasons, it is required to develop a decontamination process technique for reducing the radioactive level of these resins enough to handle by the self-disposal method. In this research, First, Investigated the structure and properties of the ion exchange resin used in a steam generator blowdown system. Second, Checked for a occurrence status of contaminated spent resin and a disposal technology. Third, identified the chemical characteristics of the waste radionuclides of the spent resin, and examined ionic bonding and separation mechanism of radioactive nuclear species and a spent resin. Finally, we carried out the decontamination experiment using chemicals, ultrasound, microbubbles, supercritical carbon dioxide to process these spent resin. In the case of the spent resin decontamination method using chemicals, the higher the concentration of the drug decontamination efficiency was higher. In the ultrasound method, foreign matter of the spent resin was removed and was found that the level of radioactivity is below of the MDA. In the microbubbles method, we found that the concentration of the radioactivity decreased after the experiment, so it can be used to the decontamination process of the spent resin. In

  19. Research and development for decontamination system of spent resin in Hanbit Nuclear Power Plant

    International Nuclear Information System (INIS)

    Sung, Gi Hong

    2015-01-01

    When reactor coolant leaks occur due to cracks of a steam generator tube, radioactive materials contained in the primary cooling water in nuclear power plant are forced out toward the secondary systems. At this time the secondary water purification resin in the ion exchange resin tower of the steam generator blowdown system is contaminated by the radioactivity of the leaked radioactive materials, so we pack this in special containers and store temporarily because we could not dispose it by ourselves. If steam generator tube leakage occurs, it produces contaminated spent resins annually about 5,000-7,000 liters. This may increase the amount of nuclear waste productions, a disposal working cost and a unit price of generating electricity in the plant. For this reasons, it is required to develop a decontamination process technique for reducing the radioactive level of these resins enough to handle by the self-disposal method. In this research, First, Investigated the structure and properties of the ion exchange resin used in a steam generator blowdown system. Second, Checked for a occurrence status of contaminated spent resin and a disposal technology. Third, identified the chemical characteristics of the waste radionuclides of the spent resin, and examined ionic bonding and separation mechanism of radioactive nuclear species and a spent resin. Finally, we carried out the decontamination experiment using chemicals, ultrasound, microbubbles, supercritical carbon dioxide to process these spent resin. In the case of the spent resin decontamination method using chemicals, the higher the concentration of the drug decontamination efficiency was higher. In the ultrasound method, foreign matter of the spent resin was removed and was found that the level of radioactivity is below of the MDA. In the microbubbles method, we found that the concentration of the radioactivity decreased after the experiment, so it can be used to the decontamination process of the spent resin. In

  20. Diphonix trademark Resin: A review of its properties and applications

    International Nuclear Information System (INIS)

    Chiarizia, R.; Horwitz, E.P.; Alexandratos, S.D.

    1995-01-01

    The recently developed Diphonix trademark resin is a new multifunctional chelating ion exchange resin containing seminally substituted diphosphonic acid ligands chemically bonded to a styrene-based polymeric matrix. Diphonix can be regarded as a dual mechanism polymer, with a sulfonic acid cation exchange group allowing for rapid access, mostly non-specific, of ions into the polymeric network, and the diphosphonic acid group responsible for specificity (recognition) for a number of metal cations. The Diphonix resin exhibits an extraordinarily strong affinity for actinides, especially in the tetra- and hexavalent oxidation states. It has potential applications in TRU and mixed waste treatment and characterization, and in the development of new procedures for rapid actinide preconcentration and separation from environmental samples. Metal uptake studies have been extended to alkaline earth cations, to transition and post transition metal species, and to metal sorption from neutral or near neutral solutions. Also the kinetic behavior of the resin has been investigated in detail. Influence of the most commonly occurring matrix constituents (Na, Ca, Al, Fe, hydrofluoric, sulfuric, oxalic and phosphoric acids) on the uptake of actinide ions has been measured. This review paper summarizes the most important results studies on the Diphonix resin and gives an overview of the applications already in existence or under development in the fields of mixed waste treatment, actinide separation procedures, treatment of radwaste from nuclear power plants, and removal of iron from copper electrowinning solutions

  1. Use of an Italian pozzolanic cement for the solidification of bead ion exchange resins

    International Nuclear Information System (INIS)

    De Angelis, G.

    1988-05-01

    Granular ion-exchange resins represent a large portion of the medium-active wastes generated at nuclear power stations. The most common practice for their confinement is to mix them with cement paste and cast the mixture in a concrete shell. Such a procedure however does not prove successful in many cases, because of the extreme swelling to which the embedded resin can give rise. This phenomenon has been investigated carefully. In particular, measurements of the swelling pressure have been made together with evaluation of the volume changes of the resin beads due to ion exchange and of the weight increase as a function of relative humidity. The ion exchange capacity, which continues even after incorporation in the cement matrix has also been put into evidence. The conclusion was drawn that a three component diagram (water - dry resin- cement) has to be prepared every time in order to identify the region corresponding to the better formulations. With this in mind the optimum waste loading of 11.5 wt% of dry resin was chosen to incorporate a mixed bed resin (Amberlite IR 120 Na + and IRA 400 Cl - in the weight ratio of 1:1) into an Italian pozzolanic cement (425 type). Several properties of the final waste form have been investigated, ranging from mechanical (crushing strength, tensile strength, flexural strength, ultrasonic pulse velocity, elastic modulus and Poisson ratio), to thermal stability, radiation stability, permeability, leachability and resistance to bacterial attack. Dimensional stability was also measured with the aim of examining the expansion phenomena which can take place in the presence of resin beads. The data obtained are encouraging for future application of the type 425 cement tested in the field of radwastes. An attempt to explain the performance of this binder, based on its intrinsic properties, was also made. (author)

  2. A STUDY ON ADSORPTION AND DESORPTION BEHAVIORS OF 14C FROM A MIXED BED RESIN

    Directory of Open Access Journals (Sweden)

    SEUNG-CHUL PARK

    2014-12-01

    Full Text Available Spent resin waste containing a high concentration of 14C radionuclide cannot be disposed of directly. A fundamental study on selective 14C stripping, especially from the IRN-150 mixed bed resin, was carried out. In single ion-exchange equilibrium isotherm experiments, the ion adsorption capacity of the fresh resin for non-radioactive HCO3− ion, as the chemical form of 14C, was evaluated as 11mg-C/g-resin. Adsorption affinity of anions to the resin was derived in order of NO3− > HCO3− ≥ H2PO4−. Thus the competitive adsorption affinity of NO3− ion in binary systems appeared far higher than that of HCO3− or H2PO4−, and the selective desorption of HCO3− from the resin was very effective. On one hand, the affinity of Co2+ and Cs+ for the resin remained relatively higher than that of other cations in the same stripping solution. Desorption of Cs+ was minimized when the summation of the metal ions in the spent resin and the other cations in solution was near saturation and the pH value was maintained above 4.5. Among the various solutions tested, from the view-point of the simple second waste process, NH4H2PO4 solution was preferable for the stripping of 14C from the spent resin.

  3. Lawps ion exchange column gravity drain of spherical resorcinol formaldehyde resin

    Energy Technology Data Exchange (ETDEWEB)

    Duignan, M. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Herman, D. T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Restivo, M. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Burket, P. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-01-28

    Experiments at several different scales were performed to understand the removal of spherical resorcinol formaldehyde (sRF) ion exchange resin using a gravity drain system with a valve located above the resin screen in the ion exchange column (IXC). This is being considered as part of the design for the Low Activity Waste Pretreatment System (LAWPS) to be constructed at the DOE Hanford Site.

  4. Apparatus for waste disposal of radioactive hazardous waste

    International Nuclear Information System (INIS)

    Burack, R.D.; Stenger, W.J.; Wolfe, C.R.

    1992-01-01

    This patent describes an apparatus for concentrating dissolved and solid radioactive materials carried in a waste water solution containing a hazardous chelating agent used for cleaning nuclear equipment. It comprises oxidizing chamber means, separator means coupled to the oxidizing chamber means; ion exchange means containing an ion exchange resin; dryer means for receiving the radioactive solids from the separator means and for producing dry solids; and packaging means for receiving the dry solids and spent ion exchange resins containing the removed dissolved radioactive materials and for packaging the dry solids and spent resins in solid form

  5. Application of THOR-Technology on resins

    International Nuclear Information System (INIS)

    Lorenzen, J.; Lindberg, M.

    2003-01-01

    The THermal Organic Reduction (THOR) process, developed and patented by studsvik utilises pyrolysis / steam reforming technology. The THOR-process provides a reliable and safe method for treating a wide variety of LLW in a unique, fluidised bed treatment system at moderate temperature. This technology is suitable for processing hazardous, mixed and dry active LLW with appropriate waste feed modifications. Both solid and liquid LLRW and ILRW streams including ion exchange resins, activated carbon (charcoal), graphite, oils, solvents and cleaning solutions with contact radiation levels of up to 4 Sv/hr can be processed. Studsvik has completed over four years of operation at its facility at Erwin, Tennessee, USA. During this period studsvik has processed more than 1,5 thousand tons of radioactive ion exchange bead resins. powdered filter media and active carbon, with a cumulative total radioactivity of about 7 (E+8) MBq. Operations have demonstrated consistent, reliable, robust operating characteristics. Due to the widely varying characteristics of the incoming waste streams various efficiencies and volume reductions have been experienced. Input waste has varied in total inorganic content from 90%. A substantial element of this variability has been the ''soluble salt'' content of the input waste streams. Final reformed residue comprises a non-dispersible, granular solid which is suitable for long-term storage or direct burial in a qualified container. Special containers, THOR-liners, are available from studsvik for the transport of waste from the customer to the Erwin facility and HICs (high integrity containers) for transport of the residues to Barnwell. The paper will give an overview of the last four years of commercial operations processing LLRW from commercial nuclear power plans. (orig.)

  6. Method of processing liquid wastes

    International Nuclear Information System (INIS)

    Naba, Katsumi; Oohashi, Takeshi; Kawakatsu, Ryu; Kuribayashi, Kotaro.

    1980-01-01

    Purpose: To process radioactive liquid wastes with safety by distillating radioactive liquid wastes while passing gases, properly treating the distillation fractions, adding combustible and liquid synthetic resin material to the distillation residues, polymerizing to solidify and then burning them. Method: Radioactive substance - containing liquid wastes are distillated while passing gases and the distillation fractions containing no substantial radioactive substances are treated in an adequate method. Synthetic resin material, which may be a mixture of polymer and monomer, is added together with a catalyst to the distillation residues containing almost of the radioactive substances to polymerize and solidify. Water or solvent in such an extent as not hindering the solidification may be allowed if remained. The solidification products are burnt for facilitating the treatment of the radioactive substances. The resin material can be selected suitably, methacrylate syrup (mainly solution of polymethylmethacrylate and methylmethacrylate) being preferred. (Seki, T.)

  7. Leach testing of simulated ion-exchange resin waste solidified in cement

    International Nuclear Information System (INIS)

    Muurinen, A.K.; Uotila, P.I.; Ovaskainen, R.M.

    Leach tests were carried out on ion-exchange resins solidified in cement. Three product mixtures, two isotopes and four leachants at two temperatures, were tested. The increase of resin content increased the leaching of Cs-137; the effect of silix admixture was negligible. The type of the leachant has a stronger influence on Co-60 than on Cs-137. The increase of temperature usually also increased leaching. (author)

  8. The PILO process: zeolites and titanates in the treatment of spent ion exchange resins

    International Nuclear Information System (INIS)

    Hultgren, Aa.; Thegerstroem, C.; Forberg, S.; Westermark, T.; Faelt, L.

    1981-01-01

    Spent ion exchange resins from power reactor operation contain more than 95% of the total radioactivity of wet reactor wastes. Cementation and bituminization are the two methods applied in Sweden up to now for the immobilization of spent resins. Over the last years, however, research and development work has resulted in a proposed process (PILO), where > 99.9 % of cesium and strontium and around 90 % of other radioactive nuclides are eluted from the spent resins and sorbed in zeolites and titanates in a chromatographic process. The inorganic sorbents are dried after loading and sintered to yield long-term stable products, while the treated resins may be incinerated to give ash residues of fairly short-lived activity. The development work has included production, characterization and testing of different zeolites and titanates, bench-scale optimization of the chromatographic process using actual spent resins, heat treatment of the loaded inorganic sorbents, and resin incineration. Over-all system design studies including transport requirements, integrated process flowsheets, and cost estimates are now in progress. The aim is to have a sufficient basis during spring 1982 to decide on the merits of a PILO plant at the planned repository for low and medium level waste (SFR), to be commissioned in 1988. (Auth.)

  9. Degradation of resins in EPICOR-II prefilters from Three Mile Island

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Johnson, D.A.

    1990-01-01

    The Low-Level Waste Data Base Development--EPICOR-II Resin/Liner Investigation Program funded by the U.S. Nuclear Regulatory Commission is investigating the chemical and physical conditions of the synthetic ion exchange resins contained in several EPICOR-Il prefilters. Those prefilters were used during cleanup of contaminated water from the Three Mile Island Nuclear Power Station after the March 1979 accident. This paper summarizes results and analyses of the third sampling of resins from prefilters PF-8 and -20. Results are compared with baseline data from tests performed on unirradiated resins supplied by Epicor, Inc. to determine if degradation has occurred due to the high internal radiation dose. Results also are compared with results from tests performed on resins obtained from the first and second samplings of those two prefilters. 17 refs., 4 figs., 4 tabs

  10. Isotopic analysis of radioactive waste packages (an inexpensive approach)

    International Nuclear Information System (INIS)

    Padula, D.A.; Richmond, J.S.

    1983-01-01

    A computer printout of the isotopic analysis for all radioactive waste packages containing resins, or other aqueous filter media is now required at the disposal sites at Barnwell, South Carolina, and Beatty, Nevada. Richland, Washington requires an isotopic analysis for all radioactive waste packages. The NRC (Nuclear Regulatory Commission), through 10 CFR 61, will require shippers of radioactive waste to classify and label for disposal all radioactive waste forms. These forms include resins, filters, sludges, and dry active waste (trash). The waste classification is to be based upon 10 CFR 61 (Section 1-7). The isotopes upon which waste classification is to be based are tabulated. 7 references, 8 tables

  11. Cementation of Radioactive Waste from a PWR with Calcium Sulfoaluminate Cement

    International Nuclear Information System (INIS)

    Li, J.

    2013-01-01

    Spent radioactive ion-exchange resin (SIER) and evaporation concentrates are radioactive wastes that are produced at by pressurized water reactor (PWR) nuclear power stations. Borate, which is used as a retardent for cement, is also present as a moderator in a PWR, therefore, borate will be present in both ion-exchange resins and evaporation concentrates. In this study the use of Calcium sulfoaluminate cements (SAC) as encapsulation medium for these waste streams was investigated. The study involved the manufacturing of different cement test samples with different amounts of SAC cement, waste resins (50% water content) and admixtures. In order to reduce hydration heat during 200 L solidification experiments, different admixtures were investigated. Initial results based on compressive strength tests and hydration temperature studies, indicated that zeolite was the best admixture for the current waste form. Experiments indicated that the addition of resin material into the current cement matrix reduces the hydration heat during curing Experimental results indicated that a combination of SAC (35 wt. %), zeolite (7 wt. %) mix with 42 wt. % resins (50% water content) and 16 wt. % of water forms a optimum cured monolith with low hydration heat. The microstructures of hydrated OPC, SAC and SAC with zeolite addition were studied using a Scanning Electron Microscopy (SEM). SEM results indicated that the SAC matrices consist of a needle type structure that changed gradually into a flake type structure with the addition of zeolite. Additionally, the presence of zeolite material inside the SAC matrix reduced the leaching rates of radionuclides significantly. In a final 200 L grouting test, measured results indicated a hydration temperature below 90oC withno thermal cracks after solidified. The influence of radiation on the compressive strength and possible gas generation (due to radiolysis) on cement waste forms containing different concentrations ion exchange resin was

  12. Regeneration of Pt-catalysts deactivated in municipal waste flue gas with H2/N2 and the effect of regeneration step on the SCR catalyst

    DEFF Research Database (Denmark)

    Due-Hansen, Johannes; Rasmussen, Søren Brik; Kustov, Arkady

    The deactivation performance of Pt-catalysts for CO oxidation has been studied in relation to use in sewage sludge municipal waste burners, where HMDS was found to poison the industrial catalyst in a similar way to the model Pt/TiO2 catalyst. A promising regeneration procedure was developed based...... on reduction with hydrogen. This procedure had negligible effect on the performance of the SCR catalyst. After treatment with 2% H2, 8% O2 in N2 for one hour, a slight better NO SCR activity was observed due to increase in the concentration V4+ sites. However, after exposure in normal NO SCR gases the activity...

  13. Regeneration of Pt-catalysts deactivated in municipal waste flue gas with H2/N2 and the effect of regeneration step on the SCR catalyst

    DEFF Research Database (Denmark)

    Due-Hansen, Johannes; Rasmussen, Søren Birk; Kustov, Arkadii

    Municipal waste flue gas was previously found to deactivate the Pt-based CO oxidation catalyst severely. In the specific case studied, siloxanes were found to cause the deactivation. An on-site method for complete regeneration of the catalyst activity was found without shutdown of the flue gas...... stream, i.e. by in situ treatment of the Pt-catalyst by reductive H2-gas. However, introduction of H2 gas in the gas stream could also affect other units in the tail pipe gas cleaning system. Of special interest here, is the effect of hydrogen gas on the performance of the deNOx + SCR catalytic process...

  14. Method of processing low-level radioactive liquid wastes

    International Nuclear Information System (INIS)

    Matsunaga, Ichiro; Sugai, Hiroshi.

    1984-01-01

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

  15. Encapsulation of radioactive waste

    International Nuclear Information System (INIS)

    Pordes, O.; Plows, J.P.

    1980-01-01

    A method is described for encapsulating a particular radioactive waste which consists of suspending the waste in a viscous liquid encapsulating material, of synthetic resin monomers or prepolymers, and setting the encapsulating material by addition or condensation polymerization to form a solid material in which the waste is dispersed. (author)

  16. Regeneration performance of CO2-rich solvents by using membrane vacuum regeneration technology: Relationships between absorbent structure and regeneration efficiency

    International Nuclear Information System (INIS)

    Yan, Shuiping; Fang, Mengxiang; Wang, Zhen; Luo, Zhongyang

    2012-01-01

    Highlights: ► MVR may be viable to successfully use less valuable heat to replace high grade steam. ► Increasing OH and amine groups will increase the regeneration efficiency. ► Absorbents with a four carbon chain length will be more attractive to MVR. ► Amino acid salts will be more appropriate for MVR. ► HRM conducted at ambient pressure and low temperature is inferior to MVR. -- Abstract: In order to give a better understanding for the selection of suitable absorbents for the novel membrane vacuum regeneration technology (MVR) which has the potential to reduce CO 2 energy requirement by utilizing the waste heat or low-grade energy, an experimental study to determine the relationships between chemical structure and vacuum regeneration behavior of CO 2 absorbents at 70 °C and 10 kPa was performed. Eleven typical absorbents with different functional groups in their chemical structures were investigated in terms of vacuum regeneration efficiencies. Results showed that the regeneration efficiency decreased with an increase of number of activated hydrogen atom in amine group and decreased with the number of hydroxyl group. Especially, more attention should be paid to these alkanolamines with one hydrogen atom in amine group and two or more hydroxyl groups in the structures due to their better comprehensive performance in regeneration, absorbent loss and CO 2 absorption aspects. Increasing the carbon chain length and amine groups in the absorbent structure contributed to the improvement of regeneration performance and reduction of absorbent volatile loss. These absorbents with a four carbon chain length bonded at amine group might be more attractive to MVR. Furthermore, polyamines were superior to monoamines in terms of higher regeneration efficiencies and lower absorbent losses. Additionally, the individual effects of the potassium carboxylate group and hydroxymethylene group were also compared in this study. Results showed that amino acid salts were more

  17. Regeneration and reuse waste from an edible oil refinery.

    Science.gov (United States)

    Boukerroui, Abdelhamid; Belhocine, Lydia; Ferroudj, Sonia

    2017-08-21

    A spent bleaching earth (SBE) from an edible oil refinery has been regenerated by thermal processing in oven, followed by washing with a cold solution of hydrochloric acid (1M). Optimal regeneration conditions have been controlled by decolorization tests of degummed and neutralized soybean oil. Optimal values of treatment (temperature 350°C, carbonization time 01 h, and HCl concentration 1M) gave a very efficient material. After bleaching oil by regenerated spent bleaching earth (RSBE), the chlorophyll-a and β-carotenes contained in crude edible oil and observed respectively at 430, 454, and 483 nm, value of λ max , are very much decreased. The results obtained after decolorization of edible oil by RSBE material indicate, that, during the process, the bleaching oil did not undergo any changes in the free fatty acid content. The peroxide value (PV) was reduced from 4.2 to 1.8 meq O 2 /kg, and the color has been improved (Lovibond color yellow/red: from 50/0.5 to 2.7/0.3, respectively). The RSBE material obtained was characterized by several techniques (FTIR, SEM). The results show that the heat treatment did not affect the mineral structure of RSBE, and the regenerated material recovered its porous structure.

  18. Use of Anion Exchange Resins for One-Step Processing of Algae from Harvest to Biofuel

    Directory of Open Access Journals (Sweden)

    Martin Poenie

    2012-07-01

    Full Text Available Some microalgae are particularly attractive as a renewable feedstock for biodiesel production due to their rapid growth, high content of triacylglycerols, and ability to be grown on non-arable land. Unfortunately, obtaining oil from algae is currently cost prohibitive in part due to the need to pump and process large volumes of dilute algal suspensions. In an effort to circumvent this problem, we have explored the use of anion exchange resins for simplifying the processing of algae to biofuel. Anion exchange resins can bind and accumulate the algal cells out of suspension to form a dewatered concentrate. Treatment of the resin-bound algae with sulfuric acid/methanol elutes the algae and regenerates the resin while converting algal lipids to biodiesel. Hydrophobic polymers can remove biodiesel from the sulfuric acid/methanol, allowing the transesterification reagent to be reused. We show that in situ transesterification of algal lipids can efficiently convert algal lipids to fatty acid methyl esters while allowing the resin and transesterification reagent to be recycled numerous times without loss of effectiveness.

  19. Corrosion of metal containers containing cemented radioactive wastes

    International Nuclear Information System (INIS)

    Duffo, G.S.; Farina, S.B.; Schulz, F.M.; Marotta, F

    2010-01-01

    Nuclear activities generate different kinds of radioactive wastes. In the case of Argentina, wastes classified as low and medium level are conditioned in metal drums for final disposal in a repository whose design is based on the use of multiple and independent barriers. Nuclear energy plants generate a large volume of mid-level radioactive wastes, consisting mainly of ion-exchange resins contaminated by fission products. Other contaminated products such as gloves, papers, clothing, rubber and plastic tubing, can be incinerated and the ashes from the combustion also constitute wastes that must be disposed of. These wastes (resins and ashes) must be immobilized in order to avoid the release of radionuclides into the environment. The wastes usually undergo a process of cementing to immobilize them. This work aims to systematically study the process of degradation by corrosion of the steel drums in contact with the cemented resins and with the ashes cemented with the addition of different types and concentrations of aggressive compounds (chloride and sulfate). The specimens are configured so that the parameters of interest for the steel in contact with the cemented materials can be measured. The variables of corrosion potential, electric resistivity of the matrix and polarization resistance (PR) were monitored and show that the presence of chloride increases the susceptibility to corrosion of the drum steel that is in contact with the cement resin matrix

  20. Long-term properties of TVO's bituminized resins

    International Nuclear Information System (INIS)

    Valkiainen, M.; Vuorinen, U.

    1989-06-01

    Long-term properties of bituminized spent ion-exchange resins from Olkiluoto power plant have been studied since 1981. This report summarizes the results on water uptake and leaching obtained up till now. It is observed that water uptake in excess of rewetting of the ion-exchange resins is taking place. Leach test in water equilibrated with cement have been performed for about five years. Separation of granular resin particles caused by density differences observed in former experiments was further studied and confirmed in this work. Anaerobic corrosion of a steel drum has been studied in laboratory conditions generally giving corrosion rates below l μm/a. Radiolytic gases will accumulate and be trapped in the waste product. The rate of swelling is estimated by a specially constructed device based on ultrasonic distance meter observing changes in level of the product surface in the drum

  1. Phenolic cation-exchange resin material for recovery of cesium and strontium. [Patent application

    Science.gov (United States)

    Ebra, M.A.; Wallace, R.M.

    1982-05-05

    A phenolic cation exchange resin with a chelating group has been prepared by reacting resorcinol with iminodiacetic acid in the presence of formaldehyde at a molar ratio of about 1:1:6. The material is highly selective for the simultaneous recovery of both cesium and strontium from aqueous alkaline solutions, such as, aqueous alkaline nuclear wate solutions. The organic resins are condensation polymers of resorcinol and formaldehyde with attached chelating groups. The column performance of the resins compares favorably with that of commercially available resins for either cesium or strontium removal. By combining Cs/sup +/ and Sr/sup 2 +/ removal in the same bed, the resins allow significant reduction of the size and complexity of facilities for processing nuclear waste.

  2. Pilot Plant for treating of spent exchange resins

    International Nuclear Information System (INIS)

    Iglesias, Alberto M.; Raffo Calderon, Maria del C.; Varani, Jose L.

    2004-01-01

    Spent exchange resins that have been accumulating during the last operational 30 years in Atucha I nuclear power plant (NPP) are a 'problematic' waste. These spent resins conform an intermediate level waste due to the total content of alpha, beta and gamma emitters (some samples of spent resins were analyzed in 2003). For this reason its treatment is more expensive since it is necessary to add more safety barriers for its final disposition and also for the radioprotection actions that are involved. Using sulfuric acid solutions it is possible to elute from the spent resins the ions that are retained. In the same operation are eluted Cobalt, Cesium and alpha emitters since that all these elements react as cations in aqueous solution. Decontamination by electrochemical methods was analyzed as an interesting method to apply after elution operation to these spent resins since that with the decontamination process it is possible to obtain a solid without activity and concentrate the activity in cells that are small in volume and its manipulation doesn't present any extra complication. Experiments made with active samples taken from the deposit were successful. Because of these results it was built a small plant to treat a batch of 100 dm 3 of wet spent exchange resins. Some problems with the material that was in the deposit together with spent resins caused that we had to plan a more complex strategy to obtain a complete decontamination of the spent resins (in this stage we used the cobalt retention cell that was described in other paper to retain Cobalt and alpha emitters and a sample of zeolites from Argentina ores to retain Cesium). Due to alpha emitters act electrochemically like cations it was possible to retain altogether with ionic Cobalt on the copper amalgam electrode. Working in the non-active lab with alcoholic solutions it was possible to retain ionic Cesium on a copper electrode (copper is covered by mercury fine film which forms a solid amalgam) with a

  3. Development of bitumization plant of radioactive waste

    International Nuclear Information System (INIS)

    Lee, S.H.; Chun, K.S.; Song, H.Y.; Park, S.H.; Park, W.J.; Ryoo, R.; Cho, W.J.; Wee, K.S.; Lee, J.D.

    1981-01-01

    This study is to develop the bitumization techniques of radioactive waste concentrates from nuclear power plants (PWR), Cold-run and hot-run bitumization pilot plant were designed and constructed, and testing was carried out. The results show that the range of pH 7 to 8, in which a soluble product could form, is suitable for feeding in the waste and for its incorporation into asphalt, domestic blown asphalt is suitable up to an exposure of 10 8 rad. The estimated viscosity of the asphalt/waste product is about 14 poises at 175 0 C and the asphalt/wastes products in simulated sea-water for three weeks showed no evidence of swelling. The leaching rate from the spent resin/asphalt products in distilled water less than 1/1000 of that from the cement-products. It is considered that the resin content which is suitable for the incorporation of spent resin into asphalt is less than 50 wt%

  4. Process for preparing wastes for non-pollutant disposal

    International Nuclear Information System (INIS)

    Debus, A.A.G.; Rosenstiel, T.L.

    1982-01-01

    In disposing of wastes, especially those containing radioactive or toxic substances, wherein the liquid or finely divided solid waste is mixed with water and a hydraulic binder and allowed to set, the binder used comprises a mixture of calcium sulphate hemihydrate, a water-dispersible melamine formaldehyde resin which when cured is hydrophobic and sufficient cross-linking agent to cure the resin. (author)

  5. Chemical derivation to enhance the chemical/oxidative stability of resorcinol-formaldehyde (R-F) resin

    International Nuclear Information System (INIS)

    Hubler, T.L.; Shaw, W.J.; Brown, G.N.; Linehan, J.C.; Franz, J.A.; Hart, T.R.; Hogan, M.O.

    1996-09-01

    Tank wastes at Hanford and SRS contain highly alkaline supernate solutions of conc. Na, K nitrates with large amounts of 137 Cs. It is desirable to remove and concentrate the highly radioactive fraction for vitrification. One candidate ion exchange material for removing the radiocesium is R-F resin. This report summarizes studies into synthesis and characterization of 4-derivatized R-F resins prepared in pursuit of more chemically/oxidatively robust resin. 85% 4-fluororesorcinol/15% phenol formaldehyde resin appears to have good stability in alkaline solution, although there may be some nucleophilic displacement reaction during synthesis; further studies are needed

  6. Leaching of nuclear power reactor waste forms

    International Nuclear Information System (INIS)

    Endo, L.S.; Villalobos, J.P.; Miyamoto, H.

    1987-01-01

    The leaching tests for immobilized power reactor wastes carried out at IPEN are described. These wastes forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. Three years leaching results are reported. The cesium diffuvity coefficients determined out of these results are about 1 x 10 -8 cm 2 /s for boric acid waste form and 9 x 10 -9 cm 2 /s for ion-exchange resin waste. Strontium diffusivity coefficients found are about 3 x 10 -11 cm 2 /s and 9 x 10 -11 cm 2 /s respectively. (Author) [pt

  7. Effect of the chemical structure of anion exchange resin on the adsorption of humic acid: behavior and mechanism.

    Science.gov (United States)

    Shuang, Chendong; Wang, Jun; Li, Haibo; Li, Aimin; Zhou, Qing

    2015-01-01

    Polystyrenic (PS) anion-exchange resin and polyacrylic (PA) anion-exchange resin were used to investigate the effect of resin chemical structure on the adsorption of humic acid (HA). Due to the rearrangement of HA to form layers that function as barricades to further HA diffusion, PS resin exhibited 12.4 times slower kinetics for the initial adsorption rate and 8.4 times for the diffusion constant in comparison to that of the PA resin. An HA layer and a spherical cluster of HA can be observed on the surface of the PS and PA resins after adsorption, respectively. The considerable difference in HA adsorption between the PS and PA resins was due to the difference in molecule shape for interaction with different resin structures, which can essentially be explained by the hydrophobicity and various interactions of the PS resin. A given amount of HA occupies more positively charged sites and hydrophobic sites on the PS resin than were occupied by the same amount of HA on the PA resin. Increased pH resulted in an increase of HA adsorption onto the PA resin but a decrease in adsorption onto PS resin, as the non-electrostatic adsorption led to electrostatic repulsion between the HA attached to the resin and the HA dissolved in solution. These results suggest higher rates of adsorption and higher regeneration efficiency for interaction of HA with more hydrophilic anion exchange materials. Copyright © 2014 Elsevier Inc. All rights reserved.

  8. Study of plasma off-gas treatment from spent ion exchange resin pyrolysis.

    Science.gov (United States)

    Castro, Hernán Ariel; Luca, Vittorio; Bianchi, Hugo Luis

    2017-03-23

    Polystyrene divinylbenzene-based ion exchange resins are employed extensively within nuclear power plants (NPPs) and research reactors for purification and chemical control of the cooling water system. To maintain the highest possible water quality, the resins are regularly replaced as they become contaminated with a range of isotopes derived from compromised fuel elements as well as corrosion and activation products including 14 C, 60 Co, 90 Sr, 129 I, and 137 Cs. Such spent resins constitute a major proportion (in volume terms) of the solid radioactive waste generated by the nuclear industry. Several treatment and conditioning techniques have been developed with a view toward reducing the spent resin volume and generating a stable waste product suitable for long-term storage and disposal. Between them, pyrolysis emerges as an attractive option. Previous work of our group suggests that the pyrolysis treatment of the resins at low temperatures between 300 and 350 °C resulted in a stable waste product with a significant volume reduction (>50%) and characteristics suitable for long-term storage and/or disposal. However, another important issue to take into account is the complexity of the off-gas generated during the process and the different technical alternatives for its conditioning. Ongoing work addresses the characterization of the ion exchange resin treatment's off-gas. Additionally, the application of plasma technology for the treatment of the off-gas current was studied as an alternative to more conventional processes utilizing oil- or gas-fired post-combustion chambers operating at temperatures in excess of 1000 °C. A laboratory-scale flow reactor, using inductively coupled plasma, operating under sub-atmospheric conditions was developed. Fundamental experiments using model compounds have been performed, demonstrating a high destruction and removal ratio (>99.99%) for different reaction media, at low reactor temperatures and moderate power consumption

  9. Plasma arc pyrolysis of radioactive ion exchange resin

    International Nuclear Information System (INIS)

    Pickles, C.A.; Toguri, J.M.

    1992-01-01

    This paper reports on two ion exchange resins (IRN 77 and IRN 78) which were pyrolysed in a plasma-arc furnace. Both continuous and batch tests were performed. Volume reduction ratios of 10 to 1 and 10 to 3.5 were achieved for IRN 78 and IRN 77 respectively. The product of the resin pyrolysis was a char which contained the radioactive elements such as cobalt. The off-gases consisted of mainly hydrogen and carbon monoxide. There was a relatively small amount of dust in the off-gases. At the present time radioactive ion exchange resign is being kept in storage. The volume of this waste is increasing and it is important that the volume be reduce. The volume reduction ratio should be of the order of ten-to-one. Also, it is required that the radioactive elements can be collected or fixed in a form which could easily be disposed of. Plasma arc treatment offers considerable potential for the processing of the waste

  10. Treatment of aqueous diethyl phthalate by adsorption using a functional polymer resin.

    Science.gov (United States)

    Xu, Zhengwen; Zhang, Weiming; Pan, Bingcai; Lv, Lu; Jiang, Zhengmao

    2011-01-01

    To study the adsorptive separation efficiency, adsorption and desorption performances of diethyl phthalate (DEP) were investigated with a functional polymer resin (NDA-702). A macroporous polymer resin (XAD-4) and a coal-based granular activated carbon (AC-750) were chosen for comparison. The kinetic adsorption data obeyed the pseudo-second-order rate model, and the adsorption processes were limited by both film and intraparticle diffusions. Adsorption equilibrium data were well fitted by the Freundlich equation, and the larger uptake and higher selection of NDA-702 than AC-750 and XAD-4 was probably due to the microporous structure, phenyl rings and polar groups on NDA-702. Thermodynamic adsorption studies indicated that the test adsorbents spontaneously adsorbed DEP, driven mainly by enthalpy change. Continuous fixed-bed runs demonstrated that there no significant loss of the resin's adsorption capacity and there was complete regeneration of NDA-702. The results suggest that NDA-702 has excellent potential as an adsorption material for water treatment.

  11. Decontamination impacts on solidification and waste disposal

    International Nuclear Information System (INIS)

    Kempf, C.R.; Soo, P.

    1988-01-01

    Research to determine chemical and physical conditions which could lead to thermal excursions, gas generation, and/or general degradation of decontamination-reagent-loaded resins has shown that IRN-78, IONAC A-365, and IRN-77 organic ion exchange resin moisture contents vary significantly depending on the counter ion ''loading.'' The extent/vigor of the reaction is very highly dependent on the degree of dewatering of the resins and on the method of solution addition. The heat generation may be due, in part, to the heat of neutralization. In studies of the long-term compatibility effects of decontamination waste resins in contact with waste package container materials in the presence of decontamination reagents, radiolysis products and gamma irradiation, it has been found that the corrosion of carbon steel and austenitic stainless steel in mixed bed resins is enhanced by gamma irradiation. However, cracking in high density polyethylene is essentially eliminated because of the rapid removal of oxygen from the environment by gamma-induced oxidation of the large resin mass. 13 refs., 10 figs., 3 tabs

  12. INEL studies concerning solidification of low-level waste in cement

    International Nuclear Information System (INIS)

    Mandler, J.W.

    1989-01-01

    The Idaho National Engineering Laboratory (INEL) has performed numerous studies addressing issues concerning the solidification of low-level radioactive waste in cement. These studies have been performed for both the Nuclear Regulatory Commission (NRC) and the Department of Energy (DOE). This short presentation will only outline the major topics addressed in some of these studies, present a few conclusions, and identify some of the technical concerns we have. More details of the work and pertinent results will be given in the Working Group sessions. The topics that have been addressed at the INEL which are relevant to this Workshop include (1) solidification of ion-exchange resins and evaporator waste in cement at commercial nuclear power plants, (2) leachability and compressive strength of power plant waste solidified in cement, (3) suggested guidelines for preparation of a solid waste process control program (PCP), (4) cement solidification of EPICOR-II resin wastes, and (5) performance testing of cement-solidified EPICOR-II resin wastes

  13. The management of spent resins at rad waste installation (RWI)

    International Nuclear Information System (INIS)

    Johan, Bahdir

    2002-01-01

    During years 2000 RWI has been managed 1.596 liters the spent resins from MPR, multipurpose reactor G.A. Siwabessy (RSG-GAS). The spent resins containing 6 0 C o, 1 37 C s and 8 9 S r. The immobilization of spent resins using Portland cement type I and sand in shell 950 liter has been carried out. The immobilization product quality was characterized by measurement of surface exposure, density, compression strength and leaching rate. The samples each being 3000 g are made by taking cement slurry from the 950 liter shell. The samples are molded in the form of cylinder of 50 mm diameter and height 50 mm. The density was determined by weighing and measuring sample's volume. The compression strength was determinate using P aul Weber a pparatus. The leaching rate by immerse of the samples in the demineralized water and the rate the contact surface dose measures using survey meter. The results show that the density ρ= (1,72 ± 0,01) s/d 1,73 ± 0,03) g.cm - 3 , the compression strength Γ= (20,10 ± 0,02 s/d 20,18 ± 0,02) N.mm - 2 , the leaching rate R n = (1,32x10 - 2 s /d 2,0x10 - 3 ) g.cm - 2 . h - 1 , the contact surface rate D 0 = (0,374 s/d 0,79) mrem.jam - 1 a nd 1m distance from the surface D 1 = (0,18 s/d 0,19) mrem.jam - 1 h ave been in compliance with the criteria recommended by IAEA the quality of spent resins concrete in shell 950 liters (number 22C to 27C) is qualified for ultimate disposal

  14. Ion Exchange Resin and Clay Vitrification by Plasma Discharges

    International Nuclear Information System (INIS)

    Diaz A, Laura V.; Pacheco S, Joel O.; Pacheco P, Marquidia; Monroy G, Fabiola; Emeterio H, Miguel; Ramos F, Fidel

    2006-01-01

    The lack of treatment of a low and intermediate level radioactive waste (LILRW) lead us to propose a vitrification process based on a plasma discharge; this technique incorporates LILRW into a matrix glass composed of ceramic clays material. The Mexican Institute of Nuclear Research (ININ), uses an ion exchange resin IRN 150 (styrene-divinilbence copolymer) in the TRIGA MARK III nuclear reactor. The principal objective of this resin is to absorb particles containing heavy metals and low-level radioactive particles. Once the IRN 150 resin filter capacity has been exceeded, it should be replaced and treated as LILRW. In this work, a transferred plasma system was realized to vitrify this resin taking advantage of its high power density, enthalpy and chemical reactivity as well as its rapid quenching and high operation temperatures. In order to characterize the morphological structure of these clay samples, Scanning Electron Microscopy (SEM), X-Ray Diffraction (XRD) and Thermogravimetric analysis (TGA) techniques were applied before and after the plasma treatment

  15. Supercritical water oxidation of ion exchange resins: Degradation mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Leybros, A.; Roubaud, A. [CEA Marcoule, DEN DTCD SPDE LFSM, F-30207 Bagnols Sur Ceze (France); Guichardon, P. [Ecole Cent Marseille, F-13451 Marseille 20 (France); Boutin, O. [Aix Marseille Univ, UMR CNRS 6181, F-13545 Aix En Provence 4 (France)

    2010-07-01

    Spent ion exchange resins are radioactive process wastes for which there is no satisfactory industrial treatment. Supercritical water oxidation could offer a viable treatment alternative to destroy the organic structure of resins and contain radioactivity. IER degradation experiments were carried out in a continuous supercritical water reactor. Total organic carbon degradation rates in the range of 95-98% were obtained depending on operating conditions. GC-MS chromatography analyses were carried out to determine intermediate products formed during the reaction. Around 50 species were identified for cationic and anionic resins. Degradation of poly-styrenic structure leads to the formation of low molecular weight compounds. Benzoic acid, phenol and acetic acid are the main compounds. However, other products are detected in appreciable yields such as phenolic species or heterocycles, for anionic IERs degradation. Intermediates produced by intramolecular rearrangements are also obtained. A radical degradation mechanism is proposed for each resin. In this overall mechanism, several hypotheses are foreseen, according to HOO center dot radical attack sites. (authors)

  16. Reducing resin use in floor drain processing system

    International Nuclear Information System (INIS)

    Flint, W.; Hobart, S.A.; Miller, A.D.

    1995-01-01

    The Kewaunee Nuclear Power Plant utilizes two mixed bed demineralizers for processing floor drain wastes. These demineralizers were originally designed for stream generator blowdown treatment, but were not needed for that purpose. Effluent from the resin beds is monitored for radioactivity and released for discharge. Plant radwaste inleakage volumes and resin disposal volumes were low in comparison with industry averages, but decontamination factors through the treatment system were less than desirable. Release criteria for discharges always had been met, but plant personnel wished to decrease their already low discharges of radioactive species, reduce their resin disposal costs, and provide a margin of safety in the unlikely event that fuel damage would be experienced during an operating cycle. This paper describes the study initiated to address those issues, the findings of the study, and results of implementing some of the study recommendations

  17. Spray-type drying unit for spent ion exchange resins, sludges and radioactive concentrates

    International Nuclear Information System (INIS)

    Raibaud, J.

    1986-01-01

    The process for drying radwaste in the liquid form or in aqueous suspension is a very attractive solution from the standpoint of volume reduction. Most of the existing drying facilities are not well adapted for drying the varieties of aqueous waste produced by the nuclear research centres and nuclear power plants, such as: - ion exchange resins, bead type or powdered resins, - centrifuge sludges, - settling sludges, - evaporator bottoms. Technicatome has selected the LEAFLASH process developed by Rhone Poulenc Recherches for drying the nuclear aqueous waste. This process has been well tried at full scale in a large number of industrial branches. The advantages of the process have been confirmed by the results obtained in operating a pilot facility. They include: - high flexibility in operation: - quick start-up and stoppage procedures, - adaptation to a wide spectrum of liquid waste without significant changes in the adjustment of the device. - compactness, - low power consumption, - complete drying of the waste for any initial concentration [fr

  18. Testing of variables which affect stablity of cement solidified low-level waste

    International Nuclear Information System (INIS)

    Boris, G.F.

    1989-01-01

    This paper describes the test program undertaken to investigate variables which could affect the stability of cement solidified low-level waste and to evaluate the effect of these variables on certain tests prescribed in the Technical Position on Waste Form. The majority of the testing was performed on solidified undepleted bead resin, however, six additional waste types, suggested by the NRC, were tested. The tested variables included waste loading, immersion duration, depletion level, ambient cure duration, curing environment, immersion medium and waste type. Of these, lower waste loadings, longer ambient cures prior to testing and immersion in demineralized water versus simulated sea water and potable water resulted in higher compressive strengths for bead resin samples. Immersion times longer than 90 days did not affect the resin samples. Compressive strengths for other waste types varied depending upon the waste. The strengths of all waste types exceeded the minimum criterion by at least a factor of four, up to a factor of forty. The higher waste loadings exhibit strengths less than the lower waste loadings

  19. Accurate determination of 41Ca concentrations in spent resins from the nuclear industry by Accelerator Mass Spectrometry

    International Nuclear Information System (INIS)

    Nottoli, Emmanuelle; Bourlès, Didier; Bienvenu, Philippe; Labet, Alexandre; Arnold, Maurice; Bertaux, Maité

    2013-01-01

    The radiological characterisation of nuclear waste is essential for managing storage sites. Determining the concentration of Long‐Lived RadioNuclides (LLRN) is fundamental for their long-term management. This paper focuses on the measurement of low 41 Ca concentrations in ions exchange resins used for primary fluid purification in Pressurised Water Reactors (PWR). 41 Ca concentrations were successfully measured by Accelerator Mass Spectrometry (AMS) after the acid digestion of resin samples, followed by radioactive decontamination and isobaric suppression through successive hydroxide, carbonate, nitrate and final CaF 2 precipitations. Measured 41 Ca concentrations ranged from 0.02 to 0.03 ng/g, i.e. from 0.06 to 0.09 Bq/g. The 41 Ca/ 60 Co activity ratios obtained were remarkably reproducible and in good agreement with the current ratio used for resins management. - Highlights: • In the context of radioactive waste management, this study aimed at measuring 41 Ca in spent resins using Accelerator Mass Spectrometry. • A chemical treatment procedure was developed to quantitatively recover calcium in solution and selectively extract it. • Developed firstly on synthetic matrices, the chemical treatment procedure was then successfully applied to real resin samples. • Accelerator mass spectrometry allowed measuring concentrations of 41 Ca in spent resins as low as 0.02 ng/g of dry resin. • Final results are in agreement with current data used for spent resins management

  20. Method of processing radioactive wastes

    International Nuclear Information System (INIS)

    Funabashi, Kiyomi; Sugimoto, Yoshikazu; Kikuchi, Makoto; Yusa, Hideo.

    1979-01-01

    Purpose: To obtain solidified radioactive wastes at high packing density by packing radioactive waste pellets in a container and then packing and curing a thermosetting resin therein. Method: Radioactive liquid wastes are dried into power and subjected to compression molding. The pellets thus obtained are supplied in a predetermined amount from the hopper to the inside of a drum can. Then, thermosetting plastic and a curing agent are filled in the drum can. Gas between the pellets is completely expelled by the intrusion of the thermosetting resin and the curing agent among the pellets. Thereafter, the drum can is heated by a heater and curing is effected. After the curing, the drum can is sealed. (Kawakami, Y.)

  1. Powder wastes confinement block and manufacturing process of this block

    International Nuclear Information System (INIS)

    Dagot, L.; Brunel, G.

    1996-01-01

    This invention concerns a powder wastes containment block and a manufacturing process of this block. In this block, the waste powder is encapsulated in a thermo hardening polymer as for example an epoxy resin, the encapsulated resin being spread into cement. This block can contain between 45 and 55% in mass of wastes, between 18 and 36% in mass of polymer and between 14 and 32% in mass of cement. Such a containment block can be used for the radioactive wastes storage. (O.M.). 4 refs

  2. Immobilisation of ion exchange resins in cement

    International Nuclear Information System (INIS)

    Howard, C.G.; Jolliffe, C.B.; Lee, D.J.

    1990-02-01

    Over the last seven years, Low Oxidation State Metal Ion reagents (LOMI) have been used to decontaminate the 100 MW(e) Steam Generating Heavy Water Ractor (SGHWR) at Winfrith. The use of these reagents has resulted in a dilute ionic solution containing activation products which are produced by corrosion of metallic components in the reactor. It has been demonstrated that the amount of activity in the solution can be reduced using organic ion exchanger resins. These resins consist of a cross linked polystyrene with sulphonic acid or quaternary ammonium function groups and can be successfully immobilised in blended cement systems. The formulation which has been developed is produced from a 9 to 1 blend of ground granulated blast furnace slag (BFS) and ordinary Portland cement (OPC) containing 28% ion exchange resin in the water saturated form. If 6% Microsilica is added to the blended cement the waste loading can be increased to 36 w/o. (author)

  3. Characterisation of plastic packaging waste for recycling: problems related to current approaches

    DEFF Research Database (Denmark)

    Götze, Ramona; Astrup, Thomas Fruergaard

    2013-01-01

    criteria of recycling processes. A lack of information in current waste characterisation practise on polymer resin composition, black coloured material content and the influence of surface adherent material on physico-chemical characteristics of plastic packaging waste were identified. These shortcomings...... were addressed by a resin type-based sorting analysis and a washing test for plastic packaging material from Danish household waste. Preliminary results show that, for a quarter of the hand sorted material, no resin type could be identified and that Polypropylene and Polyethylene terephthalate were...... the dominating resin types in plastic packaging. The suggested washing procedure caused a decrease of 70% of the ash content of the plastic material. The analysed metals and nutrients were reduced by up to 24%...

  4. Wet oxidative destruction of spent ion-exchange resins using hydrogen peroxide

    International Nuclear Information System (INIS)

    Srinivas, C.; Ramaswamy, M.; Theyyunni, T.K.

    1994-01-01

    Spent organic ion exchange resins are generated in large quantities during the operation of nuclear facilities. Wet oxidation as a mode of treatment of these gel-type ion exchange resins was investigated using H 2 O 2 as oxidant in the presence of CuSO 4 as catalyst. Experiments using commercial samples were conducted at 95-100 deg C under reflux conditions at atmospheric pressure. It was found that the reaction of cation resin with H 2 O 2 was instantaneous whereas with anion resin, there was a lag time. For efficient utilization of the oxidant, low rate of addition of H 2 O 2 , 0.01M concentration of CuSO 4 and neutral pH in mixed resin reactions, were found to be useful. Foaming was noted during reactions involving anion resins. This could be controlled by silicone-based agents. The residual solution left after resin oxidation is aqueous in nature and is expected to contain all the radioactivity originally present in the resin. Preliminary experiments showed that it could be efficiently trapped using available inorganic sorbents. Wet oxidation system offers a simple method of converting organic waste into environmentally acceptable inorganic products. (author). 8 refs., 6 figs., 2 tabs

  5. Immobilization in cement of ion exchange resins from Spanish nuclear reactors

    International Nuclear Information System (INIS)

    Huebra, A.G. de la; Murillo, R.; Ortiz, S.J.

    1990-01-01

    Ion exchange materials used at nuclear power plants can be immobilized in cements less expensive than polymer matrices. Cement solidification of spent ion exchange resins shows swelling and cracking troubles (during setting time, or of storage). The objective of this study was to select the types of cement that produce the best quality on immobilization of three kinds of resins and to set up cement formulations containing the maximum possible loading of resin. Four cements were selected to carried out the study. After a study of hydration-dehydration phenomena of ion exchange resins, a systematic work has been carried out on immobilization. Tests were performed to study compressive strength and underwater stability by changing water/cement ratio and resin/cement ratio. Mixtures made with water, cement and resin only were loaded with 10% by weight dry resin. Mixtures with higher loadings show poor workability. Tests were carried out by adding organic plasticizers and silica products to improve waste loading. Plasticizers reduced water demand and silica products permit the use of more water. Leaching tests have been performed at 40 O C. In conclusion Blast Furnace Slag is the best cement for immobilization of ion exchange resin both bead and powdered form for mechanical strength, stability and leaching

  6. Test procedures for polyester immobilized salt-containing surrogate mixed wastes

    International Nuclear Information System (INIS)

    Biyani, R.K.; Hendrickson, D.W.

    1997-01-01

    These test procedures are written to meet the procedural needs of the Test Plan for immobilization of salt containing surrogate mixed waste using polymer resins, HNF-SD-RE-TP-026 and to ensure adequacy of conduct and collection of samples and data. This testing will demonstrate the use of four different polyester vinyl ester resins in the solidification of surrogate liquid and dry wastes, similar to some mixed wastes generated by DOE operations

  7. Produced water treatment using polymeric resins; Resinas polimericas para tratamento da agua produzida

    Energy Technology Data Exchange (ETDEWEB)

    Louvisse, Ana Maria Travalloni; Freire, Norma de Oliveira [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil). Centro de Pesquisas; Queiros, Yure Gomes de Carvalho; Silva, Carla Michele Frota da; Barros, Cintia Chagas; Lucas, Elizabeth Fernandes [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil). Inst. de Macromoleculas

    2008-07-01

    The treatment and disposal of oily waters from oil production and refining processes represent an important technological pass for attendance of the ambient legislation and to the politics of the Cia of search of the environmental excellence. The reuse or reinjection of the co-produced water has been considered an interesting strategical alternative, however, the water quality requirement for these processes demands a pretreatment step, considering the exit effluent from primary and secondary treatment processes currently used. This work presents resulted of the polymeric resin development for tertiary treatment of oily waters with low oil and grasses content and fine solids, including the dissolved fraction. These resins can adsorb, in reversible form, organic components. Its high adsorption capacity is determined by the polarity, superficial area, range of porosity and a wide distribution of particle size and pore. Another important characteristic is the possibility of its use in some cycles of work due to its weak forces of interaction between the contaminants and the surface of the resin. Regeneration can be carried through of diverse forms, including the use of solvent, with or without the variation of temperature and/or pH. The gotten results indicated a high resin adsorption capacity, with concentrated oily water treatment volume 10,000 times the volume of resin used. (author)

  8. Method for processing powdery radioactive wastes

    International Nuclear Information System (INIS)

    Yasumura, Keijiro; Matsuura, Hiroyuki; Tomita, Toshihide; Nakayama, Yasuyuki.

    1978-01-01

    Purpose: To solidify radioactive wastes with ease and safety at a high reaction speed but with no boiling by impregnating the radioactive wastes with chlorostyrene. Method: Beads-like dried ion exchange resin, powdery ion exchange resin, filter sludges, concentrated dried waste liquor or the like are mixed or impregnated with a chlorostyrene monomer dissolving therein a polymerization initiator such as methyl ethyl ketone peroxide and benzoyl peroxide. Mixed or impregnated products are polymerized to solid after a predetermined of time through curing reaction to produce solidified radioactive wastes. Since inflammable materials are used, this process has a high safety. About 70% wastes can be incorporated. The solidified products have a strength as high as 300 - 400 kg/cm 3 and are suitable to ocean disposal. The products have a greater radioactive resistance than other plastic solidification products. (Seki, T.)

  9. Factorial experimental design for recovering heavy metals from sludge with ion-exchange resin

    International Nuclear Information System (INIS)

    Lee, I.H.; Kuan, Y.-C.; Chern, J.-M.

    2006-01-01

    Wastewaters containing heavy metals are usually treated by chemical precipitation method in Taiwan. This method can remove heavy metals form wastewaters efficiently, but the resultant heavy metal sludge is classified as hazardous solid waste and becomes another environmental problem. If we can remove heavy metals from sludge, it becomes non-hazardous waste and the treatment cost can be greatly reduced. This study aims at using ion-exchange resin to remove heavy metals such as copper, zinc, cadmium, and chromium from sludge generated by a PCB manufacturing plant. Factorial experimental design methodology was used to study the heavy metal removal efficiency. The total metal concentrations in the sludge, resin, and solution phases were measured respectively after 30 min reaction with varying leaching agents (citric acid and nitric acid); ion-exchange resins (Amberlite IRC-718 and IR-120), and temperatures (50 and 70 deg. C). The experimental results and statistical analysis show that a stronger leaching acid and a higher temperature both favor lower heavy metal residues in the sludge. Two-factors and even three-factor interaction effects on the heavy metal sorption in the resin phase are not negligible. The ion-exchange resin plays an important role in the sludge extraction or metal recovery. Empirical regression models were also obtained and used to predict the heavy metal profiles with satisfactory results

  10. Biogenic glutamic acid-based resin: Its synthesis and application in the removal of cobalt(II)

    International Nuclear Information System (INIS)

    Jamiu, Zakariyah A.; Saleh, Tawfik A.; Ali, Shaikh A.

    2017-01-01

    Highlights: • A novel resin embedded with metal chelating glutamic acid was synthesized. • The biogenic amino acid residues imparted remarkable efficacy to remove Co(II). • The resin showed excellent ability to remove various metals from wastewater. - Abstract: Inexpensive biogenic glutamic acid has been utilized to synthesize a cross-linked dianionic polyelectrolyte (CDAP) containing metal chelating ligands. Cycloterpolymerization, using azoisobutyronitrile as an initiator, of N,N-diallylglutamic acid hydrochloride, sulfur dioxide and a cross-linker afforded a pH-responsive cross-linked polyzwitterionic acid (CPZA) which upon basification with NaOH was converted into CDAP. The new resin, characterized by a multitude of spectroscopic techniques as well as Scanning Electron Microscopy (SEM) and Brunauer–Emmett–Teller (BET) analyses, was evaluated for the removal of Co(II) as a model case under different conditions. The adsorption capacity of 137 mg g"−"1 does indeed make the resin as one of the most effective sorbents in recent times. The resin leverages its cheap natural source and ease of regeneration in combination with its high and fast uptake capacities to offer a great promise for wastewater treatment. The resin has demonstrated remarkable efficiency in removing toxic metal ions including arsenic from a wastewater sample.

  11. Biogenic glutamic acid-based resin: Its synthesis and application in the removal of cobalt(II)

    Energy Technology Data Exchange (ETDEWEB)

    Jamiu, Zakariyah A.; Saleh, Tawfik A.; Ali, Shaikh A., E-mail: shaikh@kfupm.edu.sa

    2017-04-05

    Highlights: • A novel resin embedded with metal chelating glutamic acid was synthesized. • The biogenic amino acid residues imparted remarkable efficacy to remove Co(II). • The resin showed excellent ability to remove various metals from wastewater. - Abstract: Inexpensive biogenic glutamic acid has been utilized to synthesize a cross-linked dianionic polyelectrolyte (CDAP) containing metal chelating ligands. Cycloterpolymerization, using azoisobutyronitrile as an initiator, of N,N-diallylglutamic acid hydrochloride, sulfur dioxide and a cross-linker afforded a pH-responsive cross-linked polyzwitterionic acid (CPZA) which upon basification with NaOH was converted into CDAP. The new resin, characterized by a multitude of spectroscopic techniques as well as Scanning Electron Microscopy (SEM) and Brunauer–Emmett–Teller (BET) analyses, was evaluated for the removal of Co(II) as a model case under different conditions. The adsorption capacity of 137 mg g{sup −1} does indeed make the resin as one of the most effective sorbents in recent times. The resin leverages its cheap natural source and ease of regeneration in combination with its high and fast uptake capacities to offer a great promise for wastewater treatment. The resin has demonstrated remarkable efficiency in removing toxic metal ions including arsenic from a wastewater sample.

  12. Method of processing radioactive liquid wastes

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  13. Polyethylene solidification of low-level wastes

    International Nuclear Information System (INIS)

    Kalb, P.D.; Colombo, P.

    1985-02-01

    This topical report describes the results of an investigation on the solidification of low-level radioactive waste in polyethylene. Waste streams selected for this study included those which result from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those which remain problematic for solidification using contemporary agents (ion exchange resins). Four types of commercially available low-density polyethylenes were employed which encompass a range of processing and property characteristics. Process development studies were conducted to ascertain optimal process control parameters for successful solidification. Maximum waste loadings were determined for each waste and polyethylene type. Property evaluation testing was performed on laboratory-scale specimens to assess the potential behavior of actual waste forms in a disposal environment. Waste form property tests included water immersion, deformation under compressive load, thermal cycling and radionuclide leaching. Recommended waste loadings of 70 wt % sodium sulfate, 50 wt % boric acid, 40 wt % incinerator ash, and 30 wt % ion exchange resins, which are based on process control and waste form performance considerations are reported. 37 refs., 33 figs., 22 tabs

  14. Application of advanced oxidative process in treatment radioactive waste

    International Nuclear Information System (INIS)

    Kim, Catia; Sakata, Solange K.; Ferreira, Rafael V.P.; Marumo, Julio T.

    2009-01-01

    The ion exchange resin is used in the water purification system in both nuclear research and power reactors. Combined with active carbon, the resin removes dissolved elements from water when the nuclear reactor is operating. After its consumption, it becomes a special type of radioactive waste. The usual treatment to this type of waste is the immobilization with Portland cement, which is simple and low cost. However, its low capacity of immobilization and the increase volume of waste have been the challenges. The development of new technologies capable of destroying this waste completely by increasing its solidification is the main target due to the possibility of both volume and cost reduction. The objective of this work was to evaluate ion exchange resin degradation by Advanced Oxidative Process using Fenton's Reagent (H 2 O 2 / Fe +2 ) in different concentration and temperatures. One advantage of this process is that all additional organic compounds or inorganic solids produced are oxidized easily. The degradation experiments were conducted with IRA-400 resin and Fenton's Reagents, varying the H 2O 2 concentration (30% e 50%) and heat temperature (25, 60 and 100 deg C). The resin degradation was confirmed by the presence of BaCO 3 as a white precipitate resulting from the reaction between the Ba(OH) 2 and the CO 2 from the resin degradation. All experiments run in duplicate. Higher degradation was observed with Fenton's Reagent (Fe +2 /H 2 O 2 30%) at 100 deg C after 2 hours. (author)

  15. Method of solidifying radioactive wastes

    International Nuclear Information System (INIS)

    Tomita, Toshihide; Minami, Yuji; Matsuura, Hiroyuki

    1984-01-01

    Purpose: To enable complete curing even when radioactive wastes contain those materials hindering the curing reaction, for example, copper hydroxide. Method: After admixing an alkaline substance to radioactive concentrated liquid wastes containing copper hydroxide or other amphoteric substances, they are dried, powderized and then cured with thermosetting resins. The thermosetting resins usable herein include, for example, those prepared by mixing an unsaturated polyester with a monomer such as styrene. When a polymerization initiator such as methyl ethyl ketone peroxide and a polymerization promotor are added to the mixture, it takes places curing reaction at normal temperature. Suitable alkaline substances usable herein are those which are insoluble to the liquid wastes and do not change the chemical form under heating and drying. (Yoshihara, H.)

  16. Long-term properties of bituminized waste products

    International Nuclear Information System (INIS)

    Snellman, M.; Valkiainen, M.

    1985-10-01

    This study is a survey of the factors of importance for the long term behaviour of bituminized ion exchange resins. Phenomena occurring in bituminized waste products affected by the treatment, storage and disposal are identified. Test methods have been developed for characterizing product properties, which are important for the long-term behaviour. The long-term properties of bituminized ion-exchange resins are studied in a repository environment with access of water equilibrated with concrete. In these circumstances the most important properties are related to the interactions of bituminized waste with the surrounding barriers. The most important phenomena are water uptake due to rehydration of the resins and subsequent swelling of the product

  17. Long-term properties of bituminized waste products

    International Nuclear Information System (INIS)

    Snellman, M.; Valkiainen, M.

    1985-10-01

    This study is a survey of the factors of importance for long term behaviour of bituminized ion exchange resins. Phenomena occurring in bituminized waste products affected by the treatment, storage and disposal are identified. Test methods have been developed for characterizing product properties, which are important for the long-term behaviour. The long-term properties of bituminized ion-exchange resins are studied in a repository environment with access of water equilibrated with concrete. In these circumstances the most important properties are ralated to the interactions of bituminized waste with the surrounding barriers. The most important phenomena are water uptake due to rehydration of the resins and subsequent swelling of the product. (author)

  18. Design of Ion-Exchange Resins Through EDTA and DTPA Modified Ligands

    Directory of Open Access Journals (Sweden)

    2014-07-01

    Catechol, resorcinol, and their admixtures with EDTA and DTPA moieties were converted into polymeric resins by alkaline polycondensation with formaldehyde. The resins were characterized by FTIR spectroscopy, elemental analysis, ion-exchange capacity, and distribution coefficient (D for heavy metal and radionuclide such as Cs and Sr. 137Cs and 90Sr constitutes a major source of heat in nuclear waste streams and in regards to recent nuclear event their remediation in complex solution – sea water - represent an important issue.

  19. Adsorption, regeneration and reuse of activated carbon for elimination of COD from municipal waste water. Final report. Untersuchungen zur Aktivkohle-Adsorption, -Regeneration und Wiederverwendung beim Einsatz zur Elimination des CSB aus kommunalem Abwasser. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Sekoulov, I.; Holst, H.J.

    1994-02-01

    One practicability to reduce COD-loadings of secondary effluents is the activated carbon fixed bed-process. In this project activated carbons (GAC) of different raw products (peat, wood, coal) were loaded with filtrated municipal waste water (mean COD-concentration: 60 mg CSB/L). After the break-through of adsorption-colums, the carbon was regenerated by means of superheated steam (optimal conditions: [theta]=300 C, pressure=2 bar, thermal energie=10 KW). Regeneration experiments demonstrated, that the results of steam-treatment of GAC is independent in respects to the preloading of activated carbons and the raw products. The mass of COD, detected in the condensed (liquid) steam-phase, always was in quantity about 10% of the further COD adsorbed to the GAC. Gaseous reaction products could not evaluated by means of the experimental design. But former perculation cycles of steam-treated carbons and measurments of the specific area (BET-method) indicated, that the adsorption-capacity of regenerated GAC seams to be in the same quantity, than unloaded carbons. Measurements (GC/MS) of single organic components showed, that humic-substances are catalysed at steam temperatures about 300 C. It seams, that chlorinated hydrocarbons could formed by means of this regeneration method. BOD experiments demonstrated, that substances of the steam-treatment could be biological degraded about 50% to 60%. (orig.)

  20. Liquid and Gaseous Waste Operations Department annual operating report, CY 1992

    Energy Technology Data Exchange (ETDEWEB)

    Gillespie, M.A.; Maddox, J.J.; Scott, C.B.

    1993-03-01

    A total of 6.05 x 10{sup 7} gal of liquid waste was decontaminated by the Process Waste Treatment Plant (PWTP) ion exchange system during CY 1992. This averaged to 115 gpm throughout the year. When necessary, a wastewater sidestream of 50--80 gpm was treated through the use of a natural zeolite treatment system. An additional 8.00 x 10{sup 6} gal (average of 15 gpm throughout the year) were treated by the zeolite system. Therefore, the average total flow treated at the PWTP for CY 1992 was 130 gpm. In mid-June, the zeolite system was repiped to allow it the capability to treat the ion exchange system`s discharge due to rising Cs problems in the wastewater. While being used to treat the ion exchange system`s discharge, it cannot treat a sidestream of wastewater. During the year, the regeneration of the cation exchange resins resulted in the generation of 7.83 x 10{sup 3} gal of liquid low-level waste (LLLW) concentrate and 1.15 x 10{sup 4} gal of LLLW evaporator feed. The head-end softening process (precipitation/clarification) generated 604 drums (4.40 x 10{sup 3} ft{sup 3}) of solid low-level waste sludge. The zeolite treatment system generated approximately 8.40 x 10{sup 2} ft{sup 3} of spent zeolite resin, which was turned over to the Solid Waste Operations Department for disposal. See Table 1 for a monthly summary of activities at the PWTP. Figures 1, 2, 3, and 4 show a comparison of operations at the PWTP in 1992 with previous years. Figure 5 shows a comparison of annual rainfall at Oak Ridge National Laboratory (ORNL) since 1987. A total of 1.55 x 10{sup 8} gal of liquid waste (average of 294 gpm throughout the year) was treated at the Nonradiological Wastewater Treatment Plant (NRWTP). Of this amount, 1.40 x 10{sup 7} gal were treated by the precipitation/clarification process for removal of heavy metals. Twenty-five boxes (1.60 x 10{sup 3} ft{sup 3}) of solid sludge generated by the precipitation/clarification process were removed from the filter press room.

  1. Liquid and Gaseous Waste Operations Department annual operating report, CY 1992

    Energy Technology Data Exchange (ETDEWEB)

    Gillespie, M.A.; Maddox, J.J.; Scott, C.B.

    1993-03-01

    A total of 6.05 x 10[sup 7] gal of liquid waste was decontaminated by the Process Waste Treatment Plant (PWTP) ion exchange system during CY 1992. This averaged to 115 gpm throughout the year. When necessary, a wastewater sidestream of 50--80 gpm was treated through the use of a natural zeolite treatment system. An additional 8.00 x 10[sup 6] gal (average of 15 gpm throughout the year) were treated by the zeolite system. Therefore, the average total flow treated at the PWTP for CY 1992 was 130 gpm. In mid-June, the zeolite system was repiped to allow it the capability to treat the ion exchange system's discharge due to rising Cs problems in the wastewater. While being used to treat the ion exchange system's discharge, it cannot treat a sidestream of wastewater. During the year, the regeneration of the cation exchange resins resulted in the generation of 7.83 x 10[sup 3] gal of liquid low-level waste (LLLW) concentrate and 1.15 x 10[sup 4] gal of LLLW evaporator feed. The head-end softening process (precipitation/clarification) generated 604 drums (4.40 x 10[sup 3] ft[sup 3]) of solid low-level waste sludge. The zeolite treatment system generated approximately 8.40 x 10[sup 2] ft[sup 3] of spent zeolite resin, which was turned over to the Solid Waste Operations Department for disposal. See Table 1 for a monthly summary of activities at the PWTP. Figures 1, 2, 3, and 4 show a comparison of operations at the PWTP in 1992 with previous years. Figure 5 shows a comparison of annual rainfall at Oak Ridge National Laboratory (ORNL) since 1987. A total of 1.55 x 10[sup 8] gal of liquid waste (average of 294 gpm throughout the year) was treated at the Nonradiological Wastewater Treatment Plant (NRWTP). Of this amount, 1.40 x 10[sup 7] gal were treated by the precipitation/clarification process for removal of heavy metals. Twenty-five boxes (1.60 x 10[sup 3] ft[sup 3]) of solid sludge generated by the precipitation/clarification process were removed from the filter

  2. Liquid and Gaseous Waste Operations Department annual operating report, CY 1992

    International Nuclear Information System (INIS)

    Gillespie, M.A.; Maddox, J.J.; Scott, C.B.

    1993-03-01

    A total of 6.05 x 10 7 gal of liquid waste was decontaminated by the Process Waste Treatment Plant (PWTP) ion exchange system during CY 1992. This averaged to 115 gpm throughout the year. When necessary, a wastewater sidestream of 50--80 gpm was treated through the use of a natural zeolite treatment system. An additional 8.00 x 10 6 gal (average of 15 gpm throughout the year) were treated by the zeolite system. Therefore, the average total flow treated at the PWTP for CY 1992 was 130 gpm. In mid-June, the zeolite system was repiped to allow it the capability to treat the ion exchange system's discharge due to rising Cs problems in the wastewater. While being used to treat the ion exchange system's discharge, it cannot treat a sidestream of wastewater. During the year, the regeneration of the cation exchange resins resulted in the generation of 7.83 x 10 3 gal of liquid low-level waste (LLLW) concentrate and 1.15 x 10 4 gal of LLLW evaporator feed. The head-end softening process (precipitation/clarification) generated 604 drums (4.40 x 10 3 ft 3 ) of solid low-level waste sludge. The zeolite treatment system generated approximately 8.40 x 10 2 ft 3 of spent zeolite resin, which was turned over to the Solid Waste Operations Department for disposal. See Table 1 for a monthly summary of activities at the PWTP. Figures 1, 2, 3, and 4 show a comparison of operations at the PWTP in 1992 with previous years. Figure 5 shows a comparison of annual rainfall at Oak Ridge National Laboratory (ORNL) since 1987. A total of 1.55 x 10 8 gal of liquid waste (average of 294 gpm throughout the year) was treated at the Nonradiological Wastewater Treatment Plant (NRWTP). Of this amount, 1.40 x 10 7 gal were treated by the precipitation/clarification process for removal of heavy metals. Twenty-five boxes (1.60 x 10 3 ft 3 ) of solid sludge generated by the precipitation/clarification process were removed from the filter press room

  3. Evaluation of new macroporous resins for the removal of uranium and plutonium from waste streams

    International Nuclear Information System (INIS)

    Koenst, J.W.; Herald, W.R.

    1976-01-01

    Organic ion exchange resins were evaluated for 238 Pu(IV), 238 Pu(VI), and 233 U(VI) removal from water. The capacity of the resins and equilibrium coefficients (Kd) were compared with each other and to bone char--an inorganic adsorbent consisting of hydroxyapatite (HAP) for which data is available. Bone char gave the best results for the removal of 238 Pu(IV), Amberlite XE279 (one of the new macroporous resins) gave the best results for 238 Pu(VI), and another macroporous resin, Dowex-MSA-1, gave good results for 233 U(VI). Kd values were shown to be a function of pH

  4. Liquid waste processing at Comanche Peak

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  5. SETH 200: new mobile unit for spent ion-exchange resins embedding in polymers

    International Nuclear Information System (INIS)

    de Buzonniere, A.; Raibaud, J.; Augustin, X.

    1985-01-01

    The CEA embedding process for low- and medium-activity waste in thermosetting resins (polyester or epoxy) has been used industrially. Recent developments (elimination of chemical pretreatment thanks to a new epoxy formulation and technological breakthroughs in the operating techniques) have greatly increased the potential of the process and have allowed with Technicatome's industrial experience, the elaboration of a new mobile unit easily operated and very competitive, particularly in spent resin processing

  6. Flowsheet Validation For The Permanganate Digestion Of REILLEX(trademark) HPQ Anion Resin

    International Nuclear Information System (INIS)

    Kyser, E.

    2009-01-01

    The flowsheet for the digestion of Reillex(trademark) HPQ was validated both under the traditional alkaline conditions and under strongly acidic conditions. Due to difficulty in performing a pH adjustment in the large tank where this flowsheet must be performed, the recommended digestion conditions were changed from pH 8-10 to 8 M HNO 3 . Thus, no pH adjustment of the solution is required prior to performing the permanganate addition and digestion and the need to sample the digestion tank to confirm appropriate pH range for digestion may be avoided. Neutralization of the acidic digestion solution will be performed after completion of the resin digestion cycle. The amount of permanganate required for this type of resin (Reillex(trademark) HPQ) was increased from 1 kg/L resin to 4 kg/L resin to reduce the amount of residual resin solids to a minimal amount ( 2 ) solids (1.71 kg/L resin) and involves the generation of a significant liquid volume due to the low solubility of permanganate. However, since only two batches of resin (40 L each) are expected to be digested, the total waste generated is limited.

  7. Degradation of ion spent resin using the Fenton's reagent

    International Nuclear Information System (INIS)

    Araujo, Leandro Goulart de

    2013-01-01

    The most common method for spent radioactive ion exchange resin treatment is its immobilization in cement, which reduces the radionuclides release into the environment. Although this method is efficient, it increases considerably the final volume of the waste due to the low incorporation capacity. The objective of this work was to develop a degradation method of spent resins arising from the nuclear research reactor located at the Nuclear and Energy Research Institute (IPEN-CNEN/SP), using an Advanced Oxidation Process (AOP) with Fenton's reagents. This method would allow a higher incorporation in cement. Three different resins were evaluated: cationic, anionic and a mixture of both resins. The reactions were conducted varying the catalyst concentration (25, 50, 100 and 150 mM), the volume of hydrogen peroxide (320 to 460 mL), and three different temperatures, 50, 60 and 70 deg C. Degradation of about 98% was achieved using a 50 mM catalyst solution and 330 mL of hydrogen peroxide solution. The most efficient temperature was 60 deg C. (author)

  8. Acid decomposition processing system for radioactive wastes

    International Nuclear Information System (INIS)

    Oomine, Toshimitsu.

    1984-01-01

    Purpose: To perform plutonium recovery at a low energy consumption irrespective of the plutonium density within the wastes. Method: In a decomposing and volume-reducing device for combustible or less combustible wastes containing transuranic elements using an acid, the wastes are in contact with nitric acid before feeding to a reactor. Then, the transuranic elements are transferred into the nitric acid, which is then in contact with ion exchange resins. After adsorbing the transuranic elements to the ion exchange resins, the nitric acid removed with the transuranic elements is caused to flow into a reaction vessel or heating vessel and used as a decomposing and oxidizing agent. (Seki, T.)

  9. Improvement of incineration efficiency of spent ion exchange resins on the incinerator at nuclear power plants. Manufacturing the solids of the resins mixed with paraffin wax and their incinerating test results on actual incinerator

    International Nuclear Information System (INIS)

    Izumi, Takeshi; Ohtsu, Takashi; Inagawa, Hirofumi; Kawakami, Takashi; Hagiwara, Masahiro; Ino, Takao; Ishiyama, Yuji

    2011-01-01

    In nuclear power plants, ion exchange resins are used at water purification systems such as condensate demineralizers. After usage, used ion exchange resins are stored at plants as low level radioactive wastes. Ion exchange resins contain water and so, those are flame resistant materials. At present, ion exchange resins are incinerated with other inflammable materials at incinerators. Furthermore, ion exchange resins are fine particle beads and are easy to be scattered in all directions, so operators must pay attentions for treatment. Then, we have developed the new solidification system of ion exchange resins with paraffin wax. Ion exchange resins are mixed and extruded with paraffin wax and these solids are enabled to incinerate at existing incinerators. In order to demonstrate this new method, we made the large amount of solids and incinerated them at actual incinerator. From these results, we have estimated to be able to incinerate the solids only at actual incinerator. (author)

  10. Synthesis of Novel Polymeric Resins by Gamma Irradiation for Separation of In(III) ions from Cd(II) in Aqueous Media

    International Nuclear Information System (INIS)

    Massoud, A.; Abou El-Nour, F.; Killa, H.

    2012-01-01

    In this work, Zn(II)polymethacrylates and poly(acrylamide-acrylic acid) were prepared by gamma irradiation polymerization technique of the corresponding monomer at 30 kGy. The polymeric resins were mixed with Indium ions to determine its capacity in aqueous solutions using batch experiment. The adsorption efficiency of obtained polymeric resins toward In(III) and Cd(II) in different experimental conditions was established. Batch and column methods were applied for separation of indium and cadmium. The effects of various eluants such as H 2 SO 4 , NH 4 NO 3 , HNO 3 and HCl on the recovery of both metal ions were studied. The polymeric resins may be regenerated using 3M HCl solutions.

  11. Leaching studies on ion exchange resins immobilized in bitumen matrix

    International Nuclear Information System (INIS)

    Grosche Filho, C.E.; Chandra, U.; Villalobos, J.P.

    1986-01-01

    Many samples of bitumen mixed with ion exchange resin labelled with 134 Cs were prepared to study radionuclide leaching from bitumen waste forms. The resins used in tests were nuclear grade mixed cationic/aniocic bead resins. Different bitumen types were assayed: two distilled and two oxidized bitumens. Leached fractions were collected accordingly to the recommendation of ISO method, with complete exchange of leachant after each sampling. The volume collected for analysis was one liter. Marinelli beckers were used to improve counting efficiency in a Ge(Li) detector. The diffusion coefficients for samples prepared with distilled and oxidized bitumens were obtained. Mathematical models of transport phenomena applied to cylindrical geometry was employed to fit experimental data aiming to explain leaching mechanism and predict the long time behavior of immobilized radionuclides. (M.C.K.) [pt

  12. SuperLig® 639 Resin Performance using 8 Molar LAW Feed

    Energy Technology Data Exchange (ETDEWEB)

    Nash, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Restivo, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hamm, L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-01-15

    All prior testing with SuperLig® 639 has been done with the aqueous concentration of LAW at ~5 M [Na+], where the resin sinks, and can be used in a conventional down-flow column orientation. However, the aqueous LAW stream from the Waste Treatment Plant is expected to be ~8 M [Na+]. The resin would float in this higher density liquid, potentially disrupting the ability to achieve a good decontamination due to poor packing of the resin that leads to channeling. Testing was completed with a higher salt concentration in the feed simulant (7.8 M [Na+]) in an engineering-scale apparatus with two columns, each containing ~0.9 L of resin. Testing of this system used a simulant of the LAW solution, and substituted ReO4 - as a surrogate for TcO4 -. Results were then compared using computer modeling. Bench-scale testing was also performed, and examined an unconstrained resin bed, while engineering-scale tests used both constrained and unconstrained beds in a two-column, lead and lag sequential arrangement.

  13. SuperLig® 639 Resin Performance using 8 Molar LAW Feed

    International Nuclear Information System (INIS)

    Nash, C.; Williams, M.; Restivo, M.; Hamm, L.; McCabe, D.

    2016-01-01

    All prior testing with SuperLig® 639 has been done with the aqueous concentration of LAW at ~5 M [Na+], where the resin sinks, and can be used in a conventional down-flow column orientation. However, the aqueous LAW stream from the Waste Treatment Plant is expected to be ~8 M [Na+]. The resin would float in this higher density liquid, potentially disrupting the ability to achieve a good decontamination due to poor packing of the resin that leads to channeling. Testing was completed with a higher salt concentration in the feed simulant (7.8 M [Na+]) in an engineering-scale apparatus with two columns, each containing ~0.9 L of resin. Testing of this system used a simulant of the LAW solution, and substituted ReO4 - as a surrogate for TcO4 -. Results were then compared using computer modeling. Bench-scale testing was also performed, and examined an unconstrained resin bed, while engineering-scale tests used both constrained and unconstrained beds in a two-column, lead and lag sequential arrangement.

  14. Sampling and characterization of spent exchange resins of Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Varani, Jose L.; Cernadas, D.; Iglesias, Alberto M.; Raffo Calderon, Maria del C.

    2004-01-01

    The present storage facilities for spent resins in Atucha I NPP would be full within the next 5 years, accordingly some tasks are being planned to conditioning these waste and so generate enough storage capacity for the remaining life of the plant. Among this tasks is the characterization of spent resins that has several objectives: to know their chemical and radiochemical composition; the homogeneity of these parameters in the total volume of spent resins; the existence or not of compact zones; the proportion of 'crud'; the breakage degree of the micro-spheres; etc. The first step was to analyse the criterion to follow for sampling resins in the storage deposit of 40 m 3 . In order to take some samples from different points, a special device was required. It had to be introduced closed in the resin bed, opened to take the sample and then closed again to return to the surface. A device used in cereal industry for sampling silos to different depths was modified in its internal capacity for reducing operator dose and increasing the length of rod in order to reaching the bottom of the pit. The device was tested in cold mock up before to taking actual samples. Active resins samples, five in total up to now, were taken from deposit to different depths and kept in lead containers. After analysing the samples, the following average results were extracted: 1.7 x 10 5 Bq/g of Co-60, 9.7 x 10 5 Bq/g of Cs-137 and 774 Bq/g of total alpha, which corresponds to intermediate activity waste. The differences between the values of activity of the different samples are of up to 310 % for Co-60 and of up to 788 % for the Cs-137 what indicates a great inhomogeneity. The direct observation of resin grains, placed in a transparent glass burette, did not demonstrate an important proportion of broken or divided resins. (author)

  15. Cesium ion exchange using actual waste: Column size considerations

    International Nuclear Information System (INIS)

    Brooks, K.P.

    1996-04-01

    It is presently planned to remove cesium from Hanford tank waste supernates and sludge wash solutions using ion exchange. To support the development of a cesium ion exchange process, laboratory experiments produced column breakthrough curves using wastes simulants in 200 mL columns. To verify the validity of the simulant tests, column runs with actual supernatants are being planned. The purpose of these actual waste tests is two-fold. First, the tests will verify that use of the simulant accurately reflects the equilibrium and rate behavior of the resin compared to actual wastes. Batch tests and column tests will be used to compare equilibrium behaviors and rate behaviors, respectively. Second, the tests will assist in clarifying the negative interactions between the actual waste and the ion exchange resin, which cannot be effectively tested with simulant. Such interactions include organic fouling of the resin and salt precipitation in the column. These effects may affect the shape of the column breakthrough curve. The reduction in column size also may change the shape of the curve, making the individual effects even more difficult to sort out. To simplify the evaluation, the changes due to column size must be either understood or eliminated. This report describes the determination of the column size for actual waste testing that best minimizes the effect of scale-down. This evaluation will provide a theoretical basis for the dimensions of the column. Experimental testing is still required before the final decision can be made. This evaluation will be confined to the study of CS-100 and R-F resins with NCAW simulant and to a limited extent DSSF waste simulant. Only the cesium loading phase has been considered

  16. Treatment of liquids of exchange resins washing from RA6

    International Nuclear Information System (INIS)

    Becker, F.L.; Menghini, J.; Rodriguez, D.S.

    2012-01-01

    In this work nanometric magnetite nanoparticles have been synthesized by hydrochemical co-precipitation method to achieve the adsorption and extraction of the contaminant species present in the water coming from the regeneration of exchange resins in Research Reactor RA6. It has been possible to obtain a mean nanoparticle size of 16 nm, necessary to increase the specific surface area of the material and thus, its adsorbent capability. The synthesis parameters and adsorption conditions made it possible to achieve a treatment efficiency superior to 80% for Cs-137 (author)

  17. Ion exchange removal of chromium (iii) from tannery wastes by using a strong acid cation exchange resin amberlite ir-120 h+ and its hybrids

    International Nuclear Information System (INIS)

    Ahmad, T.

    2014-01-01

    A strong acid cation exchange resin Amberlite IR-120 H+ and its hybrids with Mn(OH)/sub 2/, Cu(OH)/sub 2/ and Fe(OH)/sub 3/ are used for the removal of chromium (III) from spent tannery bath. The experimental data give good fits with the Langmuir sorption model. The thermodynamic parameters entropy (delta S), enthalpy (delta H) and free energy (delta G) changes are computed, which reveal that the chromium removal from tannery wastes by ion exchangers is an endothermic, physical sorption and entropically driven process. The rate of sorption is found to increase with the increase of resin dosage, stirring speed and temperature. Different kinetic models such as film diffusion, particle diffusion and Lagergren pseudo first order are used to evaluate the mechanism of the process. It is found that the hybrid ion exchange resins have better removal capacity as compared to the parent ion exchanger. The increase in the removal capacity is found to be in the order of the corresponding PZC values of the hybrid ion exchangers. Further, it is suggested that the higher exchange capacity is the result of Donnan effect and specific adsorption of chromium by the oxides / hydroxides present inside the matrix of the organic cation exchanger. (author)

  18. HTGR fuel development: loading of uranium on carboxylic acid cation-exchange resins using solvent extraction of nitrate

    International Nuclear Information System (INIS)

    Haas, P.A.

    1975-09-01

    The reference fuel kernel for recycle of 233 U to HTGR's (High-Temperature Gas-Cooled Reactors) is prepared by loading carboxylic acid cation-exchange resins with uranium and carbonizing at controlled conditions. The purified 233 UO 2 (NO 3 ) 2 solution from a fuel reprocessing plant contains excess HNO 3 (NO 3 - /U ratio of approximately 2.2). The reference flowsheet for a 233 U recycle fuel facility at Oak Ridge uses solvent extraction of nitrate by a 0.3 M secondary amine in a hydrocarbon diluent to prepare acid-deficient uranyl nitrate. This nitrate extraction, along with resin loading and amine regeneration steps, was demonstrated in 14 runs. No significant operating difficulties were encountered. The process is controlled via in-line pH measurements for the acid-deficient uranyl nitrate solutions. Information was developed on pH values for uranyl nitrate solution vs NO 3 - /U mole ratios, resin loading kinetics, resin drying requirements, and other resin loading process parameters. Calculations made to estimate the capacities of equipment that is geometrically safe with respect to control of nuclear criticality indicate 100 kg/day or more of uranium for single nitrate extraction lines with one continuous resin loading contactor or four batch loading contactors. (auth)

  19. Formulation study on immobilization of spent ion exchange resins in polymer cement

    International Nuclear Information System (INIS)

    Xia Lili; Lin Meiqiong; Bao Liangjin; Fan Xianhua

    2006-01-01

    The aim of this study is to develop a formulation of cement-solidified spent radioactive ion exchange resin form. The solidified form consists of a sort of composite cement, epoxide resin emulsion, and spent ion exchange resins. The composite cement is made up of quick-setting sulphoaluminate cement, silica powder, zeolite, and fly ash in the proportion 1:0.05:0.10:0.05. Sixteen combinations of composite cement, epoxide resin emulsion and mixed anion-cation exchange resins are selected according to a three-factors-four-levels normal design table with the compression strength as the evaluation criterion. The resulted formulation is as follows: the mass ratio of polymer emulsion to composite cement is 0.55:1, the loading of mixed anion-cation exchange resins is 0.3, and the anionic-to-cationic exchange resins ratio is 2:1. The polymer cement solidified forms were tested after 28 d curing for Cs + and Sr 2+ leaching rates, pH and conductivity of the leaching water, and radiation-resistant property in addition to their compressive strength. The measurement results indicate that the performance of thus prepared solidified forms can meet the requirements of the National Standard GB14569.1-93 for near earth's surface disposal of low radioactive waste. (authors)

  20. Study of methods for removing strontium, plutonium, and ruthenium from Savannah River Plant waste supernate

    International Nuclear Information System (INIS)

    Wiley, J.R.

    1976-06-01

    As a part of long-term waste management studies at the Savannah River Laboratory, tests were made to study removal of strontium, plutonium, and ruthenium from simulated and actual waste supernates. Plutonium was sorbed by Duolite ARC-359 ion exchange resin, the same resin that is used to remove cesium from waste supernate. Strontium was removed from supernate by sorption on a chelating resin Chelex 100, or by precipitation as Sr 3 (PO 4 ) 2 . Activities of 137 Cs, 90 Sr, and 238-241 Pu remaining in processed waste supernate should be 1-10 nanocuries of each element per gram of salt. Of the methods that were tested, none was adequate for plant-scale removal of ruthenium

  1. Recovery of transplutonium elements from nuclear reactor waste

    International Nuclear Information System (INIS)

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

    1977-01-01

    A method of separating actinide values from nitric acid waste solutions resulting from reprocessing of irradiated nuclear fuels comprises oxalate precipitation of the major portion of actinide and lanthanide values to provide a trivalent fraction suitable for subsequent actinide/lanthanide partition, exchange of actinide and lanthanide values in the supernate onto a suitable cation exchange resin to provide an intermediate-lived raffinate waste stream substantially free of actinides, and elution of the actinide values from the exchange resin. The eluate is then used to dissolve the trivalent oxalate fraction prior to actinide/lanthanide partition or may be combined with the reprocessing waste stream and recycled. 5 claims, 2 figures

  2. Resin impregnation process for producing a resin-fiber composite

    Science.gov (United States)

    Palmer, Raymond J. (Inventor); Moore, William E. (Inventor)

    1994-01-01

    Process for vacuum impregnation of a dry fiber reinforcement with a curable resin to produce a resin-fiber composite, by drawing a vacuum to permit flow of curable liquid resin into and through a fiber reinforcement to impregnate same and curing the resin-impregnated fiber reinforcement at a sufficient temperature and pressure to effect final curing. Both vacuum and positive pressure, e.g. autoclave pressure, are applied to the dry fiber reinforcement prior to application of heat and prior to any resin flow to compact the dry fiber reinforcement, and produce a resin-fiber composite of reduced weight, thickness and resin content, and improved mechanical properties. Preferably both a vacuum and positive pressure, e.g. autoclave pressure, are also applied during final curing.

  3. Distribution of 14 elements from two solutions simulating Hanford HLW Tank 102-SY (acid-dissolved sludge and acidified supernate) on four cation exchange resins and five anion exchange resins having different functional groups

    International Nuclear Information System (INIS)

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

    1995-01-01

    As part of the Tank Waste Remediation System program at Los Alamos, we evaluated a series of cation exchange and anion exchange resins for their ability to remove hazardous components from radioactive high-level waste (HLW). The anion exchangers were Reillex TM HPQ, a polyvinyl pyridine resin, and four strong-base polystyrene resins having trimethyl, tri ethyl, tri propyl, and tributyl amine as their respective functional groups. The cation exchange resins included Amberlyst TM 15 and Amberlyst tM XN-1010 with sulfonic acid functionality, Duolite TM C-467 with phosphonic acid functionality, and poly functional Diphonix TM with di phosphonic acid, sulfonic acid, and carboxylic acid functionalities. We measured the distributions of 14 elements on these resins from solutions simulating acid-dissolved sludge (pH 0.6) and acidified supernate (pH 3.5) from underground storage tank 102-SY at the Hanford Reservation near Richland, Washington, USA. To these simulants, we added the appropriate radionuclides and used gamma spectrometry to measure fission products (Ce, Cs, Sr, Tc, and Y), actinides (U, Pu, and Am), and matrix elements (Cr, Co, Fe, Mn, Zn, and Zr). For each of the 252 element/resin/solution combinations, distribution coefficients (Kds) were measured for dynamic contact periods of 30 minutes, 2 hours, and 6 hours to obtain information about sorption kinetics from these complex media. Because we measured the sorption of many different elements, the tabulated results indicate which unwanted elements are most likely to interfere with the sorption of elements of special interest. On the basis of these 756 measured Kd values, we conclude that some of the tested resins appear suitable for partitioning hazardous components from Hanford HLW. (author). 10 refs., 11 tabs

  4. Ion Exchange Studies for Removal of Sulfate from Hanford Tank Waste Envelope C (241-AN-107) Using SuperLig 655 Resin

    International Nuclear Information System (INIS)

    Kurath, D.E.; Bontha, J.R.; Blanchard, D.L.; Fiskum, S.K.; Rapko, B.M.

    2000-01-01

    BNFL Inc. is evaluating various pretreatment technologies to mitigate the impacts of sulfate on the LAW vitrification system. One pretreatment technology for separating sulfate from LAW solutions involves the use of SuperLig(reg. sign) 655 (SL-655), a proprietary ion exchange material developed and supplied by IBC Advanced Technologies, Inc., American Fork, UT. This report describes testing of SL-655 with diluted ([Na] approximately 5 M) waste from Hanford Tank 241-AN-107 at Battelle, Pacific Northwest Division. Batch contact studies were conducted from 4 to 96 hours to determine the sulfate distribution coefficient and reaction kinetics. A small-scale ion exchange column test was conducted to evaluate sulfate removal, loading, breakthrough, and elution from the SL-655. In all of these tests, an archived 241-AN-107 tank waste sample (pretreated to remove Cs, Sr, and transuranics elements) was used. The experimental details and results are described in this report. Under the test conditions, SL-655 was found to have no significant ion exchange affinity for sulfate in this matrix. The batch contact study resulted in no measurable difference in the aqueous sulfate concentration following resin contact (K d ∼ 0). The column test also demonstrated SL-655 had no practical affinity for sulfate in the tested matrix. Within experimental error, the sulfate concentration in the column effluent was equal to the concentration in the feed after passing 3 bed volumes of sample through the columns. Furthermore, some, if not all, of the decreased sulfate concentration in these first three column volumes of effluent can be ascribed to mixing and dilution of the 241-AN-107 feed with the interstitial liquid present in the column at the start of the loading cycle. Finally, ICP-AES measurements on the eluate solutions showed the presence of barium as soon as contact with the feed solution is completed. Barium is a metal not detected in the feed solution. Should the loss of barium be

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

    International Nuclear Information System (INIS)

    Kwak, Kyung Kil; Ji, Young Yong

    2010-12-01

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

  6. Maximum Potential Hydrogen Gas Retention in the sRF Resin Ion Exchange Column for the LAWPS Process

    Energy Technology Data Exchange (ETDEWEB)

    Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wells, Beric E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bottenus, Courtney LH [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schonewill, Philip P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2018-01-22

    The Low-Activity Waste Pretreatment System (LAWPS) is being developed to provide treated supernatant liquid from the Hanford tank farms directly to the Low-Activity Waste (LAW) Vitrification Facility at the Hanford Tank Waste Treatment and Immobilization Plant. The design and development of the LAWPS is being conducted by Washington River Protection Solutions, LLC. A key process in LAWPS is the removal of radioactive Cs in ion exchange (IX) columns filled with spherical resorcinol-formaldehyde (sRF) resin. One accident scenario being evaluated is the loss of liquid flow through the sRF resin bed after it has been loaded with radioactive Cs and hydrogen gas is being generated by radiolysis. In normal operations, the generated hydrogen is expected to remain dissolved in the liquid and be continuously removed by liquid flow. For an accident scenario with a loss of flow, hydrogen gas can be retained within the IX column both in the sRF resin and below the bottom screen that supports the resin within the column. The purpose of this report is to summarize calculations that estimate the upper-bound volume of hydrogen gas that can be retained in the column and potentially be released to the headspace of the IX column or to process equipment connected to the IX column and, thus, pose a flammability hazard.

  7. Treatment of spent ion-exchange resins in shaft-type reactor with fuel-plasma source of heating

    International Nuclear Information System (INIS)

    Dmitriev, S. A.; Knyazev, I. A.; Lifanov, F. A.; Polkanov, M. A.; Shvetsov, S. Yu; Savkin, A. E.

    1999-01-01

    The method of high-temperature conditioning the spent radioactive ion-exchange resins in combination with other combustible and incombustible radioactive waste in the plasma shaft furnace with obtaining a crystalline glass-like matrix as a final product has been developed. The method was tested on the pilot plant consisted of ceramic plasma melter, steel water-cooled shaft furnace, lined by fire-resistant concrete, and system of gas purification. The capacity of the furnace was within the limits from 10 to 15 kg/h. The volume and mass reduction factors of treated waste were 36 and 7.6, accordingly. The content of gaseous products of thermal decomposition of a waste at an output of the shaft furnace and properties of obtained slag compound were determined. Based on test results the proposals on creation of compact plasma plant for treatment of mixed radioactive waste including spent ion exchange resins were developed. (author). 3 refs., 2 tabs., 2 figs

  8. Feasibility of Thermoelectric Waste Heat Recovery from Research Reactor

    International Nuclear Information System (INIS)

    Lee, Byunghee

    2015-01-01

    A thermoelectric generator has the most competitive method to regenerate the waste heat from research reactors, because it has no limitation on operating temperature. In addition, since the TEG is a solid energy conversion device converting heat to electricity directly without moving parts, the regenerating power system becomes simple and highly reliable. In this regard, a waste heat recovery using thermoelectric generator (TEG) from 15-MW pool type research reactor is suggested and the feasibility is demonstrated. The producible power from waste heat is estimated with respect to the reactor parameters, and an application of the regenerated power is suggested by performing a safety analysis with the power. The producible power from TEG is estimated with respect to the LMTD of the HX and the required heat exchange area is also calculated. By increasing LMTD from 2 K to 20K, the efficiency and the power increases greatly. Also an application of the power regeneration system is suggested by performing a safety analysis with the system, and comparing the results with reference case without the power regeneration

  9. Liquid radioactive waste processing improvement of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Nery, Renata Wolter dos Reis; Martinez, Aquilino Senra; Monteiro, Jose Luiz Fontes

    2005-01-01

    The study evaluate an inorganic ion exchange to process the low level liquid radwaste of PWR nuclear plants, so that the level of the radioactivity in the effluents and the solid waste produced during the treatment of these liquid radwaste can be reduced. The work compares two types of ion exchange materials, a strong acid cation exchange resin, that is the material typically used to remove radionuclides from PWR nuclear plants wastes, and a mordenite zeolite. These exchange material were used to remove cesium from a synthetic effluent containing only this ion and another effluent containing cesium and cobalt. The breakthrough curves of the zeolite and resin using a fix bed reactor were compared. The results demonstrated that the zeolite is more efficient than the resin in removing cesium from a solution containing cesium and cobalt. The results also showed that a bed combining zeolite and resin can process more volume of an effluent containing cesium and cobalt than a bed resin alone. (author)

  10. Biodegradation testing of TMI-2 EPICOR-II waste forms

    International Nuclear Information System (INIS)

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

    1988-06-01

    ASTM biodegradation tests were conducted on waste forms containing high specific activity ion exchange resins from EPICOR-II prefilters. Those tests were part of a program to test waste forms in accordance with the NRC Branch Technical Position on Waste Form. Small waste forms were manufactured using two different solidification agents, Portland Type I-II cement and vinyl ester-styrene (VES). Ion exchange material was taken from two EPICOR-II prefilters; PF-7, which contained all organic material, and PF-20, which contained organic resins and a layer of inorganic zeolites. Test results showed that the VES waste forms supported microbial growth, while cement waste forms did not support that growth. Growth was also observed adjacent to some VES waste forms. Radiation levels found in the ion exchange resins used in this study were not found to inhibit microbial growth. The extent of degradation of the waste forms could not be determined using the ASTM tests specified by the NRC Branch Technical Position on Waste Form. As a result of this work, a different testing methodology is recommended, which would provide direct verification of waste form capabilities. That methodology would evaluate solidification materials without using the ASTM procedures or subsequent compression testing. The proposed tests would provide exposure to a wide range of microbial species, use appropriately sized specimens, provide for possible use of alternate carbon sources, and extend the test length. Degradation would be determined directly by measuring metabolic activity or specimen weight loss. 16 refs., 15 figs., 3 tabs

  11. Hydrogen sulfide production by sulfate-reducing bacteria utilizing additives eluted from plastic resins.

    Science.gov (United States)

    Tsuchida, Daisuke; Kajihara, Yusuke; Shimidzu, Nobuhiro; Hamamura, Kengo; Nagase, Makoto

    2011-06-01

    In the present study it was demonstrated that organic additives eluted from plastic resins could be utilized as substrates by sulfate-reducing bacteria. Two laboratory-scale experiments, a microcosm experiment and a leaching experiment, were conducted using polyvinyl chloride (PVC) as a model plastic resin. In the former experiment, the conversion of sulfate to sulfide was evident in microcosms that received plasticized PVC as the sole carbon source, but not in those that received PVC homopolymer. Additionally, dissolved organic carbon accumulated only in microcosms that received plasticized PVC, indicating that the dissolved organic carbon originated from additives. In the leaching experiment, phenol and bisphenol A were found in the leached solutions. These results suggest that the disposal of waste plastics in inert waste landfills may result in the production of H(2)S.

  12. The effects of gamma radiation on polymer matrix waste forms

    International Nuclear Information System (INIS)

    Johnson, D.I.; Burnay, S.G.; Phillips, D.C.

    1986-06-01

    A study has been made of the volume and weight changes, mechanical properties, and radiolytic gas production of polymer matrix waste forms during γ irradiation in open containers. The work has been commissioned by the Department of the Environment as part of its radioactive waste management research programme. The materials included polyester, vinyl ester, epoxide and polystyrene resins containing ion exchangers; and polyester and epoxide resins containing a PWR evaporator concentrate. (author)

  13. Corrosion susceptibility of steel drums containing cemented intermediate level nuclear wastes

    Science.gov (United States)

    Duffó, Gustavo S.; Farina, Silvia B.; Schulz, Fátima M.; Marotta, Francesca

    2010-10-01

    Cementation processes are used as immobilization techniques for low or intermediate level radioactive waste for economical and safety reasons and for being a simple operation. In particular, ion-exchange resins commonly used for purification of radioactive liquid waste from nuclear reactors are immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability. Combustible solid radioactive waste can be incinerated and the resulting ashes can also be immobilized before storage. The immobilized resins and ashes are then contained in steel drums that may undergo corrosion depending on the presence of certain contaminants. The work described in this paper was aimed at evaluating the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins and incineration ashes containing different concentrations of aggressive species (mostly chloride and sulphate ions). A special type of specimen was designed to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 1 year. The results show the deleterious effect of chloride on the expected lifespan of the waste containers.

  14. Corrosion susceptibility of steel drums containing cemented intermediate level nuclear wastes

    International Nuclear Information System (INIS)

    Duffo, Gustavo S.; Farina, Silvia B.; Schulz, Fatima M.; Marotta, Francesca

    2010-01-01

    Cementation processes are used as immobilization techniques for low or intermediate level radioactive waste for economical and safety reasons and for being a simple operation. In particular, ion-exchange resins commonly used for purification of radioactive liquid waste from nuclear reactors are immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability. Combustible solid radioactive waste can be incinerated and the resulting ashes can also be immobilized before storage. The immobilized resins and ashes are then contained in steel drums that may undergo corrosion depending on the presence of certain contaminants. The work described in this paper was aimed at evaluating the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins and incineration ashes containing different concentrations of aggressive species (mostly chloride and sulphate ions). A special type of specimen was designed to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 1 year. The results show the deleterious effect of chloride on the expected lifespan of the waste containers.

  15. The effects of aging on compressive strength of low-level radioactive waste form samples

    International Nuclear Information System (INIS)

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

    1996-06-01

    The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program, funded by the US Nuclear Regulatory Commission (NRC), is (a) studying the degradation effects in organic ion-exchange resins caused by radiation, (b) examining the adequacy of test procedures recommended in the Branch Technical Position on Waste Form to meet the requirements of 10 CFR 61 using solidified ion-exchange resins, (c) obtaining performance information on solidified ion-exchange resins in a disposal environment, and (d) determining the condition of liners used to dispose ion-exchange resins. Compressive tests were performed periodically over a 12-year period as part of the Technical Position testing. Results of that compressive testing are presented and discussed. During the study, both portland type I-II cement and Dow vinyl ester-styrene waste form samples were tested. This testing was designed to examine the effects of aging caused by self-irradiation on the compressive strength of the waste forms. Also presented is a brief summary of the results of waste form characterization, which has been conducted in 1986, using tests recommended in the Technical Position on Waste Form. The aging test results are compared to the results of those earlier tests. 14 refs., 52 figs., 5 tabs

  16. Diffusion through composite materials made with thermosetting resins

    International Nuclear Information System (INIS)

    Morin, Bruno.

    1981-08-01

    Medium and low-level radioactive wastes may be coated in a solid matrix mainly made with thermosetting resins: the study of water and cesium migration through composite materials made with thermosetting resins is usefull to compare the water tightness of different coatings. Disks with a thickness of two millimeters were used to measure the water absorption. Diffusion cells including a plane membrane the thickness of which was at least 70μ were used to measure the diffusion of cesium 137. The diffusion coefficient of water in pure thermosetting resins, polyester or epoxyde, is about 10 -9 cm 2 .s -1 ; the diffusion coefficients of cesium in the same materials are about 10 -12 cm 2 .s -1 ; the introduction of solid particles in these polymers generally induces an acceleration of the diffusion process: the diffusion coefficient may reach 10 -8 cm 2 .s -1 . This lost of water-tightness may be reduced either by rendering insoluble the filler mixed to the polymer, or by diminushing the porosity of the interfacial zones by improving the bonding between the polymer and the filler [fr

  17. Method of encapsulating solid radioactive waste material for storage

    International Nuclear Information System (INIS)

    Bunnell, L.R.; Bates, J.L.

    1976-01-01

    High-level radioactive wastes are encapsulated in vitreous carbon for long-term storage by mixing the wastes as finely divided solids with a suitable resin, formed into an appropriate shape and cured. The cured resin is carbonized by heating under a vacuum to form vitreous carbon. The vitreous carbon shapes may be further protected for storage by encasement in a canister containing a low melting temperature matrix material such as aluminum to increase impact resistance and improve heat dissipation. 8 claims

  18. New treatment centers for radioactive waste from Russian designed VVER-reactors

    International Nuclear Information System (INIS)

    Chrubasik, A.

    1997-01-01

    The nuclear power plants using Russian designed VVER-type reactors, were engineered and designed without any wastes treatment facilities. The liquid and solid waste were collected in storage tanks and shelters. After many years of operation, the storage capabilities are exhausted. The treatment of the stored and still generated waste represents a problem of reactor safety and requires a short term solution. NUKEM has been commissioned to design and construct several new treatment centers to remove and process the stored waste. This paper describes the process and lessons learned on the development of this system. The new radioactive waste treatment center (RWTC) includes comprehensive systems to treat both liquid and solid wastes. The process includes: 1) treatment of evaporator concentrates, 2) treatment of ion exchange resins, 3) treatment of solid burnable waste, 4) treatment of liquid burnable waste, 5) treatment of solid decontaminable waste, 6) treatment of solid compactible waste. To treat these waste streams, various separate systems and facilities are needed. Six major facilities are constructed including: 1. A sorting facility with systems for waste segregation. 2. A high-force compactor facility for volume reduction of non-burnable waste. 3. An incinerator facility for destruction of: 1) solid burnable waste, 2) liquid burnable waste, 3) low level radioactive ion exchange resins. 4. A facility for melting of incineration residue. 5. A cementation facility for stabilization of: 1) medium level radioactive ion exchange resins, 2) solid non compactible waste, 3) compacted solid waste. 6. Separation of radionuclides from evaporator concentrates. This presentation will address the facilities, systems, and lessons learned in the development of the new treatment centers. (author)

  19. Process for affixing radioactive contamination on contaminated materials or wastes. Its application

    International Nuclear Information System (INIS)

    Aubouin, Guy; Aude, Georges; Tassigny, Christian de.

    1982-01-01

    The invention concerns a process for affixing radioactive contamination on materials or waste matter in order to ensure that the materials are transferred in complete safety or to package them when their activity is low. Under this process at least one coat of a resin polymerizable at ambient temperature, for example an epoxy resin, a polyester resin, a vinyl resin or a mixture of thermohardening resin and thermoplastic resin, is sprayed on to the contaminated material part by means of an electrostatic gun [fr

  20. Condensate-polisher resin-leakage quantification and resin-transport studies

    International Nuclear Information System (INIS)

    Stauffer, C.C.; Doss, P.L.

    1983-04-01

    The objectives of this program were to: (1) determine the extent of resin leakage from current generation condensate polisher systems, both deep bed and powdered resin design, during cut-in, steady-state and flow transient operation, (2) analyze moisture separator drains and other secondary system samples for resin fragments and (3) document the level of organics in the secondary system. Resin leakage samples were obtained from nine-power stations that have either recirculating steam generators or once through steam generators. Secondary system samples were obtained from steam generator feedwater, recirculating steam generator blowdown and moisture separator drains. Analysis included ultraviolet light examination, SEM/EDX, resin quantification and infrared analysis. Data obtained from the various plants were compared and factors affecting resin leakage were summarized

  1. Proceedings of the Sandia Laboratories workshop on the use of titanate ion exchangers for defense waste management

    International Nuclear Information System (INIS)

    Schwoebel, R.L.; Northrup, C.J.

    1978-01-01

    Abstracts and visual aids from the following talks are presented: removal of radionuclides from Hanford defense waste solutions; waste management programs at Savannah River Plant; application of defense waste decontamination; americium and curium recovery from nuclear waste using inorganic ion exchanger materials; removal of trace 106 Ru in nuclear waste processing; and titanate characterization and consolidation processes. Copies of three memos are included: 90 Sr radiation effects on sodium titanate loaded macroreticular resin; 238 239 Pu content in defense waste; and preparation and physical properties of sodium titanate in ion exchange resin

  2. OPG Western waste management facility resin overpacking project

    International Nuclear Information System (INIS)

    Rae, G.A.; Van de Bospoort, P.; Pearson, S.D.

    2007-01-01

    Liners containing radioactive resins are stored in in-ground containers. Over time, degradation of the liners has occurred and there is potential for eventual leakage. The liners require overpacking in more robust packages to allow for extended storage and final placement in the Deep Geologic Repository. This paper will discuss the equipment design for safe venting, weather protection, radiation shielding, and remote handling of the liners. Alternative considerations and reasoning for final equipment design will be addressed. It will present issues encountered and how they were overcome as well as the logistical overview of the project, including milestones and time tables. (author)

  3. Determination of Na+ and K+ ions in the high-level liquid waste by ion chromatography (IC)

    International Nuclear Information System (INIS)

    Chen Lianzhong; Ma Guilan

    1992-01-01

    The determination of Na + and k + ions in the high-level liquid waste is investigated using ion chromatography. In order to protect the low capacity ion exchange resin in single column IC and remove the transition metal as well as other heavy metal ions that are contained in liquid waste, the pretreatment column with EDTA chelating resin is used. Those impurity metal ions are strongly absorbed by EDTA chelating resin and 100% of Na + and K + ions in the solution are eluted. The ability of the decontamination of EDTA chelating resin is satisfactory. The sample of the high-level liquid waste is diluted appropriately, then an aliquot of the sample is passed through the pretreatment column with EDTA chelating resin, the eluate is analysed by single column ion chromatography. The precision of this method is better than 5% for the determination of Na + and K + ions (at μg· ml -1 level)

  4. Development of volume-reduction system for ion exchange resin using inductively coupled plasma

    International Nuclear Information System (INIS)

    Fujisawa, Morio; Katagiri, Genichi

    2002-01-01

    The spent ion exchange resin generated as radioactive waste in water purifying system at nuclear power stations or related facilities of nuclear power has been stored in the site, and its volume has been increasing year by year. We had developed a full-scale system of IC plasma volume-reduction system for the spent resin, and have performed basic performance test using some samples imitating the spent resin. As the results, the imitation of the resin can be reduced in volume by more than 90% so that the processing performance in actual scale was proved to be effective. In addition, it was clarified that the residuum after volume-reduction process is easy to mix with cement, and solidity containing 30wt% residuum provides high strength of 68 MPa. Therefore, we evaluate the application of this process to stabilization of the disposal to be very effective. (author)

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

    International Nuclear Information System (INIS)

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

    1987-03-01

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

  6. New Anion-Exchange Resins for Improved Separations of Nuclear Materials

    International Nuclear Information System (INIS)

    Bartsch, Richard A.; Barr, Mary E.

    2001-01-01

    Improved separations of nuclear materials will have a significant impact upon a broad range of DOE activities. DOE-EM Focus Areas and Crosscutting Programs have identified improved methods for the extraction and recovery of radioactive metal ions from process, waste, and environmental waters as critical needs for the coming years. We propose to develop multifunctional anion-exchange resins that facilitate anion uptake by carefully controlling the structure of the anion receptor site. Our new ion-exchange resins interface the field of ion-specific chelating ligands with robust, commercial ion-exchange technology to provide materials which exhibit superior selectivity and kinetics of sorption and desorption. The following Focus Areas and Crosscutting Programs have described needs that would be favorably impacted by the new material: Efficient Separations and Processing - radionuclide removal from aqueous phases; Plutonium - Pu, Am or total alpha removal to meet regulatory requirement s before discharge to the environment; Plumes - U and Tc in groundwater, U, Pu, Am, and Tc in soils; Mixed Waste - radionuclide partitioning; High-Level Tank Waste - actinide and Tc removal from supernatants and/or sludges. The basic scientific issues which need to be addressed are actinide complex speciation along with modeling of metal complex/functional site interactions in order to determine optimal binding-site characteristics. Synthesis of multifunctionalized extractants and ion-exchange materials that implement key features of the optimized binding site, and testing of these materials, will provide feedback to the modeling and design activities. Resin materials which actively facilitate the uptake of actinide complexes from solution should display both improved selectivity and kinetic properties. The long-range implications of this research, however, go far beyond the nuclear complex. This new methodology of ''facilitated uptake'' could revolutionize ion-exchange technology

  7. New anion-exchange resins for improved separations of nuclear materials

    International Nuclear Information System (INIS)

    Barr, M.E.; Bartsch, R.A.

    1998-01-01

    'The overall objective of this research is to develop a predictive capability which allows the facile design and implementation of multi-functionalized anion-exchange materials which selectively sorb metal complexes of interest from targeted process, waste, and environmental streams. The basic scientific issues addressed are actinide complex speciation along with modeling of the metal complex/functional-site interactions in order to determine optimal binding-site characteristics. The new ion-exchange resins interface the rapidly developing field of ion-specific chelating ligands with robust, commercial ion-exchange technology. Various Focus Areas and Crosscutting Programs have described needs that would be favorably impacted by the new materials: Efficient Separations and Processing; Plutonium; Plumes; Mixed Waste; High-Level Tank Waste. Sites within the DOE complex which would benefit from the improved anion-exchange technology include Hanford, INEL, Los Alamos, Oak Ridge, and Savannah River. As of April 1998, this report summarizes work after 1.6 years of a 3-year project. The authors technical approach combines empirical testing with theoretical modeling (applied in an iterative mode) in order to determine optimal binding-site characteristics. They determine actinide-complex speciation in specific media, then develop models for the metal complex/functional-site interactions Synthesis and evaluation of multi-functionalized extractants and ion-exchange materials that implement key features of the optimized binding site provide feedback to the modeling and design activities. Resin materials which actively facilitate the uptake of actinide complexes from solution should display both improved selectivity and kinetic properties. The implementation of the bifunctionality concept involves N-derivatization of pyridinium units from a base poly(4-vinylpyridine) resin with a second cationic site such that the two anion-exchange sites are linked by spacer arms of varying

  8. A new low-cost method of reclaiming mixed foundry waste sand based on wet-thermal composite reclamation

    OpenAIRE

    Fan Zitian; Liu Fuchu; Long Wei

    2014-01-01

    A lot of mixed clay-resin waste sand from large-scale iron foundries is discharged every day; so mixed waste sand reclamation in low cost and high quality has a great realistic significance. In the study to investigate the possibility of reusing two types of waste foundry sands, resin bonded sand and clay bonded sand which came from a Chinese casting factory, a new low-cost reclamation method of the mixed foundry waste sand based on the wet-thermal composite reclamation was proposed. The wast...

  9. Experience with intermediate strong base anion resins at three mile island nuclear station

    International Nuclear Information System (INIS)

    Federick, K.H.; Reed, J.G.; Glass, R.M.

    1978-01-01

    A-104 resin appears to have been an excellent replacement for upflow regenerated Stratabeds at Three Mile Island. It has resulted in a 50% increase in operating throughputs, lower maintenance and greater ease of operation for the makeup trains. It should be stressed that differences in raw water characteristics or equipment design might render this experience not applicable to any given specific installation. A thorough evaluation of all aspects of such a conversion is recommended before any changes are attempted. 1 ref

  10. A big picture prospective for wet waste processing management

    International Nuclear Information System (INIS)

    Gibson, J.D.

    1996-01-01

    This paper provides an overview of general observations made relative to the technical and economical considerations being evaluated by many commercial nuclear power plants involving their decision making process for implementation of several new wet waste management technologies. The waste management processes reviewed include the use of, Reverse Osmosis, Non-Precoat Filters, Resin Stripping ampersand Recycling, Evaporation ampersand Calcination (RVR trademark, ROVER trademark ampersand Thermax trademark), Compression Dewatering (PressPak trademark), Incineration (Resin Express trademark), Survey ampersand Free Release (Green Is Clean) and Quantum Catalytic Extraction Processing (QCEP trademark). These waste management processes are reviewed relative to their general advantages and disadvantages associated with the processing of various wet waste streams including: reactor make-up water, floor drain sludges and other liquid waste streams such as boric acid concentrates and steam generator cleaning solutions. A summary of the conclusions generally being derived by most utilities associated with the use of these waste management processes is also provided

  11. Radioactive waste sealing container

    International Nuclear Information System (INIS)

    Tozawa, S.; Kitamura, T.; Sugimoto, S.

    1984-01-01

    A low- to medium-level radioactive waste sealing container is constructed by depositing a foundation coating consisting essentially of zinc, cadmium or a zinc-aluminum alloy over a steel base, then coating an organic synthetic resin paint containing a metal phosphate over the foundation coating, and thereafter coating an acryl resin, epoxy resin, and/or polyurethane paint. The sealing container can consist of a main container body, a lid placed over the main body, and fixing members for clamping and fixing the lid to the main body. Each fixing member may consist of a material obtained by depositing a coating consisting essentially of cadmium or a zinc-aluminum alloy over a steel base

  12. Receipt and processing of RBOF/RRF liquid waste in H-Tank Farm

    International Nuclear Information System (INIS)

    Marra, J.E.

    1994-01-01

    The Receiving Basin for Off-Site Fuels/Resin Regeneration Facility (RBOF/RRF) currently generates approximately 50,000 gallons of wastewater per month. This waste is sent to the 211-H General Purpose (GP) evaporator and/or the 241-H Tank Farm (HTF). The primary criteria for selecting the destination of the waste are solids content and radioactively.The waste is typically sent to the GP evaporator if it has low solids content and low activity. Currently, approximately 70% of the waste water produced at RBOF/RRF meets the criteria for acceptance by the GP evaporator. In June 1993, High Level Waste Engineering opened a Technical Issue (TI) related to processing of RBOF/RRF directly through the 1H Cesium Removal Column (CRC) to the F/H Effluent Treatment Facility (ETF). In March 1994, additional emphasis was placed on this effort after it was determined that the 1H evaporator had a failed tube bundle. As a result, The TI was expanded to include evaluations of methods to increase the acceptance rate of wastewater at the GP (i.e., to ensure that the 70% of RBOF/RRF wastewater that currently meets the GP acceptance criteria is actually processed at the GP). Since March 1994, waste receipts from RBOF/RRF have averaged less than the 30,000 gallons/month allotted in the HLW System Plan. In addition, the RBOF/RRF waste sent to HTF has successfully been processed through the 2H evaporator. Based on this progress, no additional effort should be expended to reduce the amount of RBOF/RRF sent to HTF, either by increasing the criteria for acceptance of RBOF/RRF waste at the GP evaporator or by evaluating alternate treatment options (such as processing through the 1H CRC or installing treatment equipment in the RBOF/RRF)

  13. Extracorporeal adsorption of anti-factor VIII allo-antibodies on randomly functionalized polystyrene resins.

    Science.gov (United States)

    Huguet, Hélène-Céline; Lasne, Dominique; Rothschild, Chantal; Siali, Rosa; Jozefonvicz, Jacqueline

    2004-02-01

    The occurrence of anti-factor VIII (FVIII) allo-antibodies is a severe complication of the treatment of haemophilia A patients, leading to the inhibition of transfused FVIII activity. The effective elimination of these inhibitory antibodies plays a decisive role in the management of affected patients. To achieve this, immunoadsorption devices employing synthetic adsorbers, which selectively eliminate inhibitors, are of interest in the treatment strategy of haemophilia A patients with inhibitors. Adsorbers consisting of polystyrene-based beads substituted with sulphonate and L-tyrosyl methylester groups, which mimic part of epitope of FVIII molecule recognized by inhibitors, exhibit selective binding capacities towards anti-FVIII antibodies. The adsorption of FVIII inhibitors was investigated by simulating an extracorporeal circulation of haemophilic plasma over these functionalized resins. These innovative adsorbers are able to remove around 25% of anti-FVIII antibodies in 15 minutes depending on the plasma tested. Furthermore, they do not modify the amount of essential plasmatic proteins or residual immunoglobulins G. Experiments which were carried out using different plasmas with various inhibitor titres demonstrate a good reproducibility regarding the adsorption capacity of the synthetic resin. The characteristics of adsorption are similar on either native or regenerated resins. Both the purely synthetic nature of the resin and its easy processability demonstrate the real advantages over currently available protocols. This synthetic adsorber is a major technological advance in selective removal of FVIII inhibitory antibodies.

  14. Characteristics of resin floc dispersion of anion and cation exchange resin in precoat filter using powdered ion exchange resin

    Energy Technology Data Exchange (ETDEWEB)

    Adachi, Tetsurou (Nitto Denko Corp., Ibaraki, Osaka (Japan)); Sawa, Toshio; Shindoh, Toshikazu

    1989-09-01

    The filtration performance of mixed filter aid consisting of powdered anion and cation exchange resins used in the precoat filter is closely related to the characteristics of resin floc dispersion. The factors related to resin floc dispersion of anion and cation exchange resin were investigated by measuring the specific settle volume of resin floc as an evaluating index in addition to the measurement of physical, chemical and electrochemical properties of powdered ion exchange resin. The effect of adsorption of iron oxide and polymer electrolyte and of ion exchange were determined. In addition, considered floc dispersion with adsorbing iron oxide, it was assumed that the amount and filling ratio of resin floc were related to summation and multiplication of surface electric charge respectively. An experimental expression was obtained for simulation of the change of specific settle volume of resin floc by particle size, surface area, ion exchange capacity and degree of ionization of the powdered ion exchange resin. (author).

  15. Characteristics of resin floc dispersion of anion and cation exchange resin in precoat filter using powdered ion exchange resin

    International Nuclear Information System (INIS)

    Adachi, Tetsurou; Sawa, Toshio; Shindoh, Toshikazu.

    1989-01-01

    The filtration performance of mixed filter aid consisting of powdered anion and cation exchange resins used in the precoat filter is closely related to the characteristics of resin floc dispersion. The factors related to resin floc dispersion of anion and cation exchange resin were investigated by measuring the specific settle volume of resin floc as an evaluating index in addition to the measurement of physical, chemical and electrochemical properties of powdered ion exchange resin. The effect of adsorption of iron oxide and polymer electrolyte and of ion exchange were determined. In addition, considered floc dispersion with adsorbing iron oxide, it was assumed that the amount and filling ratio of resin floc were related to summation and multiplication of surface electric charge respectively. An experimental expression was obtained for simulation of the change of specific settle volume of resin floc by particle size, surface area, ion exchange capacity and degree of ionization of the powdered ion exchange resin. (author)

  16. Gas Generation Testing of Spherical Resorcinol-Formaldehyde (sRF) Resin

    Energy Technology Data Exchange (ETDEWEB)

    Colburn, Heather A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bryan, Samuel A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Camaioni, Donald M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mahoney, Lenna A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Adami, Susan R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2018-01-19

    This report describes gas generation testing of the spherical resorcinol-formaldehyde (sRF) resin that was conducted to support the technology maturation of the LAWPS facility. The current safety basis for the LAWPS facility is based primarily on two studies that had limited or inconclusive data sets. The two studies indicated a 40% increase in hydrogen generation rate of water (as predicted by the Hu model) with sRF resin over water alone. However, the previous studies did not test the range of conditions (process fluids and temperatures) that are expected in the LAWPS facility. Additionally, the previous studies did not obtain replicate test results or comparable liquid-only control samples. All of the testing described in this report, conducted with water, 0.45M nitric acid, and waste simulants with and without sRF resin, returned hydrogen generation rates that are within the current safety basis for the facility of 1.4 times the Hu model output for water.

  17. A complete process for the treatment of low level radioactive wastes using composite ion exchange resins based on polyurethane foam

    International Nuclear Information System (INIS)

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

    2007-01-01

    The CFC-PU foam and HMO-PU foams have been employed for the removal of radioactive cesium and strontium respectively from actual Cat. III radioactive liquid waste received from reprocessing plant. Batch studies have been carried out in order to optimize the loading of above chemicals. The results of column and batch studies have indicated that there is a good agreement between them. After passing about 1000 bed volumes, the average DF and volume reduction factor obtained were 20 and 200 respectively. A method has been developed for the wet digestion of spent resins in alkaline KMnO 4 medium. The digested foam has been immobilized in cement matrix and the matrices were characterized with respect to compressive strength and leach resistance. (author)

  18. Image resolution influence on determination of resin injection rock mass

    Science.gov (United States)

    Wang, Weixing; Hakami, Eva

    2006-01-01

    In the context of nuclear waste repositories, an important approach to understanding brittle rock mass behavior to integrate new and powerful observational and numerical methods with multi-functional 3-D imaging and visualization techniques. Since 1994, Swedish Nuclear Fuel and Waste Management Company (SKB) have identified the need for a better understanding of radionuclide transport and retention processes in fractured rock. As a cooperation project between Sweden and China, we sampled a number of rock specimens for analyze rock fracture network by optical image technique. The samples are resin injected, in which way; opened fractures can be seen clearly by means of UV (Ultraviolet) light illumination. In the study period, we used different optical focuses to obtain the images from the same samples; we found that Image resolution influences on porosity determination of resin injected rock mass. This paper presents and discusses the six issues based on our research results: (1) Fracture porosity increases as camera focus distance decreases; (2) Porosity increases as illumination increases in resin injected fracture images; (3) To roughly estimate the porosity, the low resolution image can be used; (4) To collect more details of fracture information, the high resolution image is needed; (5) The resolution of image should be determined based on the aim of fracture analysis; (6) To acquire high resolution image, constructing a special illumination (standard) box maybe helpful to avoid light reflection and diffusion.

  19. The management of intermediate level wastes in Sweden

    International Nuclear Information System (INIS)

    Hultgren, Aa.; Thegerstroem, C.

    1980-01-01

    A brief overview of current practices and research in Sweden on the management of intermediate level wastes is given. Intermediate level wastes include spent resins, filters and core components from the six power reactors in operation; radioactive wastes from nuclear fuel development at Studsvik and from non-nuclear applications are a minor contribution. (Auth.)

  20. Method of processing solidification product of radioactive waste

    International Nuclear Information System (INIS)

    Daime, Fumiyoshi.

    1988-01-01

    Purpose: To improve the long-time stability of solidification products by providing solidification products with liquid tightness, gas tightness, abrasion resistance, etc., of the products in the course of the solidification for the treatment of radioactive wastes. Method: The surface of solidification products prepared by mixing solidifying agents with powder or pellets is entirely covered with high molecular polymer such as epoxy resin. The epoxy resin has excellent properties such as radiation-resistance, heat resistance, water proofness and chemical resistance, as well as have satisfactory mechanical properties. This can completely isolate the solidification products of radioactive wastes from the surrounding atmosphere. (Yoshino, Y.)

  1. Ion exchange resins destruction in a stirred supercritical water oxidation reactor

    International Nuclear Information System (INIS)

    Leybros, A.; Roubaud, A.; Guichardon, P.; Boutin, O.

    2010-01-01

    Spent ion exchange resins (IERs) are radioactive process wastes for which there is no satisfactory industrial treatment. Supercritical water oxidation offers a viable treatment alternative to destroy the organic structure of resins, used to remove radioactivity. Up to now, studies carried out in supercritical water for IER destruction showed that degradation rates higher than 99% are difficult to obtain even using a catalyst or a large oxidant excess. In this study, a co-fuel, isopropanol, has been used in order to improve degradation rates by initiating the oxidation reaction and increasing temperature of the reaction medium. Concentrations up to 20 wt% were tested for anionic and cationic resins. Total organic carbon reduction rates higher than 99% were obtained from this process, without the use of a catalyst. The influence of operating parameters such as IERs feed concentration, nature and counterions of exchanged IERs were also studied. (authors)

  2. Purification of radioactive waste oil by a supercritical fluid

    International Nuclear Information System (INIS)

    Yoo, Jaeryong; Sung, Jinhyun; Park, Kwangheon; Kim, Hongdoo; Kim, Hakwon; Lim, Taeyoon; Yim, Sanghak; Yoon, Weonseob

    2006-01-01

    The radioactive waste oil from the nuclear industry is potentially hazardous due to its possibility to contaminate soil and underwater. Pollutants in waste oil are generally radioactive heavy metals or organo-metals. Radioactive waste oils are highly viscous fluids that are similar to used-motor oils. Several processes have been developed to regenerated used motor oil, such as acid clay treatment, chemical addition, vacuum distillation, thermal cracking and hydrofinishing. However, these technologies are difficult to apply to separating radioactive nuclides from radioactive waste oils. In recent years, our laboratory developed a membrane method for the regeneration of used motor oils. We applied supercritical Co2 (scCO2) as a viscosity reducing additive to waste oils at a lower process temperature in order to improve membrane permeability and thus the energy saving. However, the membrane cannot filter the contaminants in radioactive waste oil that are not particles, such as radioactive ions in impurity water in the oil. In this paper, we suggest a method extracting clean oil from the radioactive waste oil rather than filtering by a supercritical fluid. We selected R22, a refrigerant, as a solvent for extraction. R22 has a mild critical point - 96.1 .deg. and 49.9bar. Regeneration of waste oils by extracting clean oil using a supercritical fluid such as R22 is easy to handle and reduce secondary wastes. In this paper, we examine the feasibility of R22 in extracting clean oil from radioactive waste oils

  3. Leach studies on cement-solidified ion exchange resins from decontamination processes at operating nuclear power stations

    International Nuclear Information System (INIS)

    McIsaac, C.V.; Akers, D.W.; McConnell, J.W.; Morcos, N.

    1992-01-01

    The effects of varying pH and leachant compositions on the physical stability and leachability of radionuclides and chelating agents were determined for cement-solidified decontamination ion-exchange resin wastes collected from two operating commercial light water reactors. Small scale waste-form specimens were collected during waste solidifications performed at the Brunswick Steam Electric Plant Unit 1 and at the James A. FitzPatrick Nuclear Power Station. The collected specimens were leach tested, and their compressive strength was measured in accordance with the Nuclear Regulatory Commission's ''Technical Position on Waste Form'' (Revision 1), from the Low-Level Waste Management Branch. Leachates from these studies were analyzed for radionuclides, selected transition metals, and chelating agents to assess the leachability of these waste form constituents. Leachants used for the study were deionized water, simulated seawater, and groundwater compositions similar to those found at Barnwell, South Carolina and Hanford, Washington. Results of this study indicate that initial leachant pH does not affect leachate pH or releases from cement-solidified decontamination ion-exchange resin waste forms. However, differences in leachant composition and the presence of chelating agents may affect the releases of radionuclides and chelating agents. In addition, results from this study indicate that the cumulative releases of radionuclides and chelating agents observed for forms that disintegrated were similar to those for forms that maintained their general physical integrity

  4. Resin Viscosity Influence on Fiber Compaction in Tapered Resin Injection Pultrusion Manufacturing

    Science.gov (United States)

    Masuram, N. B.; Roux, J. A.; Jeswani, A. L.

    2018-06-01

    Viscosity of the liquid resin effects the chemical and mechanical properties of the pultruded composite. In resin injection pultrusion manufacturing the liquid resin is injected into a specially designed tapered injection chamber through the injection slots present on top and bottom of the chamber. The resin is injected at a pressure so as to completely wetout the fiber reinforcements inside the tapered injection chamber. As the resin penetrates through the fibers, the resin also pushes the fibers away from the wall towards the center of chamber causing compaction of the fiber reinforcements. The fibers are squeezed together due to compaction, making resin penetration more difficult; thus higher resin injection pressures are required to efficaciously penetrate through the compacted fibers and achieve complete wetout. The impact of resin viscosity on resin flow, fiber compaction, wetout and on the final product is further discussed. Injection chamber design predominantly effects the resin flow inside the chamber and the minimum injection pressure required to completely wet the fibers. Therefore, a desirable injection chamber design is such that wetout occurs at lower injection pressures and at low internal pressures inside the injection chamber.

  5. Estimation of tritium trapped in the used ion exchange column resins using water as scavenger

    International Nuclear Information System (INIS)

    Kumaravel, S.; Ramakrishna, V.; Nair, B.S.K.; Ganesh, G.; Tripathi, R.M.

    2018-01-01

    Estimation of Tritium trapped in the used resins of Ion exchange (IX) columns apart from the gross beta activity in heavy water systems of Pressurised Heavy Water Reactors (PHWR) is mandatory before their disposal as radioactive waste. The gross beta activity is estimated by Gieger- Muller (GM) counter and by using gamma spectrometer. Accurate estimation of tritium activity of the resin without compromising the counting efficiency is a challenging task because, if a fixed quantity of the resin is directly added to the scintillation solution and counted on a liquid scintillation analyser (LSA), it is prone to interfere with counting efficiency drastically and results in unquantifiable errors and other practical difficulties. In this study a standard technique using light water as scavenger medium to precisely quantify the total activity of tritium trapped in the resin by a systematic approach was carried out

  6. Leaching studies of radionuclides from solidified wastes with thermosetting resin

    International Nuclear Information System (INIS)

    Suzuki, K.; Kuribayashi, H.; Morimitsu, W.; Ono, I.

    1982-01-01

    This paper reports on studies of the leachability of Co-60 and Cs-137 from simulated LWR radwastes solidified with thermosetting resin and evaluates the effects of chemical fixation on leachability. It is concluded that insolubilization by a nickel-ferrocyanide compound offers an effective chemical fixation of these radionuclides and is a recommended pretreating method for radwastes that are to be solidified. 2 figures

  7. Effects of blood contamination on resin-resin bond strength.

    Science.gov (United States)

    Eiriksson, Sigurdur O; Pereira, Patricia N R; Swift, Edward J; Heymann, Harald O; Sigurdsson, Asgeir

    2004-02-01

    Incremental placement and curing of resin composites has been recommended. However, this requires longer operating time, and therefore, increased risk of contamination. The purpose of this study was to evaluate the effects of blood contamination on microtensile bond strengths (microTBS) between resin interfaces and to determine the best decontamination method to re-establish the original resin-resin bond strength. The top surfaces of 64, 4-mm composite blocks (Z-250, Renew, APX, Pertac II) were untreated as the control, or were treated as follows: blood applied and dried on the surface (Treatment 1), blood applied, rinsed, dried (Treatment 2), blood applied, rinsed, and an adhesive applied (Single Bond, One-Step, Clearfil SE, Prompt L-Pop) (Treatment 3). Fresh composite was applied and light-cured in 2-mm increments. After 24 h storage in water, the specimens were sectioned into 0.7-mm thick slabs, trimmed to a cross-sectional area of 1 mm(2), and loaded to failure at a crosshead speed of 1 mm/min using an Instron universal testing machine. Data were analyzed using two-way ANOVA and Fisher's PLSD test (pcontamination resulted in resin-resin bond strengths of only 1.0-13.1 MPa. Rinsing raised bond strengths to over 40 MPa for each material. Use of an adhesive further increased bond strengths except for Pertac II. Rinsing blood from contaminated surfaces increases the resin-resin bond strength significantly and the application of an appropriate adhesive increases the bond strength to control levels.

  8. Effect of pH on the release of radionuclides and chelating agents from cement-solidified decontamination ion-exchange resins collected from operating nuclear power stations

    International Nuclear Information System (INIS)

    McIsaac, C.V.; Akers, D.W.; McConnell, J.W.

    1991-06-01

    Data are presented on the physical stability and leachability of radionuclides and chelating agents from cement-solidified decontamination ion-exchange resin wastes collected from two operating commercial light water reactors. Small-scale waste--form specimens collected during solidifications performed at the Brunswick Steam Electric Plant Unit 1 and at the James A. FitzPatrick Nuclear Power Station were leach-tested and subjected to compressive strength testing in accordance with the Nuclear Regulatory Commission's ''Technical Position on Waste Form'' (Revision 1). Samples of untreated resin waste collected from each solidification vessel before the solidification process were analyzed for concentrations of radionuclides, selected transition metals, and chelating agents to determine the quantities of these chemicals in the waste-form specimens. The chelating agents included oxalic, citric, and picolinic acids. In order to determine the effect of leachant chemical composition and pH on the stability and leachability of the waste forms, waste-form specimens were leached in various leachants. Results of this study indicate that differences in pH do not affect releases from cement-solidified decontamination ion-exchange resin waste forms, but that differences in leachant chemistry and the presence of chelating agents may affect the releases of radionuclides and chelating agents. Also, this study indicates that the cumulative releases of radionuclides and chelating agents are similar for waste- form specimens that decomposed and those that retained their general physical form. 36 refs., 60 figs., 28 tabs

  9. Foam, Foam-resin composite and method of making a foam-resin composite

    Science.gov (United States)

    Cranston, John A. (Inventor); MacArthur, Doug E. (Inventor)

    1995-01-01

    This invention relates to a foam, a foam-resin composite and a method of making foam-resin composites. The foam set forth in this invention comprises a urethane modified polyisocyanurate derived from an aromatic amino polyol and a polyether polyol. In addition to the polyisocyanurate foam, the composite of this invention further contains a resin layer, wherein the resin may be epoxy, bismaleimide, or phenolic resin. Such resins generally require cure or post-cure temperatures of at least 350.degree. F.

  10. Regeneration of cello-oligomers via selective depolymerization of cellulose fibers derived from printed paper wastes.

    Science.gov (United States)

    Voon, Lee Ken; Pang, Suh Cem; Chin, Suk Fun

    2016-05-20

    Cellulose extracted from printed paper wastes were selectively depolymerized under controlled conditions into cello-oligomers of controllable chain lengths via dissolution in an ionic liquid, 1-allyl-3-methylimidazolium chloride (AMIMCl), and in the presence of an acid catalyst, Amberlyst 15DRY. The depolymerization process was optimized against reaction temperature, concentration of acid catalyst, and reaction time. Despite rapid initial depolymerization process, the rate of cellulose depolymerization slowed down gradually upon prolonged reaction time, with 75.0 wt% yield of regenerated cello-oligomers (mean Viscosimetric Degree of Polymerization value of 81) obtained after 40 min. The depolymerization of cellulose fibers at 80 °C appeared to proceed via a second-order kinetic reaction with respect to the catalyst concentration of 0.23 mmol H3O(+). As such, the cellulose depolymerization process could afford some degree of control on the degree of polymerization or chain lengths of cello-oligomers formed. Copyright © 2016 Elsevier Ltd. All rights reserved.

  11. Development of transition metal oxide catalysts for treatment of off-gases released during pyrolysis of organic ion exchange resins

    International Nuclear Information System (INIS)

    Sathi Sasidharan, N.; Deshingkar, D.S.; Wattal, P.K.

    2005-08-01

    The spent IX resin wastes arising from nuclear power plants have high radiation level due to fission product 137 Cesium and activation product 60 Cobalt. The pyrolysis and oxidative pyrolysis processes have potential to minimize final waste form volumes of these wastes. The major difficulty in deploying these processes for treatment of spent IX resins is release of off-gases containing large quantities of aromatic hydrocarbons, amines, sulphur dioxide, hydrogen sulphide, carbonyl sulphide etc. As an alternative to high temperature incineration of the pyrolysis off gases, feasibility of using catalytic combustion at moderate temperatures was investigated in the laboratory. Copper chromite, copper oxide-ceric oxide and vanadium pentaoxide catalysts supported on alumina were prepared and tested for oxidation of styrene monomer, toluene, ethyl benzene and trimethyl amine at 22500 hr -1 space velocity and temperature range of 300 to 500 degC. At temperatures over 475 degC, all three catatyst gave oxidation efficiency of over 97% for these compounds over concentration range of few tens of ppm to few thousands ppm. A composite catalyst bed of three catalysts comprising principally of copper chromite is proposed for treatment of IX resin pyrolysis off-gases. (author)

  12. Field lysimeter studies for performance evaluation of grouted Hanford defense wastes

    International Nuclear Information System (INIS)

    Last, G.V.; Serne, R.J.; LeGore, V.L.

    1995-02-01

    The Grout Waste Test Facility (GWTF) consisted of four large field lysimeters designed to test the leaching and migration rates of grout-solidified low-level radioactive wastes generated by Hanford Site operations. Each lysimeter was an 8-m-deep by 2-media closed-bottom caisson that was placed in the ground such that the uppermost rim remained just above grade. Two of these lysimeters were used; the other two remained empty. The two lysimeters that were used (A-1 and B-1) were backfilled with a two-layer soil profile representative of the proposed grout disposal site. The proposed grout disposal site (termed the Grout Treatment Facility Landfill) is located immediately east of the Hanford Site's 200 East Area. This soil profile consisted of a coarse sand into which the grout waste forms were placed and covered by 4 m of a very fine sand. The A-1 lysimeter was backfilled in March 1985, with a grout-solidified phosphate/sulfate liquid waste from N Reactor decontamination and ion exchange resin regeneration. The B-1 lysimeter was backfilled in September 1985 and received a grout-solidified simulated cladding removal waste representative of waste generated from fuel reprocessing operations at the head end of the Plutonium Uranium Extraction (PUREX) plant. Routine monitoring and leachate collection activities were conducted for over three years, terminating in January 1989. Drainage was collected sporadically between January 1989 and December 1992. Decontamination and decommissioning of these lysimeters during the summer of 1994, confirmed the presence of a 15 to 20-cm-long hairline crack in one of the bottom plate welds. This report discusses the design and construction of the GWTF, presents the routine data collected from this facility through January 1989 and subsequent data collected sporadically between 1989 and 1993, and provides a brief discussion concerning preliminary interpretation of the results

  13. Waste form development program. Annual report, October 1982-September 1983

    International Nuclear Information System (INIS)

    Colombo, P.; Kalb, P.D.; Fuhrmann, M.

    1983-09-01

    This report provides a summary of the work conducted for the Waste Form Development/Test Program at Brookhaven National Laboratory in FY 1983 under the sponsorship of the US Department of Energy's Low-Level Waste Management Program. The primary focus of this work is the investigation of new solidification agents which will provide improved immobilization of low-level radioactive wastes in an efficient, cost-effective manner. A working set of preliminary waste form evaluation criteria which could impact upon the movement of radionuclides in the disposal environment was developed. The selection of potential solidification agents for further investigation is described. Two thermoplastic materials, low-density polyethylene and a modified sulfur cement were chosen as primary candidates for further study. Three waste types were selected for solidification process development and waste form property evaluation studies which represent both new volume reduction wastes (dried evaporator concentrates and incinerator ash) and current problem wastes (ion exchange resins). Preliminary process development scoping studies were conducted to verify the compatibility of selected solidification agents and waste types and the potential for improved solidification. Waste loadings of 60 wt % Na 2 SO 4 , 25 wt % H 3 BO 3 , 25 wt % incinerator ash and 50 wt % dry ion exchange resin were achieved using low density polyethylene as a matrix material. Samples incorporating 65 wt % Na 2 SO 4 , 40 wt % H 3 BO 3 , 20 wt % incinerator ash and 40 wt % dry ion exchange resin were successfully solidified in modified sulfur cement. Additional improvements are expected for both matrix materials as process parameters are optimized. Several preliminary property evaluation studies were performed to provide the basis for an initial assessment of waste form acceptability. These included a two-week water immersion test and compressive load testing

  14. Characterization and disposal of ion exchange resins used in nuclear installations; Caracterizacion y disposicion de resinas de intercambio ionico utilizadas en instalaciones nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Flores E, R.M.; Ortiz O, H.B.; Olguin G, M.T.; Emeterio H, M.; Garcia M, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2006-07-01

    To dispose of an appropriate way the used ion exchange resins so much in the pool water purification systems of the TRIGA Mark III reactor like in the JS6500 gamma irradiator, of the National Institute of Nuclear Research, were carried out a series of analytic nuclear techniques and complementary conventional to those recommended by the ASTM, with the object of to control and to manage 14 lots of worn out resins appropriately. For its were identified the radioactive isotopes, the resins type, the grade of chemical pollution and the physicochemical degradation of the same ones. The lots of resins that didn't contain radioactive isotopes its were regenerated in an usual way, as long as those that if they controlled them they selected options for its final disposition. The first selected option was the extraction method of ion radioactive isotopes, concentrating the elution product by evaporation. As second option it was carried out the resins stabilization damaged by micro-encapsulation by forged to ambient temperature, using an organic polymer. Previous to the immobilization the resins were pretreated by vacuum drying, pulverization and thermal drying, however before carrying out this last, it was carried out a thermal gravimetric analysis to determine the drying conditions of the resins avoiding its chemical decomposition. (Author)

  15. Leaching studies on ion exchange resins immobilized in bitument matrix

    International Nuclear Information System (INIS)

    Grosche Filho, C.E.; Chandra, U.; Villalobos, J.P.

    1987-01-01

    To study radionuclide leaching from bitumen waste forms, many samples of bitumen mixed with ion-exchange resin labelled with 134 Cs were prepared. The resins used in the tests were nuclear grade mixed cationic/anionic bead resins. Different bittumen types were assayed: two destilled and to oxidized bitumens. Laboratory to scale samples, with surface/volume ratio (S/V) = 1, were molded to 5 cm diameter and 10 cm height. The composition of the mixtures were: 30, 40, 50 and 60% by weight of dried resin with bitumen. The leachant was deionized water with a leachant volume to sample surface rario of about 8 cm. Leached fractions were collected according to the recommendation of ISO method, with complete exchange of leachant beckers after each sampling. The volume collected for analysis was one liter. Marinelli were used for counting in a Ge(Li) detector. Up to now, results of 250 days have been accumulated. Samples prepared with distilled bitumen have shown a diffusion coefficient of the order of 10 -14 cm 2 /sec and those prepared with oxidized bitumen yielded a diffusion coefficient of the order of 10 -12 cm 2 /sec. Mathematical models of transport phenomena applied to cylindrical geometry were employed to fit experimental data. (Author) [pt

  16. Characterization of low and medium level radioactive wastes

    International Nuclear Information System (INIS)

    Nomine, J.C.; Tassigny, C. de; Billon, J.

    1983-11-01

    Leaching tests on real wastes embedded in cement, bitumens or resins are realized to study leachability of alpha-emitters or fission products and anion-cation exchange between leachate and embedded materials. Radionuclide distribution is examined by spectrogammametry on cores taken from cemented wastes. Qualitative results concerning degradation of waste blocks embedded in bitumens by bacteria in the ground are given [fr

  17. Aqueous radioactive waste bituminization

    International Nuclear Information System (INIS)

    Williamson, A.S.

    1980-08-01

    The bituminzation of decontamination and ion exchange resin stripping wastes with four grades of asphalt was investigated to determine the effects of asphalt type on the properties of the final products. All waste forms deformed readily under light loads indicating they would flow if not restrained. It was observed in all cases that product leaching rates increased as the hardness of the asphalt used to treat the waste increased. If bituminization is adopted for any Ontario Hydro aqueous radioactive wastes they should be treated with soft asphalt to obtain optimum leaching resistance and mechanical stability during interim storage should be provided by a corrosion resistant container

  18. Immobilization of Ion Exchange radioactive resins of the TRIGA Mark III Nuclear Reactor; Inmovilizacion de resinas de intercambio ionico radiactivas del reactor nuclear TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Martinez, H

    1999-07-01

    In the last decades many countries in the world have taken interest in the use, availability, and final disposal of dangerous wastes in the environment, within these, those dangerous wastes that contain radioactive material. That is why studies have been made on materials used as immobilization agent of radioactive waste that may guarantee its storage for long periods of time under drastic conditions of humidity, temperature change and biodegradation. In mexico, the development of different applications of radioactive material in the industry, medicine and investigation, have generated radioactive waste, sealed and open sources, whose require a special technological development for its management and final disposal. The present work has as a finality to develop the process and define the agglutinating material, bitumen, cement and polyester resin that permits immobilization of resins of Ionic Exchange contaminated by Barium 153, Cesium 137, Europium 152, Cobalt 60 and Manganese 54 generated from the nuclear reactor TRIGA Mark III. Ionic interchange contaminated resin must be immobilized and is analysed under different established tests by the Mexican Official Standard NOM-019-NUCL-1995 {sup L}ow level radioactive wastes package requirements for its near-surface final disposal. Immobilization of ionic interchange contaminated resins must count with the International Standards applicable in this process; in these standards, the following test must be taken in prototype examples: Free-standing water, leachability, compressive strength, biodegradation, radiation stability, thermal stability and burning rate. (Author)

  19. Method for regeneration of electroless nickel plating solution

    Science.gov (United States)

    Eisenmann, E.T.

    1997-03-11

    An electroless nickel(EN)/hypophosphite plating bath is provided employing acetic acid/acetate as a buffer and which is, as a result, capable of perpetual regeneration while avoiding the production of hazardous waste. A regeneration process is provided to process the spent EN plating bath solution. A concentrated starter and replenishment solution is provided for ease of operation of the plating bath. The regeneration process employs a chelating ion exchange system to remove nickel cations from spent EN plating solution. Phosphites are then removed from the solution by precipitation. The nickel cations are removed from the ion exchange system by elution with hypophosphorus acid and the nickel concentration of the eluate adjusted by addition of nickel salt. The treated solution and adjusted eluate are combined, stabilizer added, and the volume of resulting solution reduced by evaporation to form the bath starter and replenishing solution. 1 fig.

  20. Method for regeneration of electroless nickel plating solution

    Science.gov (United States)

    Eisenmann, Erhard T.

    1997-01-01

    An electroless nickel(EN)/hypophosphite plating bath is provided employing acetic acid/acetate as a buffer and which is, as a result, capable of perpetual regeneration while avoiding the production of hazardous waste. A regeneration process is provided to process the spent EN plating bath solution. A concentrated starter and replenishment solution is provided for ease of operation of the plating bath. The regeneration process employs a chelating ion exchange system to remove nickel cations from spent EN plating solution. Phosphites are then removed from the solution by precipitation. The nickel cations are removed from the ion exchange system by elution with hypophosphorous acid and the nickel concentration of the eluate adjusted by addition of nickel salt. The treated solution and adjusted eluate are combined, stabilizer added, and the volume of resulting solution reduced by evaporation to form the bath starter and replenishing solution.

  1. Modified sulfur cement solidification of low-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    1985-10-01

    This topical report describes the results of an investigation on the solidification of low-level radioactive wastes in modified sulfur cement. The work was performed as part of the Waste Form Evaluation Program, sponsored by the US Department of Energy's Low-Level Waste Management Program. Modified sulfur cement is a thermoplastic material developed by the US Bureau of Mines. Processing of waste and binder was accomplished by means of both a single-screw extruder and a dual-action mixing vessel. Waste types selected for this study included those resulting from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those which remain problematic for solidification using contemporary agents (ion exchange resins). Process development studies were conducted to ascertain optimal process control parameters for successful solidification. Maximum waste loadings were determined for each waste type and method of processing. Property evaluation testing was carried out on laboratory scale specimens in order to compare with waste form performance for other potential matrix materials. Waste form property testing included compressive strength, water immersion, thermal cycling and radionuclide leachability. Recommended waste loadings of 40 wt. % sodium sulfate and boric acid salts and 43 wt. % incinerator ash, which are based on processing and performance considerations, are reported. Solidification efficiencies for these waste types represent significant improvements over those of hydraulic cements. Due to poor waste form performance, incorporation of ion exchange resin waste in modified sulfur cement is not recommended.

  2. Modified sulfur cement solidification of low-level wastes

    International Nuclear Information System (INIS)

    1985-10-01

    This topical report describes the results of an investigation on the solidification of low-level radioactive wastes in modified sulfur cement. The work was performed as part of the Waste Form Evaluation Program, sponsored by the US Department of Energy's Low-Level Waste Management Program. Modified sulfur cement is a thermoplastic material developed by the US Bureau of Mines. Processing of waste and binder was accomplished by means of both a single-screw extruder and a dual-action mixing vessel. Waste types selected for this study included those resulting from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those which remain problematic for solidification using contemporary agents (ion exchange resins). Process development studies were conducted to ascertain optimal process control parameters for successful solidification. Maximum waste loadings were determined for each waste type and method of processing. Property evaluation testing was carried out on laboratory scale specimens in order to compare with waste form performance for other potential matrix materials. Waste form property testing included compressive strength, water immersion, thermal cycling and radionuclide leachability. Recommended waste loadings of 40 wt. % sodium sulfate and boric acid salts and 43 wt. % incinerator ash, which are based on processing and performance considerations, are reported. Solidification efficiencies for these waste types represent significant improvements over those of hydraulic cements. Due to poor waste form performance, incorporation of ion exchange resin waste in modified sulfur cement is not recommended

  3. Evaluation of extractants and chelating resins in polishing actinide-contaminated waste streams

    International Nuclear Information System (INIS)

    Schreiber, S.B.; Dunn, S.L.; Yarbro, S.L.

    1991-06-01

    At the Los Alamos National Laboratory Plutonium Facility, anion exchange is used for recovering plutonium from nitric acid solutions. Although this approach recovers >99%, the trace amounts of plutonium and other actinides remaining in the effluent require additional processing. We are doing research to develop a secondary unit operation that can directly polish the effluent so that actinide levels are reduced to below the maximum allowed for facility discharge. We selected solvent extraction, the only unit operation that can meet the stringent process requirements imposed; several carbonyl and phosphoryl extractants were evaluated and their performance characterized. We also investigated various engineering approaches for solvent extraction; the most promising was a chelating resin loaded with extractant. Our research now focuses on the synthesis of malonamides, and our goal is to bond these extractants to a resin matrix. 7 refs., 12 figs., 1 tab

  4. Potential of Electronic Plastic Waste as a Source of Raw Material and Energy Recovery

    International Nuclear Information System (INIS)

    Norazli Othman; Nor Ezlin Ahmad Basri; Lariyah Mohd Sidek

    2009-01-01

    Nowadays, the production of electronic equipment is one of the fastest growing industrial activities in this world. The increase use of plastic in this sector resulted in an increase of electronic plastic waste. Basically, electronic plastic material contains various chemical elements which act as a flame retardant when electronic equipment is operated. In general, the concept of recycling electronic plastic waste should be considered in order to protect the environment. For this purpose, research has been conducted to different resins of electronic plastic waste to identify the potential of electronic plastic waste as a source of raw material and energy recovery. This study was divided into two part for example determination of physical and chemical characteristics of plastic resins and calculation of heating value for plastic resins based on Dulong formula. Results of this research show that the average calorific value of electronic waste is 30,872.42 kJ/ kg (7,375 kcal/ kg). The emission factor analysis showed that the concentration of emission value that might occur during waste management activities is below the standard set by the Environment Quality Act 1974. Basically, this research shows that electronic plastic waste has the potential to become the source of raw material and energy recovery. (author)

  5. Treatment of groundwater for nitrate removal by portable ion exchange resin, OSE

    International Nuclear Information System (INIS)

    Iriburo, A.; Pessi, M.; Castagnino, G.; Garat, S.; Hackenbruch, R.; Laguardia, J.; Yelpo, L.; Amondarain, A.; Brunetto, C.

    2010-01-01

    The locations of Palmitas in the Province of Soriano is supplied with groundwater from a shallow and high nitrogen content in sedimentary aquifer (Asencio Formation). Due to lack of alternative sources, groundwater or surface water, it was decided to test the water treatment from a perforation whose tenors were of the order of 51-66 mg / L of nitrates. The methodology used for the removal of nitrate is ion exchange resins .The main issue raised in this case was the disposal of effluent from the washing of the resins, because there is no collective sanitation network Palmitas nor a sufficient stream flow for discharge . Several alternatives (installation of a transitional deposit, haulage trucks, dumping at distant points, etc.), which were ruled by their poor viability and / or high costs were studied. Finally it was decided to install a device that will have three cylinders with resins were transportable, for which should have a weight less than 75 kg and those which would be used alternately. Regeneration of the resins is carried out in the city of Mercedes, distant 40 km, where the necessary water for the discharge conditions exist with a high content of sodium chloride, resulting from ion exchange. This pilot project represents a first step in treatment for nitrate removal in groundwater using transportable resins which aims to supply the public . Due to the nature of the above location , the chosen methodology had to be adapted to fulfill their duties satisfactorily. The first results of this project to a year of commissioning implementation, which has been funded by SBI and developed by his staff, in order to be used in other places with similar problems are presented in this report

  6. Study of the effect of Kaolin in the mortar of cement matrices by confinement of ion exchange resins

    Directory of Open Access Journals (Sweden)

    Labied S.

    2018-01-01

    Full Text Available Radioactive waste arising as a result of nuclear activities should be safely managed from its generation to final disposal in an appropriate conditioned form to reduce the risk of radiation exposure of technical personnel and of the public and to limit contamination of the environment. The immobilization of low and intermediate level radioactive wastes in cementitious matrices is the most commonly used technique to produce inexpensive waste matrix that complies with regulatory requirements in order to protect humans and the environment against nuisance caused by ionizing radiation. Cement based materials are used in radioactive waste management to produce stable waste forms. This matrix constitutes the first build engineering barrier in disposal facilities. In this work, the kaolin is used to enhance the mechanical performance of the matrix of confinement of ion exchange resins by gradually replacing the sand in mortar with kaolin clay. The Kaolin clay sample was a special pure product, sourced from a foreign country. The maximum quantity of resins that can be incorporated into the mortar formulation without the packages losing their strength is 13.915% which results in a better mechanical strength at 6.7686 MPA compression with kaolin.

  7. Joule-heated glass-furnace system for the incineration of low-level radioactive wastes

    International Nuclear Information System (INIS)

    Armstrong, K.M.; Klingler, L.M.; Doty, J.W.; Kramer, D.P.

    1982-01-01

    For the past 1-1/2 years, Mound has been preparing and evaluating a commercially available joule-heated glass furnace unit, coupled with a wet scrubbing system. The purpose of the glass furnace evaluation is to advance and document incinerator technology for such combustibles as solids, resins, and sludges, and to develop a stable waste form for subsequent disposal. Four (4) waste nonradioactive types were selected to determine the combustion efficiency of the furnace unit: (1) dry solid waste composed of paper, plastics, rubber, and cloth, (2) ion exchange resin of both the anionic and cationic type, (3) filter sludge composed of diatomaceous earth, organic cellulosic filter aid, and powdered ion exchange resin, and (4) cartridge filters having glass and plastic filter surfaces and nonmetallic cores. When completed, the combustion efficiency experiments for the proposed nonradioactive waste-types revealed the ability of the furnace to easily incinerate waste at feedrates of up to 150 lb/hr. During the course of the experiments, combustibles in the offgas remained consistently low, suggesting excellent combustion efficiency. Furthermore, ash produced by the combustion process was effectively incorporated into the melt by convective currents in the glass. Future work on the glass furnace incinerator will include spiking the waste to determine radioisotope behavior in the furnace

  8. Results of field testing of waste forms using lysimeters

    International Nuclear Information System (INIS)

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

    1988-01-01

    The purpose of the field testing task, using lysimeter arrays, is to expose samples of solidified resin waste to the actual physical, chemical, and microbiological conditions of disposal enviroment. Wastes used in the experiment include a mixture of synthetic organic ion exchange resins and a mixture of organic exchange resins and an inorganic zeolite. Solidification agents used to produce the 4.8-by 7.6-cm cylindrical waste forms used in the study were Portland Type I-II cement and Dow vinyl ester-styrene. Seven of these waste forms were stacked end-to-end and inserted into each lysimeter to provide a 1-L volume. There are 10 lysimeters, 5 at ORNL and 5 at ANL-E. Lysimeters used in this study were designed to be self-contained units which will be disposed at the termination of the 20-year study. Each is a 0.91-by 3.12-m right-circular cylinder divided into an upper compartment, which contains fill material, waste forms, and instrumentation, and an empty lower compartment, which collects leachate. Four lysimeters at each site are filled with soil, while a fifth (used as a control) is filled with inert silica oxide sand. Instrumentation within each lysimeter includes porous cup soil-water samplers and soil moisture/temperature probes. The probes are connected to an on-site data acquisition and storage system (DAS) which also collects data from a field meteorological station located at each site. 9 refs

  9. Study on uranium adsorption ability of tannix resin from solution

    International Nuclear Information System (INIS)

    Le Thi Kim Dung; Le Quang Thai; Nguyen Lanh; Le Ngoc Thuy

    2004-01-01

    During past years, generated liquid waste from uranium ore processing has been treated by co-precipitation method in ITRRE. In this liquid waste treatment process, mixing liquid waste and lime, decantation, filtration of precipitate were implemented. The treated fluid has underlimited toxic concentration and ensures for moving into environment. Residue was dried and packed into drums as low level radioactive waste. Next to the advantages of this method such as simplest technology, cheapest cost, easy operation. Some amount of secondary radioactive waste as noncombustible materials must be stored with complicated technologies a highly cost. We have been researching a new liquid waste treatment system replaceable precipitation system. In the new process, insoluble tannin is utilized as adsorbent of uranium liquid waste. Advantage of insoluble tannin is expected to be possible to reduce its volume incineration as well as its adsorption ability. Those are the reasons why tannix resin is used this research subject. In this subject, we have studied adsorption capacity of uranium in Tannix, relation of adsorption rate and pH, the change of adsorption ability of column system, the pyrolysis curve of dried Tannix (author)

  10. Calpain 3 is important for muscle regeneration

    DEFF Research Database (Denmark)

    Hauerslev, Simon; Sveen, Marie-Louise; Duno, Morten

    2012-01-01

    Limb girdle muscular dystrophy (LGMD) type 2A is caused by mutations in the CAPN3 gene and complete lack of functional calpain 3 leads to the most severe muscle wasting. Calpain 3 is suggested to be involved in maturation of contractile elements after muscle degeneration. The aim of this study...... was to investigate how mutations in the four functional domains of calpain 3 affect muscle regeneration....

  11. Inclusion of radioactive ion-exchange resins into inorganic binders

    International Nuclear Information System (INIS)

    Epimakhov, V.N.; Olejnik, M.S.

    2005-01-01

    The paper is devoted to inclusion of the radioactive ion-exchange resins into the portland, slag-portland and alumina cements. The degree of filling the solidified products achieves 7-10, 12 and 18.9-19.7% correspondingly under conservation of sufficient strength (not less 5 MPa). The coefficient of waste volume increasing during solidification does not exceed 1.5 under consideration of addition of 10 mass % of clay into aluminia cement [ru

  12. Product consistency testing of three reference glasses in stainless steel and perfluoroalkoxy resin vessels

    International Nuclear Information System (INIS)

    Olson, K.M.; Smith, G.L.; Marschman, S.C.

    1995-03-01

    Because of their chemical durability, silicate glasses have been proposed and researched since the mid-1950s as a medium for incorporating high-level radioactive waste (HLW) generated from processing of nuclear materials. A number of different waste forms were evaluated and ranked in the early 1980s; durability (leach resistance) was the highest weighted factor. Borosilicate glass was rated the best waste form available for incorporation of HLW. Four different types of vessels and three different glasses were used to study the possible effect of vessel composition on durability test results from the Production Consistency Test (PCT). The vessels were 45-m 304 stainless steel vessels, 150-m 304 L stainless steel vessels, and 60-m perfluoroalkoxy (PFA) fluoropolymer resin vessels. The three glasses were the Environmental Assessment glass manufactured by Corning Incorporated and supplied by Westinghouse Savannah River company, and West Valley Nuclear Services reference glasses 5 and 6, manufactured and supplied by Catholic University of America. Within experimental error, no differences were found in durability test results using the 3 different glasses in the 304L stainless steel or PFA fluoropolymer resin vessels over the seven-day test period

  13. Kaolinite adsorption-regeneration system for dyestuff treatment by Fenton based processes.

    Science.gov (United States)

    Rosales, Emilio; Anasie, Delia; Pazos, Marta; Lazar, Iuliana; Sanromán, M Angeles

    2018-05-01

    The regeneration and reuse of adsorbents is a subject of interest nowadays in order to reduce the pollution and the wastes generated in the adsorption wastewater treatment. In this work, the regeneration of the spent kaolinite by different advanced oxidation processes (Fenton, electro-Fenton and electrokinetic-Fenton) was evaluated. Initially, it was confirmed the ability of a low cost clayey material, kaolinite, for the adsorption of model dye such as Rhodamine B showing Freundlich isotherm fitting. Then, the regeneration and consequent degradation of the pollutant in the adsorbent by Fenton based processes was carried out. The role of different parameters affecting the regeneration process (H 2 O 2 :Fe 2+ ratio, liquid:solid ratio) were evaluated. Working at 100:1 H 2 O 2 :Fe 2+ ratio and 30min near complete dye removal (around 97%) from kaolinite was obtained by Fenton treatment. After that, a two-stage treatment for adsorption-regeneration was evaluated during five treatment cycles demonstrating its viability for regeneration of the adsorbent through dye degradation. Based on the successful application of Fenton technique, the improvement of the treatment by electro-Fenton and electrokinetic-Fenton were studied for different solid:liquid ratios achieving satisfactory regeneration values. Copyright © 2017 Elsevier B.V. All rights reserved.

  14. Mathematical modelling of transport phenomena in radioactive waste-cement-bentonite matrix

    International Nuclear Information System (INIS)

    Plecas, Ilija; Dimovic, Slavko

    2010-01-01

    Document available in extended abstract form only. The leaching rate of 137 Cs from spent mix bead (anion and cation) exchange resins in a cement-bentonite matrix has been studied. Transport phenomena involved in the leaching of a radioactive material from a cement-bentonite matrix are investigated using three methods based on theoretical equations. These are: the diffusion equation for a plane source an equation for diffusion coupled to a first-order equation and an empirical method employing a polynomial equation. The results presented in this paper are from a 25-year mortar and concrete testing project that will influence the design choices for radioactive waste packaging for a future Serbian radioactive waste disposal center. Radioactive waste is waste material containing radioactive chemical elements which does not have a practical purpose. It is often the product of a nuclear process, such as nuclear fission. Waste can also be generated from the processing of fuel for nuclear reactors or nuclear weapons. The main objective in managing and disposing of radioactive (or other) waste is to protect people and the environment. This means isolating or diluting the waste so that the rate or concentration of any radionuclides returned to the biosphere is harmless. Storage as the placement of waste in a nuclear facility where isolation, environmental protection and human control are provided with the intent that the waste will be retrieved at a later time. Disposal as the emplacement of waste in an approved, specified facility (e.g. near surface or geological repository) without the intention of retrieval. The processing of radioactive wastes may be done for economic reasons (e.g. to reduce the volume for storage or disposal, or to recover a 'resource' from the waste), or safety reasons (e.g. converting the waste to a more 'stable' form, such as one that will contain the radionuclide inventory for a long time). Typically processing involves reducing

  15. Application of microwaves for incinerating waste shell moulds and cores

    Directory of Open Access Journals (Sweden)

    K. Granat

    2008-08-01

    Full Text Available In the paper, investigation results of microwave heating application for incinerating waste shell moulds and cores made of moulding sands with thermosetting resins are presented. It was found that waste shell cores or shell moulds left after casting, separated from moulding sand, can be effectively incinerated. It was evidenced that microwave heating allows effective control of this process and its results. Incineration of waste moulds and cores made of commercial grades of resin-coated moulding sand using microwave heating was found to be an effective way of their utilisation. It was determined that the optimum burning time of these wastes (except those insufficiently disintegrated and not mixed with an activating agent is maximum 240 s at the used magnetron power of 650 W. It was noticed that proper disintegration of the wastes and use of suitable additives to intensify the microwave heating process guarantee significant reduction of the process time and its full stabilisation. Application of microwave heating for incinerating waste shell moulds and cores ensure substantial and measurable economic profits due to shorter process time and lower energy consumption.

  16. Corrosion of steel drums containing simulated radioactive waste of low and intermediate level

    International Nuclear Information System (INIS)

    Farina, S.B.; Schulz Rodríguez, F.; Duffó, G.S.

    2013-01-01

    Ion-exchange resins are frequently used during the operation of nuclear power plants and constitute radioactive waste of low and intermediate level. For the final disposal inside the repository the resins are immobilized by cementation and placed inside steel drums. The eventful contamination of the resins with aggressive species may cause corrosion problems to the drums. In order to assess the incidence of this phenomenon and to estimate the lifespan of the steel drums, in the present work, the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins contaminated with different aggressive species was studied. The aggressive species studied were chloride ions (main ionic species of concern) and sulphate ions (produced during radiolysis of the cationic exchange-resins after cementation). The corrosion rate of the steel was monitored over a time period of 900 days and a chemical and morphological analysis of the corrosion products formed on the steel in each condition was performed. When applying the results obtained in the present work to estimate the corrosion depth of the drums containing the cemented radioactive waste after a period of 300 years (foreseen durability of the Low and Intermediate Level Radioactive Waste facility in Argentina), it was found that in the most unfavourable case (high chloride contamination), the corrosion penetration will be considerably lower than the thickness of the wall of the steel drums. (author)

  17. Expansive failure reactions and their prevention in the encapsulation of phenol formaldehyde type ion exchange resins in cement based systems

    Energy Technology Data Exchange (ETDEWEB)

    Constable, M.; Howard, C.G.; Johnson, M.A.; Jolliffe, C.B. (AEA Decommissioning and Waste Management, Winfrith (United Kingdom)); Sellers, R.M. (Nuclear Electric plc, Barnwood (United Kingdom))

    1992-01-01

    Lewatit DN is a phenol formaldehyde based ion exchange resin used to remove radioactive caesium from liquid waste streams such as fuel cooling ponds and effluents. This paper presents the results of a study of the encapsulation of the bead form of the resin in cement with particular reference to the mechanisms of its interaction with the encapsulant. When incorporated in pure ordinary Portland cement (OPC) at loadings in excess of 15 wt % an unstable product results due to expansion of the systems and at higher waste loadings failure results after only a few days. Evidence from differential scanning calorimetry, X-ray diffraction and scanning electron microscopy all indicate the cause of the expansive reaction to be the formation of crystals of calcium salts around and within the resin beads. Addition of BFS and sodium hydroxide prevent the formation of these salts by removal of calcium hydroxide from the system in other reactions. (author).

  18. Selective extraction of cobalt 60 and radium on CSA's SIMTM system

    International Nuclear Information System (INIS)

    Denton, M.S.; Higgins, I.R.

    1988-01-01

    CSA, has developed and promoted a new ground/well water treatment media and support system for hydrogen sulfide, iron, and manganese called The SIM System. This media was also found to have a selective affinity for arsenic, selenium, antimony, bismuth and lead. All heavy metals have not been checked, but of special rad-waste interest is Co-60 and Ra-226. Barium has been used as a stand-in for Ra in the lab, but both Co and Ba have exhibited very high affinities from Oak Ridge waters. A very attractive feature is that the adsorbed metals may be eluted with a minimum of acid, unlike ion exchangers that regenerate with difficulty or zeolites which dissolve in acids. This opens up the new dimension of extraction/concentration of these contaminants. Ordinary water softening ion exchange resin is excellent for extraction of Ra from drinking water, but, as mentioned, regenerates with great difficulty. The SIM media may be used for pre-treatment with the acid regenerant reloaded on ordinary resin for a very high concentration of Ra, or Co, for final disposal. The very stable SIM media may also be disposed of directly. Strontium (Sr-90) affinity isn't as great as for barium, but is greater than calcium. The separation factor of SIM for Sr/Ca may well be better than the value of two (2) observed for ordinary cation resin. Further studies are being made to demonstrate the full potential of this selective media for certain rad-wastes

  19. Campuran kitosan dengan resin akrilik sebagai bahan gigi tiruan penghambat Candida albicans

    Directory of Open Access Journals (Sweden)

    Titik Ismiyati

    2018-04-01

    Full Text Available The mixture of acrylic resin and chitosan as denture material to inhibit Candida albicans. The inhibition of Candida albicans in denture resin has an important role to prevent the development of denture stomatitis. Chitosan is a natural polymer compound derived from shrimp waste which can function as an antifungal Acrylik resin cannot be mixed with chitosan. To obtain a homogeneous mixture, the mixture was added a coupling agen acrylic acid and acetone. The research objective was to study the mixture of acrylic resin and chitosan with solvent acrylic acid and acetone as a denture that can inhibit the growth of Candida albicans. Methods: The samples used discs in 10 mm diameter and 2 mm thickness, made from heat cured acrylic resin mixed with chitosan dissolved in acrylat acid and acetone. They were divided into 4 groups. Group 1 was acrylic resin without chitosan as a control, group 2, 3 and 4 were the mixture of acrylic resin and 5 ml chitosan in 0.5%, 1%, and 2% concentration respectively. The fourier transform irfrared spectroscopy (FTIR and the digital optical microscope were used to synthesize and analyze. The Kruskal Wallis was used to analyze the data. The results showed that the mixture of acrylic resin with chitosan significantly inhibited the growth of Candida albicans. Conclusion: a mixture of acrylic resins and chitosan can be fungistatic, so it can be developed as an antifungal denture material.   ABSTRAK Penghambatan Candida albicans pada gigi tiruan resin akrilik dapat memainkan peran penting dalam mencegah perkembangan denture stomatitis. Kitosan adalah senyawa polimer alam yang berasal dari limbah udang yang dapat berfungsi sebagai antijamur. Resin akrilik tidak dapat bercampur dengan kitosan. Untuk mendapatkan campuran yang homogen, campuran tersebut ditambah coupling agent asam akrilat dan aseton. Penelitian ini bertujuan untuk mengkaji campuran resin akrilik dan kitosan dan asam akrilat pelarut aseton sebagai bahan gigi

  20. Development of radioactive waste management at Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    Miyanaga, I.; Sakata, S.; Ito, A.; Amano, H.

    1977-01-01

    For low- and medium-level waste treatment, main efforts have been put on the reduction of waste volume. For high-level wastes, studies are being carried out on the solidification and partitioning techniques in preparation for completion of the fuel cycle in Japan. For sea disposal of low-level wastes planned by the JAEC, significant information has been obtained regarding integrity and leaching behavior of cement solidified wastes. This paper describes the present status of development of the techniques in the following sections; 1. Treatment of Low- and Medium-Level Wastes; an incinerator with two stage ceramic filters has been tested, and the decontamination factor was found to be 10 4 for various nuclides; reverse osmosis method with a cellulose acetate membrane has been tested for laundry liquid waste, and 60 Co was removed more than 99% together with detergents; and solidification products of spent ion-exchange resin with polyethylene have been proved to be superior in mechanical properties, water resistance and volume reduction to asphalt products. 2. Safety Evaluation of Cement Solidified Wastes for Sea Disposal; homogeneous cement-solidified wastes in 200 l sealed drums did not show any cracks or defects under high hydrostatic pressure; the leaching ratio of 137 Cs for the first one year was estimated to be lower than 0.3%. 3. Treatment of High-Level Wastes; vitrification using natural zeolite has been developed and properties of the products were proved to be excellent; and a partitioning procedure consisting mainly of solvent extraction and ion-exchange method has been studied; reduction of the amount of alkaline agent by introducing a denitration technique, and reduction of resin volume by adopting a porous type resin were achieved

  1. Fire Safety Tests for Spherical Resorcinol Formaldehyde Resin: Data Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong-Sang; Peterson, Reid A.; Schweiger, Michael J.

    2012-07-30

    A draft safety evaluation of the scenario for spherical resorcinol-formaldehyde (SRF) resin fire inside the ion exchange column was performed by the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Fire Safety organization. The result of this draft evaluation suggested a potential change of the fire safety classification for the Cesium Ion Exchange Process System (CXP) emergency elution vessels, equipment, and piping, which may be overly bounding based on the fire performance data from the manufacturer of the ion exchange resin selected for use at the WTP. To resolve this question, the fire properties of the SRF resin were measured by Southwest Research Institute (SwRI), following the American Society for Testing and Materials (ASTM) standard procedures, through a subcontract managed by Pacific Northwest National Laboratory (PNNL). For some tests, the ASTM standard procedures were not entirely appropriate or practical for the SRF resin material, so the procedures were modified and deviations from the ASTM standard procedures were noted. This report summarizes the results of fire safety tests performed and reported by SwRI. The efforts by PNNL were limited to summarizing the test results provided by SwRI into one consolidated data report. All as-received SwRI reports are attached to this report in the Appendix. Where applicable, the precision and bias of each test method, as given by each ASTM standard procedure, are included and compared with the SwRI test results of the SRF resin.

  2. Method of plastic solidification of radioactive wastes

    International Nuclear Information System (INIS)

    Oikawa, Yasuo; Tokimitsu, Fujio.

    1986-01-01

    Purpose: To prevent occurrence of deleterious cracks to the inside and the surface of solidification products, as well as eliminate gaps between the products and the vessel inner wall upon plastic solidification processing for powdery or granular radioactive wastes. Method: An appropriate amount of thermoplastic resins such as styrenic polymer or vinyl acetate type polymer as a low shrinking agent is added and mixed with unsaturated polyester resins to be mixed with radioactive wastes so as to reduce the shrinkage-ratio to 0 % upon curing reaction. Thus, a great shrinkage upon hardening the mixture is suppressed to prevent the occurrence of cracks to the surface and the inside of the solidification products, as well as prevent the gaps between the inner walls of a drum can vessel and the products upon forming solidification products to the inside of the drum can. The resultant solidification products have a large compression strength and can sufficiently satisfy the evaluation standards as the plastic solidification products of radioactive wastes. (Horiuchi, T.)

  3. French processes for waste embedding. The use of epoxy resin for waste containment

    International Nuclear Information System (INIS)

    Augustin, X.; Gauthey, J.C.

    1993-01-01

    The low- and medium-level wastes generated by nuclear facilities when operating as well as during their decommissioning (dismantling, decontamination, etc.) are embedded for the purpose of obtaining a product suitable for disposal. Due to the varieties of waste produced, it was necessary to resort to multi-purpose techniques to solve problems relating to their embedding. The process for waste embedding in thermosetting polymer (polyester, epoxy) developed by the French Atomic Energy Commission (CEA) and its subsidiary TECHNICATOME is easy to operate and yields excellent results having regard to volume reduction and containment of radioisotopes (particularly caesium). The industrial development of this process has led to the design of small, flexible, fixed or mobile, embedding stations. Examples illustrating the increasing use of this process during facility dismantling are described

  4. Waste management at the Ardennes power plant

    International Nuclear Information System (INIS)

    Abraham, J.P.

    1979-01-01

    In 1976, the SENA (with the participation of EDF, CEA and CEC in the framework of a research program on the management and storage of radioactive wastes) has developed an industrial pilot plant for the encapsulation of wastes in thermosetting polyester resins. The industrial putting in operation of the plant will enable most of the wastes from the nuclear station to be processed. The quality of products will be improved and the volume and processing cost reduced

  5. Depleted ion exchange resins encapsulation with mobile unit: equipment and experience

    International Nuclear Information System (INIS)

    Cohen, S.; Faisantieu, D.

    1986-01-01

    Since 1981, STMI has been operating mobile units in EDF's PWR nuclear power plants, for spent resins encapsulation with polymer thermosetting matrices. Three mobile units are now in operation: COMET 1 and COMET 2, supplied by STMI, using polymerized styrene with proper additives as encapsulating material, and PRECED 1, on PEC-Engineering design, based on a DOW Chemical solidification process. On march 1986, more than 30 operations have been performed on EDF's PWR plants. More than 5000 liners containing encapsulated depleted ion exchange resins have been produced, while processing about 500 m 3 (i.e. 17.000 ft 3 ) of resins. During this period, those mobile units have shown their reliability and their efficiency. The produced processed waste, which have been accepted by ANDRA at the La Manche Storage Site (SSM) must meet the Fundamental Safety Rules (RFS) edicted by the Central Bureau for Nuclear Facilities Safety (SCSIN) of the French Department of Industry. The operations are carried out with excellent safety and radioprotection safety conditions, and following a very detailed Q.A. program [fr

  6. Laboratory-Scale SuperLig 639 Column Tests With Hanford Waste Simulants

    International Nuclear Information System (INIS)

    King, William D.; Spencer, William A.; Bussey, Myra Pettis

    2003-01-01

    This report describes the results of SuperLig 639 column tests conducted at the Savannah River Technology Center (SRTC) in support of the Hanford River Protection Project - Waste Treatment Plant (RPP-WTP). The RPP-WTP contract was awarded to Bechtel National Inc. (BNI) for the design, construction, and initial operation of a plant for the treatment and vitrification of millions of gallons of radioactive waste currently stored in tanks at Hanford, WA. Part of the current treatment process involves the removal of technetium from tank supernate solutions using columns containing SuperLig 639 resin. This report is part of a body of work intended to quantify and optimize the operation of the technetium removal columns with regard to various parameters (such as liquid flow rate, column aspect ratio, resin particle size, loading and elution temperature, etc.). The tests were conducted using nonradioactive simulants of the actual tank waste samples containing rhenium as a surrogate for the technetium in the actual waste. A previous report focused on the impacts of liquid flow rate and column aspect ratio upon performance. More recent studies have focused on the impacts of resin particle size, solution composition, and temperature. This report describes column loading experiments conducted varying temperature and solution composition. Each loading experiment was followed by high temperature elution of the sorbed rhenium. Results from limited testing are also described which were intended to evaluate the physical stability of SuperLig 639 resin during exposure to repeated temperature cycles covering the range of potential processing extremes

  7. Development studies for the treatment of ORNL low-level liquid waste

    International Nuclear Information System (INIS)

    Campbell, D.O.; Lee, D.D.; Dillow, T.A.

    1991-11-01

    An experimental program is under way to investigate potential separation methods for application to specific problems relating to the management of low-level liquid wastes (LLLWs) at ORNL. This report summarizes experimental results that were acquired during fiscal year 1990 and have not been previously reported elsewhere. Measurements are presented for cesium and strontium removal from simulated high-salt waste compositions, using both inorganic ion- exchange sorbents and organic ion-exchange resins, and for one experiment with actual LLLW supernate solution from Melton Valley Storage Tank W-26, using inorganic sorbents. The purpose of the study was to acquire an extensive data base to support the development of flowsheets for decontamination of the LLLW currently stored at ORNL. Experimental measurements with inorganic ion exchangers focused on batch separations of cesium using several transition-metal hexacyanoferrate(2) compositions (ferrocyanides) and of strontium using titanium oxide-based sorbents. Cesium distribution coefficients in the range of 1 x 10 6 were generally observed with nickel and cobalt ferrocyanides at pH values ≤11, yielding DFs of about 100 with 100 ppm sorbent in a single-stage batch separation. Most organic ion-exchange resins are not very effective for cesium removal from such high salt concentrations, but a new resorcinol-based resin developed at the Savannah River Site was found to be considerably superior to any other such material tested. Several chelating resins were effective for removing strontium from the waste simulants. An ion-exchange column test successfully demonstrated the simultaneous removal of both cesium and strontium from a waste simulant solution

  8. One pot synthesis of chitosan grafted quaternized resin for the removal of nitrate and phosphate from aqueous solution.

    Science.gov (United States)

    Banu, H Thagira; Meenakshi, Sankaran

    2017-11-01

    The present study deals with the synthesis of chitosan quaternized resin for efficient removal of nitrate and phosphate from aqueous solution. The resin was characterized with FTIR, SEM with EDX and XRD. Batch method was carried out to optimize various parameters such as contact time, initial concentration of nitrate and phosphate, dosage, pH, co-anions and temperature on the adsorption capacity of the adsorbent. The adsorption process illustrated that the Freundlich isotherm and the pseudo-second order are the best fitted models for the sorption of both anions. The respective negative values of ΔH° and ΔG° revealed that the adsorption of both the anions were exothermic and spontaneous. The removal efficiency of nitrate and phosphate on chitosan quaternized resin were 78% and 90% respectively with 0.1g of adsorbent and the initial concentration as 100mg/L. Nitrate and phosphate anions adsorbed effectively on chitosan quaternized resin by replacing Cl - ions from quaternary site through electrostatic attraction as well as ion-exchange mechanism. Hydrogen bonding also played important role in adsorption process. Even after 7th regeneration cycle the adsorbent retained its adsorption capacity as 23.7mg/g and 30.4mg/g for both nitrate and phosphate respectively. Copyright © 2017 Elsevier B.V. All rights reserved.

  9. Method of treating radioactive sludge waste

    International Nuclear Information System (INIS)

    Shoji, Yuichi; Matsuura, Hiroyuki; Ichihashi, Toshio

    1989-01-01

    For removing water content from sludge wastes, filtration or steam condensation may be considered, but none of them can sufficiently reduce the water content since filtration may cause clogging and steam treatment has a limit in the condensation. In view of the above, radioactive sludge wastes are dehydrated by an electroosmotic process in a vessel and then dehydrated solid contents are solidified in the vessel. Since the sludge wastes are mainly composed o fion exchange resins and iron oxides deposited to the resins, when a DC voltage is applied to the sludges containing such solid contents, a force tending to premeate them through the fine pores in the filter is exerted to water. As a result, only water is removed while the solids are being held on the filters. Since the moving direction of water is different depending on the property of the sludges, the polarity of the electrodes may be changed depending on the nature of the sludges. Thus, volume reduction can be improved and treating conditions can be controlled easily by a simple device. (N.H.)

  10. Studies Regarding As(V Adsorption from Underground Water by Fe-XAD8-DEHPA Impregnated Resin. Equilibrium Sorption and Fixed-Bed Column Tests

    Directory of Open Access Journals (Sweden)

    Mihaela Ciopec

    2014-10-01

    Full Text Available The characteristics of arsenic adsorption onto Fe-XAD8-DEHPA resin were studied on the laboratory scale using aqueous solutions and natural underground waters. Amberlite XAD8 resin was impregnated with di(2-ethylhexyl phosphoric acid (DEHPA via the dry method of impregnation. Fe(III ions were loaded onto the impregnated resin by exploiting the high affinity of arsenic towards iron. The studies were conducted by both in contact and continuous modes. Kinetics data revealed that the removal of arsenic by Fe-XAD8-DEHPA resin is a pseudo-second-order reaction. The equilibrium data were modelled with Freundlich Langmuir and Dubinin Radushkevich (D-R isotherms and it was found that the Freundlich model give the poorest correlation coefficient. The maximum adsorption capacity obtained from the Langmuir isotherm is 22.6 µg As(V/g of Fe-XAD8-DEHPA resin. The mean free energy of adsorption was found in this study to be 7.2 kJ/mol and the ΔG° value negative (−9.2 kJ/mol. This indicates that the sorption process is exothermal, spontaneous and physical in nature. The studied Fe-XAD8-DEHPA resin showed excellent arsenic removal performance by sorption, both from synthetic solution and the natural water sample, and could be regenerated simply by using aqueous NaOH or HCl solutions.

  11. Regeneration of 98Mo enriched from waste 99mTc generators

    International Nuclear Information System (INIS)

    Khujaev, S.; Mirzaeva, N.A.; Ashrapov, U.T.; Berdieva, M.; Nushtaeva, L.B.

    2002-01-01

    Radioisotope generator of technetium-99m have found widespread application in nuclear medicine for production of sodium pertechnetate solution ( 99m Tc). In technology of radioisotope generator making, which developed in Institute of Nuclear Physics of Uzbekistan Academy of Science the parent radioisotopes of molybdenum-90 are produced with neutron capture reaction of the enriched isotope of molybdenum-98 in nuclear reactor of WWR-SM. The specific activity of 99 Mo is reached to 5.0 Ci/g after activation process of original material ( 98 Mo). The basic mass of the original materials ( 98 Mo) which took for the irradiation is remained inactive. The high cost of enriched 97 Mo to make up necessity of development of method of 98 Mo regeneration with end in view of its recurring in manufacture. In general, the chromatographic type generator are produced in manufacture with aluminium oxide as sorbent for the parent radionuclides. So we studied the description conditions of molybdenum from aluminium oxide phase, its purification from possible accompanying admixtures and production of molybdenum in goods state. The systems: sulphur acid, ammonium and ammonium with hydrogen peroxide have been considered for molybdenum desorption from aluminium oxide phase. The sulphur oxid solution is most optimal medium for molybdenum desorption with concentration 0.5 M and elution rate for this makes up 2.0-2.05 ml/min. The 98 Mo yield makes up not over 87 %. The subsequent purification of the desorption of molybdenum-98 from the possible admixture are realized by using column with tetra-phosphonium alkali. At first , 98 Mo sorption is realized by anion exchange resin, after that sorbate is washing and dispiriting by solution with composition: 1.0 M NH 4 NO 3 +8.0 % NH 4 OH+10 -2 % H 2 O 2 . The 98 Mo yield makes up under 80 %. Further from obtained solution is realized by precipitation of molybdenum-98 in acid medium and separation of sediments from liquid phase, drying if sediments and

  12. [Ru(bipy)3]2+ nanoparticle-incorporate dental light cure resin to promote photobiomodulation therapy for enhanced vital pulp tissue repair

    Science.gov (United States)

    Mosca, Rodrigo C.; Young, Nicholas; Zeituni, Carlos A.; Arany, Praveen R.

    2018-02-01

    The use of nanoparticle on dental light cure resin is not new, currently several compounds (nanoadditives) are used to promote better communication between the restorative material and biological tissues. The interest for this application is growing up to enhance mechanical proprieties to dental tissue cells regeneration. Bioactive nanoparticles and complex compounds with multiple functions are the major target for optimizing the restorative materials. In this work, we incorporate [Ru(bipy)3]2+ nanoparticles, that absorbs energy at 450 nm (blue-light) and emits strongly at 620 nm (red-light), in PLGA Microspheres and insert it in Dental Light Cure Resin to promote the Photobiomodulation Therapy (PBM) effects to accelerate dental pulp repair by in vitro using cytotoxicity and proliferation assay.

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

    International Nuclear Information System (INIS)

    Zhou, H.; Colombo, P.

    1984-01-01

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

  14. Viscosity calculations of simulated ion-exchange resin waste glasses

    International Nuclear Information System (INIS)

    Kim, Cheon Woo; Park, Jong Kil; Lee, Kyung Ho; Lee, Myung Chan; Song, Myung Jae; BRUNELOT, Pierre

    2000-01-01

    An induction cold crucible melter (CCM) located in the NETEC-KEPCO has been used to vitrify simulated ion-exchange resin. During vitrification, the CCM operations were tightly constrained by glass viscosity as an important process parameter. Understanding the role of viscosity and quantifying viscosity is highly required in the determination of optimized feed formulations and in the selection of the processing temperature. Therefore, existing process models of glass viscosity based on a relationship between the glass composition, its structure polymerization, and the temperature were searched and adapted to our borosilicate glass systems. Calculated data using a viscosity model based on calculation of non-bridging oxygen (NBO) were in good agreement with the measured viscosity data of benchmark glasses

  15. Regional waste treatment with monolith disposal for low-level radioactive waste

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1983-01-01

    An alternative system is proposed for the disposal of low-level radioactive waste. This system, called REgional Treatment with MOnolith Disposal (RETMOD), is based on integrating three commercial technologies: automated package warehousing, whole-barrel rotary kiln incineration, and cement-based grouts for radioactive waste disposal. In the simplified flowsheet, all the sludges, liquids, resins, and combustible wastes are transported to regional facilities where they are incinerated. The ash is then mixed with special cement-based grouts, and the resulting mixture is poured into trenches to form large waste-cement monoliths. Wastes that do not require treatment, such as damaged and discarded equipment, are prepositioned in the trenches with the waste-cement mixture poured on top. The RETMOD system may provide higher safety margins by conversion of wastes into a solidified low-leach form, creation of low-surface area waste-cement monoliths, and centralization of waste processing into a few specialized facilities. Institutional problems would be simplified by placing total responsibility for safe disposal on the disposal site operator. Lower costs may be realized through reduced handling costs, the economics of scale, simplified operations, and less restrictive waste packaging requirements

  16. Evaluation of radioactive wastes in Instituto de Energia Atomica (Sao Paulo-Brazil)

    International Nuclear Information System (INIS)

    Sawakuchi, R.S.

    1978-01-01

    An evaluation of present and future production of radioactive waste in several departments of the Instituto de Energia Atomica has been done. Taking into account this evaluation, the criteria for disposal and convenient treatment technique have been studied. The most critical form of liquid radioactive waste is that of 131 I processing because high concentration of radiotellurium always accompanies this form of waste. Ion exchange and precipitation techniques were used to study this waste processing. Two kinds of resins were used by the ion exchange method: the strong anionic and the stron cationic. Quantitative tellurium retention has not been attained by the ion exchange method using either resins. The technique of precipitation of radioactive tellurium as ammonium tellurate was also used, allowing us to obtain more than 99% of tellurium removal. The remaining radioactive wastes can be eliminated using the storage for decay criteria with further release to the sewers in the case of liquids and burial in the case of solids. (Author) [pt

  17. Sequential determination of important ecotoxic radionuclides in nuclear waste samples

    International Nuclear Information System (INIS)

    Bilohuscin, J.

    2016-01-01

    In the dissertation thesis we focused on the development and optimization of a sequential determination method for radionuclides 93 Zr, 94 Nb, 99 Tc and 126 Sn, employing extraction chromatography sorbents TEVA (R) Resin and Anion Exchange Resin, supplied by Eichrom Industries. Prior to the attestation of sequential separation of these proposed radionuclides from radioactive waste samples, a unique sequential procedure of 90 Sr, 239 Pu, 241 Am separation from urine matrices was tried, using molecular recognition sorbents of AnaLig (R) series and extraction chromatography sorbent DGA (R) Resin. On these experiments, four various sorbents were continually used for separation, including PreFilter Resin sorbent, which removes interfering organic materials present in raw urine. After the acquisition of positive results of this sequential procedure followed experiments with a 126 Sn separation using TEVA (R) Resin and Anion Exchange Resin sorbents. Radiochemical recoveries obtained from samples of radioactive evaporate concentrates and sludge showed high efficiency of the separation, while values of 126 Sn were under the minimum detectable activities MDA. Activity of 126 Sn was determined after ingrowth of daughter nuclide 126m Sb on HPGe gamma detector, with minimal contamination of gamma interfering radionuclides with decontamination factors (D f ) higher then 1400 for 60 Co and 47000 for 137 Cs. Based on the acquired experiments and results of these separation procedures, a complex method of sequential separation of 93 Zr, 94 Nb, 99 Tc and 126 Sn was proposed, which included optimization steps similar to those used in previous parts of the dissertation work. Application of the sequential separation method for sorbents TEVA (R) Resin and Anion Exchange Resin on real samples of radioactive wastes provided satisfactory results and an economical, time sparing, efficient method. (author)

  18. Modified bitumen for embedding of radioactive wastes

    International Nuclear Information System (INIS)

    Bozkurt, C.

    1984-11-01

    This work describes new possibilities of using polymer modified bitumen as matrix material for embedding of low- and medium level radioactive wastes. Epoxy resins, polyurethane resins and styrene-butadiene-copolymers with 20-40 weight per cent are used as modifying agents. Penetration and softening point (ring and ball) of modified samples have been measured. Further the resistance to toluene and leaching rate in n-heptane have been determined. Within these polymer bitumen combinations investigated, the epoxy resins having a high epoxid equivalent weight with dicarbooxylic acid anhydrid hardeners and tertiary amin accelerators give the most dense network, highest thermodimensional stability and lowest leaching rate in organic solvents. 71 refs., 9 figs., 6 tabs

  19. Mixture Design and Its Application in Cement Solidification for Spent Resin

    International Nuclear Information System (INIS)

    Gan, Xueying; Lin, Meiqing; Chen, Hui

    1994-01-01

    The study is aimed to assess the usefulness of the mixture design for spent resin immobilization in cement. Although a considerable amount of research has been carried out to determine the limits for the composition of an acceptable resin-cement mixture, no efficient experimental strategy exists that explores the full properties of waste form against composition relationship. In order to gain an overall view, this report introduces the method of mixture design and mixture analysis, and describes the design of experiment of the 5-component mixture with the constraint conditions. The mathematic models of 28-day compressive strength varying with the ingredients are fitted, and the main effect and interaction effect of two ingredients are identified quantitatively along with the graphical interpretation using the response trace plot and contour plots

  20. Study of integral waste management systems and their effects on the environment in Thailand. Part of a coordinated programme on migration and disperios of radionuclides from storage of radioactive waste under various conditions in the terrestrial environment

    International Nuclear Information System (INIS)

    Karasuddhi, P.

    1976-02-01

    The research work carried out is divided into 3 parts: 1. Experimental study of sorption capacity of Sr-90, Cs-137 and radioactive liquid waste of OAEP onto soil and clay. 2. The fixation of radioactive sludge and resin in cement, gumcrete and bitumen. The results show that the sludge from the OAEP waste treatment plant can be fixed very well with concrete, gumcrete and bitumen, but the solidification of resin in concrete, gumcrete and bitumen is not a good method for waste treatment. In the third part, the results of environment studies around the nuclear research center are presented

  1. Radioactive waste management

    International Nuclear Information System (INIS)

    Kizawa, Hideo

    1982-01-01

    A system of combining a calciner for concentrated radioactive liquid waste and an incinerator for miscellaneous radioactive solid waste is being developed. Both the calciner and the incinerator are operated by fluidized bed method. The system features the following points: (1) Inflammable miscellaneous solids and concentrated liquid can be treated in combination to reduce the volume. (2) Used ion-exchange resin can be incinerated. (3) The system is applicable even if any final waste disposal method is adopted; calcinated and incinerated solids obtained as intermediate products are easy to handle and store. (4) The system is readily compatible with other waste treatment systems to form optimal total system. The following matters are described: the principle of fluidized-bed furnaces, the objects of treatment, system constitution, the features of the calciner and incinerator, and the current status of development. (J.P.N.)

  2. Study of the retention of radionuclides by ion-exchange resins contained in the circuits of a Pressurized Water Reactor; Etude de la retention des radionucleides dans les resines echangeuses d'ions des circuits d'une centrale nucleaire a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Gressier, F.

    2008-11-15

    Physico-chemical quality of fluids in nuclear power plant circuits must be maintained in order to limit contamination and dose rate especially when the shutdown takes place. Nevertheless, an optimum between diminishing liquid waste and limiting solid waste production has to be reached, but at affordable costs. Ion-exchange resins of purification circuits are used to fulfill this goal. In this work, different resin types have been characterized (exchange capacity, water and electrolyte sorption) and their selectivity towards Co{sup 2+}, Ni{sup 2+}, Cs{sup +} and Li{sup +} cations have been studied. We have shown that the two cation-exchange resins selectivity varies according to the nature and concentrations of their counter-ions. Moreover, flow rate (and thus hydro-kinetics) impact on species retention in a column has been characterized: the more the flow rate, the more the ionic leakage (output concentration divided by input concentration) is fast and the more the output concentration front is spread. A literature revue has enabled to put in light advantages and drawbacks of the models of interest to simulate operations of ion-exchange resins. Thus, the pure end-members mixing model associated to a non-ideality description of the resin phase based on the regular solutions model has been retained for modelling ion-exchange equilibrium. Ion-exchange kinetics has been described by mass transfer coefficients. Using the experimental results to determine model parameters, these last ones have been implemented in a speciation code CHESS, coupled with a hydrodynamic code in HYTEC. On the one hand, equilibrium experiments of ion retention have been simulated and, on the other hand, column retention tests have been modelled. Finally, selectivity variations and hydro-kinetics impacts have been simulated on some test cases so as to demonstrate the importance of taking these into account when simulating ion-exchange resins operations. (author)

  3. Centralized cement solidification technique for low-level radioactive wastes

    International Nuclear Information System (INIS)

    Matsuda, Masami; Nishi, Takashi; Izumida, Tatsuo; Tsuchiya, Hiroyuki.

    1996-01-01

    A centralized cement solidification system has been developed to enable a single facility to solidify such low-level radioactive wastes as liquid waste, spent ion exchange resin, incineration ash, and miscellaneous solid wastes. Since the system uses newly developed high-performance cement, waste loading is raised and deterioration of waste forms after land burial prevented. This paper describes the centralized cement solidification system and the features of the high-performance cement. Results of full-scale pilot plant tests are also shown from the viewpoint of industrial applicability. (author)

  4. Influence of natural sorbents in immobilization of radioactive waste in cement

    International Nuclear Information System (INIS)

    Plecas, I.; Dimovic, S.

    2006-01-01

    Leach characteristics of 137 Cs and 60 Co radionuclides from spent mix bead ion exchange resins and both ordinary Portland cement and cement mixed with two kind of natural sorbents, (bentonite and clinoptilolite) have been studied using International Atomic Energy's (IAEA) standard leach method. A study is undertaken to determine the waste immobilization performance of low-level wastes in cement-natural sorbents mixtures. The solidification matrix was a standard Portland cement mixed with 290-350 (kg/m 3 ) spent mix bead exchange resins, with or without 1-10 % of bentonite or/and clinoptilolite The leaching rates from the cement-bentonite matrix as 60 Co: (1.20-9.72)x10 -5 (cm/d) and for 137 Cs: (1.00-9.22)x10 -4 (cm/d), after 300 days were measured. From the leaching data the apparent diffusivity of cobalt and cesium in cement bentonite or/and clinoptilolite matrix with a waste load of 350 (kg/m 3 ) spent mix bead exchange resins was measured as 60 Co: (1.0-5.9)x10 -6 (cm 2 /d) and for 137 Cs: (0.48-2.4)x10 -4 (cm 2 /d) after 300 days. The compressive strength of these samples is determined following the ASTM standards. These results are part of a 30-year mortar and concrete testing project which will influence the design of radioactive waste management for a future Serbian radioactive waste disposal center. (author)

  5. Review of disposal techniques for radioactively contaminated organic ion-exchange resins

    International Nuclear Information System (INIS)

    Inman, J.R.; Mack, J.

    1993-03-01

    Organic ion-exchange resins are used in the UK nuclear industry to remove radioactive nuclides from dilute aqueous solution. Resins represent a significant proportion of the organic content of ILW and LLW, particularly ILW. Spent resins are destined to be disposed of in the UK deep repository. There are concerns regarding the potential effects of organic materials on long-term repository performance, and these effects have been the subject of much recent research work. The object of this study has been to conduct a worldwide review of treatment and conditioning techniques available for spent organic ion-exchange resins with the intention of recommending the best option for dealing with the waste in the UK. Data on available techniques have been gathered together, and are presented in tabular form at the back of the report. The techniques have been reviewed and compared considering safety, practicality and cost, and a best option selected on the basis of current knowledge. On balance it would appear that wet oxidation using hydrogen peroxide with residue encapsulation in BFS/OPC is the most appropriate technique, probably implemented using a mobile plant. This conclusion and recommendation is not however clear cut and further advice regarding the acceptability of organic material in the repository is necessary before a definite recommendation can be made. (Author)

  6. Method of processing radioactive liquid waste

    International Nuclear Information System (INIS)

    Motojima, Kenji; Kawamura, Fumio.

    1981-01-01

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

  7. Cesium Ion Exchange Program at the Hanford River Protection Project Waste Treatment Plant

    International Nuclear Information System (INIS)

    CHARLES, NASH

    2004-01-01

    The River Protection Project - Hanford Tank Waste Treatment and Immobilization Plant will use cesium ion exchange to remove 137Cs from Low Activity Waste down to 0.3 Ci/m3 in the Immobilized LAW, ILAW product. The project baseline for cesium ion exchange is the elutable SuperLig, R, 644, SL-644, resin registered trademark of IBC Advanced Technologies, Inc., American Fork, UT or the Department of Energy approved equivalent. SL-644 is solely available through IBC Advanced Technologies. To provide an alternative to this sole-source resin supply, the RPP--WTP initiated a three-stage process for selection and qualification of an alternative ion exchange resin for cesium removal in the RPPWTP. It was recommended that resorcinol formaldehyde RF be pursued as a potential alternative to SL-644

  8. Solidification of low-level wastes by inorganic binder

    International Nuclear Information System (INIS)

    Sasaki, M.T.; Shimojo, M.; Suzuki, K.; Kajikawa, A.; Karasawa, Y.

    1995-01-01

    The use of an alkali activated slag binder has been studied for solidification and stabilization of low-level wastes in nuclear power stations and spent fuel processing facilities. The activated slag effectively formed waste products having good physical properties with high waste loading for sodium sulfate, sodium nitrate, calcium pyrophosphate/phosphate and spent ion-exchange resins. Moreover, the results of the study suggest the slag has the ability to become a common inorganic binder for the solidification of various radioactive wastes. This paper also describes the fixation of radionuclides by the activated slag binder

  9. Method of producing solidification product of radioactive waste

    International Nuclear Information System (INIS)

    Masuda, Shunji; Iwami, Etsuji; Kadota, Keishi.

    1989-01-01

    Layers of thermosetting resin composition capable of curing at normal temperature are formed to a thickness of 2 to 5 mm at the bottom of a container. As the thermosetting resin composition capable of curing at normal temperature, there can be mentioned, for example, unsaturated polyester resin comprising a polymerizable monomer and an unsaturated polyester. After the layers are cured, a mixture of radioactive wastes and the thermosetting resin composition capable of curing at normal temperature is filled on the layer. After curing, thermosetting resin composition capable of curing at normal temperature is filled so as to fill gaps between the curing product and the container caused by curing shrinkage and at the upper surface of the curing products. After curing, plastic layers are formed at the surface. This can avoid residual bubbles in the layers or development of cracks. Further, leaching rate of Na ions is low and water proofness can be improved as well. (T.M.)

  10. Correlations of norbornenyl crosslinked polyimide resin structures with resin thermo-oxidative stability, resin glass transition temperature and composite initial mechanical properties

    Science.gov (United States)

    Alston, William B.

    1988-01-01

    PMR (polymerization of monomeric reactants) methodology was used to prepare 70 different polyimide oligomeric resins and 30 different unidirectional graphite fiber/polyimide composites. Monomeric composition as well as chain length between sites of crosslinks were varied to examine their effects on resin thermo-oxidative stability and glass transition temperature (Tg) of the cured/postcured resins. A linear correlation of decreasing 316 C resin weight loss/surface area versus (1) decreasing aliphatic content, or (2) increasing benzylic/aliphatic content stoichiometry ratio over a wide range of resin compositions was observed. An almost linear correlation of Tg versus molecular distance between the crosslinks was also observed. An attempt was made to correlate Tg with initial composite mechanical properties (flexural strength and interlaminar shear strength). However, the scatter in mechanical strength data prevented obtaining a clear correlation. Instead, only a range of composite mechanical properties was obtained at 25, 288, and 316 C. Perhaps more importantly, what did become apparent during the correlation study was (1) the PMR methodology could be used to prepare composites from resins containing a wide variety of monomer modifications, (2) that these composites almost invariably provided satisfactory initial mechanical properties as long as the resins formulated exhibited satisfactory processing flow, and (3) that PMR resins exhibited predictable rates of 316 C weight loss/surface area based on their benzylic/aliphatic stoichiometery ratio.

  11. Incineration, pyrolysis and gasification of electronic waste

    Science.gov (United States)

    Gurgul, Agnieszka; Szczepaniak, Włodzimierz; Zabłocka-Malicka, Monika

    2017-11-01

    Three high temperature processes of the electronic waste processing: smelting/incineration, pyrolysis and gasification were shortly discussed. The most distinctive feature of electronic waste is complexity of components and their integration. This type of waste consists of polymeric materials and has high content of valuable metals that could be recovered. The purpose of thermal treatment of electronic waste is elimination of plastic components (especially epoxy resins) while leaving non-volatile mineral and metallic phases in more or less original forms. Additionally, the gaseous product of the process after cleaning may be used for energy recovery or as syngas.

  12. Incineration, pyrolysis and gasification of electronic waste

    Directory of Open Access Journals (Sweden)

    Gurgul Agnieszka

    2017-01-01

    Full Text Available Three high temperature processes of the electronic waste processing: smelting/incineration, pyrolysis and gasification were shortly discussed. The most distinctive feature of electronic waste is complexity of components and their integration. This type of waste consists of polymeric materials and has high content of valuable metals that could be recovered. The purpose of thermal treatment of electronic waste is elimination of plastic components (especially epoxy resins while leaving non-volatile mineral and metallic phases in more or less original forms. Additionally, the gaseous product of the process after cleaning may be used for energy recovery or as syngas.

  13. Process and device for processing radioactive wastes

    International Nuclear Information System (INIS)

    1974-01-01

    A method is described for processing liquid radioactive wastes. It includes the heating of the liquid wastes so that the contained liquids are evaporated and a practically anhydrous mass of solid particles inferior in volume to that of the wastes introduced is formed, then the transformation of the solid particles into a monolithic structure. This transformation includes the compressing of the particles and sintering or fusion. The solidifying agent is a mixture of polyethylene and paraffin wax or a styrene copolymer and a polyester resin. The device used for processing the radioactive liquid wastes is also described [fr

  14. Polyvinyl chloride resin

    International Nuclear Information System (INIS)

    Kim, Hong Jae

    1976-06-01

    This book contains polyvinyl chloride resin industry with present condition such as plastic industry and polyvinyl chloride in the world and Japan, manufacture of polyvinyl chloride resin ; suspension polymerization and solution polymerization, extruding, injection process, hollow molding vinyl record, vacuum forming, polymer powders process, vinyl chloride varnish, vinyl chloride latex, safety and construction on vinyl chloride. Each chapter has descriptions on of process and kinds of polyvinyl chloride resin.

  15. Study of the retention of radionuclides by ion-exchange resins contained in the circuits of a Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Gressier, F.

    2008-11-01

    Physico-chemical quality of fluids in nuclear power plant circuits must be maintained in order to limit contamination and dose rate especially when the shutdown takes place. Nevertheless, an optimum between diminishing liquid waste and limiting solid waste production has to be reached, but at affordable costs. Ion-exchange resins of purification circuits are used to fulfill this goal. In this work, different resin types have been characterized (exchange capacity, water and electrolyte sorption) and their selectivity towards Co 2+ , Ni 2+ , Cs + and Li + cations have been studied. We have shown that the two cation-exchange resins selectivity varies according to the nature and concentrations of their counter-ions. Moreover, flow rate (and thus hydro-kinetics) impact on species retention in a column has been characterized: the more the flow rate, the more the ionic leakage (output concentration divided by input concentration) is fast and the more the output concentration front is spread. A literature revue has enabled to put in light advantages and drawbacks of the models of interest to simulate operations of ion-exchange resins. Thus, the pure end-members mixing model associated to a non-ideality description of the resin phase based on the regular solutions model has been retained for modelling ion-exchange equilibrium. Ion-exchange kinetics has been described by mass transfer coefficients. Using the experimental results to determine model parameters, these last ones have been implemented in a speciation code CHESS, coupled with a hydrodynamic code in HYTEC. On the one hand, equilibrium experiments of ion retention have been simulated and, on the other hand, column retention tests have been modelled. Finally, selectivity variations and hydro-kinetics impacts have been simulated on some test cases so as to demonstrate the importance of taking these into account when simulating ion-exchange resins operations. (author)

  16. Treatment of low level radioactive liquid wastes using composite ion-exchange resins based on polyurethane foam

    International Nuclear Information System (INIS)

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

    2010-01-01

    Composite ion-exchange resins were prepared by coating copper-ferrocyanide (CFC) and hydrous manganese oxide (HMO) powders on polyurethane (PU) foam. Polyvinyl acetate/Acetone was used as a binder. The foam was loaded with about five times its weight with CFC and HMO powders. The distribution coefficients of CFC-PU foam and HMO-PU foam for cesium and strontium respectively were estimated. Under similar conditions the HMO-PU foam showed higher capacity as well as better kinetics for removal of strontium than CFC-PU foam for Cs. The pilot plant scale studies were conducted using a mixed composite ion-exchange resin bed. About 1000 bed volumes could be passed before attaining a DF of 10 from an initial value of 60-80. The spent resin was digested in alkaline KMnO 4 and the digested liquid was fixed in cement matrix. The matrices were characterized with respect to compressive strength and leach resistance. (author)

  17. Immobilisation in cement of ion exchange resins arising from the purification of reagents used for the decontamination of reactor circuits

    International Nuclear Information System (INIS)

    Howard, C.G.; Jolliffe, C.; Lee, D.J.

    1988-04-01

    The aim of the programme is to show that ion exchange resins used to remove activity from decontaminating agents used in water reactors can be successfully immobilised in cement. To achieve this, blends of Ordinary Portland Cement and ground granulated Blast Furnace Slag (ratio 9:1) have been used. Improvements in the properties of the product and the waste loading of 50 w/o damp resin can be achieved using microsilica, a finely divided form of silicon dioxide, as an additive to the blended cement. This report contains data on the effects of anion resins, and mixed anion/cation resins, on the performance of the cemented product. The effects of organic acids, especially picolinic and formic acids, bound to anion resins have also been investigated. In addition, formulations developed have been assessed at commercial scale (200 litres of cemented product) for their process and product characteristics. The final part of the report deals with the long-term product performance of samples prepared from cation resins which are now nearly one year old. (author)

  18. Low-level liquid waste decontamination by organic ion exchange

    International Nuclear Information System (INIS)

    Lee, D.D.; Campbell, D.O.; Dillow, T.A.

    1990-01-01

    Improved processes are being developed to treat contaminated liquid wastes that have been and continue to be generated at Oak Ridge National Laboratory. Promising results have been obtained for cesium removal with a new resorcinol-based organic resin developed at the Savannah River Site. In tests of cesium removal, it was superior to other available resins, such as Duolite CS-100, with the distribution coefficient being limited primarily by competition from potassium and nearly independent of the sodium concentration. The optimum pH was approximately 12.5 in high NaNO 3 concentrations (>2 M). A fairly low flow velocity was required to yield sharp breakthrough of the loaded cesium. The resin was much less effective for strontium removal, which was limited by competition from sodium. If both cesium and strontium must be removed, another resin column or a mixed bed with a chelating resin should be used

  19. Influence of Coating with Some Natural Based Materials on the Erosion Wear Behavior of Glass Fiber Reinforced Epoxy Resin

    OpenAIRE

    Aseel Basim Abdul Hussein; Emad Saadi AL-Hassani; Reem Alaa Mohamed

    2015-01-01

    In the present study, composites were prepared by Hand lay-up molding. The composites constituents were epoxy resin as a matrix, 6% volume fractions of glass fibers (G.F) as reinforcement and 3%, 6% volume fractions of preparation natural material (Rice Husk Ash, Carrot Powder, and Sawdust) as filler. Studied the erosion wear behavior and coating by natural wastes (Rice Husk Ash) with epoxy resin after erosion. The results showed the non – reinforced epoxy have lower resistance erosion than n...

  20. Spent catalyst waste management. A review. Part 1. Developments in hydroprocessing catalyst waste reduction and use

    Energy Technology Data Exchange (ETDEWEB)

    Marafi, M.; Stanislaus, A. [Petroleum Refining Department, Petroleum Research and Studies Center, Kuwait Institute for Scientific Research, P.O. Box 24885, 13109-Safat (Kuwait)

    2008-04-15

    Solid catalysts containing metals, metal oxides or sulfides, which play a key role in the refining of petroleum to clean fuels and many other valuable products, become solid wastes after use. In many refineries, the spent catalysts discarded from hydroprocessing units form a major part of these solid wastes. Disposal of spent hydroprocessing catalysts requires compliance with stringent environmental regulations because of their hazardous nature and toxic chemicals content. Various options such as minimizing spent catalyst waste generation by regeneration and reuse, metals recovery, utilization to produce useful materials and treatment for safe disposal, could be considered to deal with the spent catalyst environmental problem. In this paper, information available in the literature on spent hydroprocessing catalyst waste reduction at source by using improved more active and more stable catalysts, regeneration, rejuvenation and reuse of deactivated catalysts in many cycles, and reusing in other processes are reviewed in detail with focus on recent developments. Available methods for recycling of spent hydroprocessing catalysts by using them as raw materials for the preparation of active new catalysts and many other valuable products are also reviewed. (author)

  1. Extraction behaviour of Am(III) and Eu(III) from nitric acid medium in TEHDGA-HDEHP impregnated resins

    Energy Technology Data Exchange (ETDEWEB)

    Saipriya, G.; Kumar, T. [Bhabha Atomic Research Centre Facilities, Kalpakkam (India). Kalpakkam Reprocessing Plant; Kumaresan, R.; Nayak, P.K.; Venkatesan, K.A.; Antony, M.P. [Indira Gandhi Center for Atomic Research, Kalpakkam (India). Fuel Chemistry Div.

    2016-07-01

    The extraction behaviour of Am(III) and Eu(III) from nitric acid medium was studied in the solvent impregnated resins containing extractants such as tetra-bis(2-ethylhexyl)diglycolamide (TEHDGA) or bis-(2-ethylhexyl)phosphoric acid (HDEHP) or mixture of TEHDGA+HDEHP. The rate of extraction of Am(III) and Eu(III) from 1 M nitric acid and the effect of various parameters, such as the concentration of nitric acid in aqueous phase and concentration of TEHDGA and HDEHP in resin phase, on the distribution coefficient of Am(III) and Eu(III) was studied. The distribution coefficient of Am(III) and Eu(III) in HDEHP-impregnated resin decreased and that in TEHDGA-impregnated resin increased, with increase in the concentration of nitric acid. However, in (TEHDGA+HDEHP) - impregnated resin, synergic extraction was observed at lower nitric acid concentration and antagonism at higher nitric acid concentration. The mechanism of Am(III) and Eu(III) extraction in the combined resin was investigated by slope analysis method. The extraction of various metal ions present in the fast reactor simulated high-level liquid waste was studied. The separation factor of Am(III) over Eu(III) was studied using citrate-buffered diethylenetriaminepentaacetic acid (DTPA) solution.

  2. Removal of organochlorine pesticides from water using virgin and regenerated granular activated carbon

    Directory of Open Access Journals (Sweden)

    MIRJANA B. NINKOVIĆ

    2010-04-01

    Full Text Available Public water systems use granular activated carbon in order to eliminate pesticides. After saturation, the used activated carbon is regenerated and reused in order to reduce the costs of water production and minimize waste. In this study, the adsorption of 10 different chlorinated pesticides from water using columns packed with commercial virgin and regenerated granular activated carbon was simulated in order to compare their adsorption capacities for different chlorinated pesticides. The breakthrough curves showed that chlorinated pesticides from the group of hexachlorocyclohexane (HCH were poorly adsorbed, followed by cyclodiens as averagely adsorbed and the derivatives of halogenated aromatic hydrocarbons (DDT as strongly adsorbed. However, the adsorption capacity of regenerated granular activated carbon was considerably lower for tested pesticides compared to the virgin granular carbon. In addition, rinsing of the pesticides after the saturation point is a far more efficient process on regenerated carbon.

  3. Radioactive waste material disposal

    Science.gov (United States)

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

    1995-01-01

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

  4. Determination of 238Pu, 239+240Pu, 241Pu and 241Am in radioactive waste from IPEN reactor

    International Nuclear Information System (INIS)

    Geraldo, Bianca; Taddei, Maria Helena T.; Cheberle, Sandra M.; Ferreira, Marcelo T.

    2011-01-01

    Ion exchange resin is a common type of radioactive waste arising from treatment of coolant water of the main circuit of research and nuclear power reactors. This waste contains high concentrations of fission and activation products. The management of this waste includes its characterization in order to determine and quantify specific radionuclides including those known as difficult-to-measure radionuclides (RDM). The analysis of RDMs generally involves expensive and time-consuming complex radiochemical analysis for purification and separation of the radionuclides. The objective of this work is to show an easy methodology for quantifying plutonium and americium isotopes in spent ion exchange resin, used for purification of the cooling water of the IEA-R1 reactor located at the Nuclear and Energy Research Institute, IPEN-CNEN/SP. The resins were destroyed by acid digestion, followed by purification and separation of the Pu and Am isotopes with anionic and chromatographic resins. 238 Pu, 239 + 24 '0Pu, and 24 '1Am isotopes were analyzed in an alpha spectrometer equipped with surface barrier detectors. 241 Pu isotope was analyzed by liquid scintillation counting. Chemical recovery yield ranged from 73 to 98% for Pu and 77 to 98% for Am, demonstrating that the methodology is suitable for identification and quantification of the isotopes studied in spent resins. (author)

  5. Evaluation of the Compressive Strength of Cement-Spent Resins Matrix Mixed with Bio char

    International Nuclear Information System (INIS)

    Zalina Laili; Muhamad Samudi Yasir; Zalina Laili; Mohd Abdul Wahab; Nur Azna Mahmud; Nurfazlina Zainal Abidin

    2015-01-01

    The evaluation of compressive strength of cement-spent resins matrix mixed with bio char was investigated. In this study, bio char with different percentage (5 %, 8 %, 11 % 14 % and 18 %) was used as alternative admixture material for cement solidification of spent resins. Some properties of the physical and chemical of spent resins and bio char were also investigated. The performance of cemented spent resins with the addition of bio char was evaluated based on their compressive strength and the water resistance test. The compressive strength was evaluated at three different curing periods of 7, 14 and 28 days, while 4 weeks of immersion in distilled water was chosen for water resistance test. The result indicated that the compressive strength at 7, 14 and 28 days of curing periods were above the minimum criterion for example > 3.45 MPa of acceptable level for cemented waste form. Statistical analysis showed that there was no significant relationship between the compressive strength of the specimen and the percentage of bio char content. Result from the water resistance test showed that only one specimen that contained of 5 % of bio char failed the water resistance test due to the high of spent resins/ bio char ratio. The compressive strength of cement solidified spent resins was found increased after the water resistance test indicating further hydration occurred after immersed in water. The results of this study also suggest that the specimen with 8 %, 11 %, 14 % and 18 % of bio char content were resistance in water and suitable for the leaching study of radionuclides from cement-bio char-spent resins matrix. (author)

  6. Leaching studies of low-level radioactive waste forms

    International Nuclear Information System (INIS)

    Dayal, R.; Arora, H.; Milian, L.; Clinton, J.

    1985-01-01

    A research program has been underway at the Brookhaven National Laboratory to investigate the release of radionuclides from low-level waste forms under laboratory conditions. This paper describes the leaching behavior of Cs-137 from two major low-level waste streams, that is, ion exchange bead resin and boric acid concentrate, solidified in Portland cement. The resultant leach data are employed to evaluate and predict the release behavior of Cs-137 from low-level waste forms under field burial conditions

  7. Dealing with operational power station wastes

    Energy Technology Data Exchange (ETDEWEB)

    Pepper, R B [Central Electricity Generating Board, London (UK). Nuclear Health and Safety Dept.

    1981-08-01

    The disposal of wastes from nuclear power stations is discussed. Liquid and gaseous wastes, from magnox stations, which are of low level activity, are dispersed to the sea or estuaries on coastal sites or for the case of Trawfynyeld, to the nearby lake. Low activity solid wastes are either disposed of on local authority tips or in shallow land burial sites. Intermediate level wastes, consisting mainly of wet materials such as filter sludges and resins from cooling ponds, are at present stored in shielded storage tanks either dry or under water. Only one disposal route for intermediate waste is used by Britain, namely, sea-dumping. Materials for sea dumping have to be encapsulated in a durable material for example, concrete.

  8. Plants for water recycling, oxygen regeneration and food production

    Science.gov (United States)

    Bubenheim, D. L.

    1991-01-01

    During long-duration space missions that require recycling and regeneration of life support materials the major human wastes to be converted to usable forms are CO2, hygiene water, urine and feces. A Controlled Ecological Life Support System (CELSS) relies on the air revitalization, water purification and food production capabilities of higher plants to rejuvenate human wastes and replenish the life support materials. The key processes in such a system are photosynthesis, whereby green plants utilize light energy to produce food and oxygen while removing CO2 from the atmosphere, and transpiration, the evaporation of water from the plant. CELSS research has emphasized the food production capacity and efforts to minimize the area/volume of higher plants required to satisfy all human life support needs. Plants are a dynamic system capable of being manipulated to favour the supply of individual products as desired. The size and energy required for a CELSS that provides virtually all human needs are determined by the food production capacity. Growing conditions maximizing food production do not maximize transpiration of water; conditions favoring transpiration and scaling to recycle only water significantly reduces the area, volume, and energy inputs per person. Likewise, system size can be adjusted to satisfy the air regeneration needs. Requirements of a waste management system supplying inputs to maintain maximum plant productivity are clear. The ability of plants to play an active role in waste processing and the consequence in terms of degraded plant performance are not well characterized. Plant-based life support systems represent the only potential for self sufficiency and food production in an extra-terrestrial habitat.

  9. Post-irradiation hardness of resin-modified glass ionomer cements and a polyacid-modified composite resin

    International Nuclear Information System (INIS)

    Yap, A.U.J.

    1997-01-01

    This study examined the post-irradiation hardness of resin-modified glass ionomer cements and a polyacid-modified composite resin using a digital microhardness tester. Change in hardness of these materials over a period of 6 months was compared to that of conventional glass ionomer cements and a composite resin. With the exception of the composite resin, all materials showed a significant increase in hardness over 24 h after their initial set. Dual-cure resin-modified glass ionomer cements showed decreased hardness with increased storage time in saline at 37 o C. Results suggest that the addition of resins to glass ionomer cements does not improve initial hardness and does not negate the acid-base reaction of conventional cements. Resin addition may, however, lead to increased water sorption and decreased hardness. (author)

  10. Thermal cycling effects on adhesion of resin-bovine enamel junction among different composite resins.

    Science.gov (United States)

    Chen, Wen-Cheng; Ko, Chia-Ling; Wu, Hui-Yu; Lai, Pei-Ling; Shih, Chi-Jen

    2014-10-01

    Thermal cycling is used to mimic the changes in oral cavity temperature experienced by composite resins when used clinically. The purpose of this study is to assess the thermal cycling effects of in-house produced composite resin on bonding strength. The dicalcium phosphate anhydrous filler surfaces are modified using nanocrystals and silanization (w/NP/Si). The resin is compared with commercially available composite resins Filtek Z250, Z350, and glass ionomer restorative material GIC Fuji-II LC (control). Different composite resins were filled into the dental enamel of bovine teeth. The bond force and resin-enamel junction graphical structures of the samples were determined after thermal cycling between 5 and 55°C in deionized water for 600 cycles. After thermal cycling, the w/NP/Si 30wt%, 50wt% and Filtek Z250, Z350 groups showed higher shear forces than glass ionomer GIC, and w/NP/Si 50wt% had the highest shear force. Through SEM observations, more of the fillings with w/NP/Si 30wt% and w/NP/Si 50wt% groups flowed into the enamel tubule, forming closed tubules with the composite resins. The push-out force is proportional to the resin flow depth and uniformity. The push-out tubule pore and resin shear pattern is the most uniform and consistent in the w/NP/Si 50wt% group. Accordingly, this developed composite resin maintains great mechanical properties after thermal cycling. Thus, it has the potential to be used in a clinical setting when restoring non-carious cervical lesions. Copyright © 2014 Elsevier Ltd. All rights reserved.

  11. Short-Term Effects of Reduced-Impact Logging on Copaifera spp. (Fabaceae Regeneration in Eastern Amazon

    Directory of Open Access Journals (Sweden)

    Carine Klauberg

    2017-07-01

    Full Text Available Timber management directly influences the population dynamics of tree species, like Copaifera spp. (copaíba, which provide oil-resin with ecological and economic importance. The aim of this study was to evaluate the structure and population dynamics of Copaifera in unmanaged and managed stands by reduced-impact logging (RIL in eastern Amazon in Pará state, Brazil. Based on a stem map of the study area, 40 Copaifera trees were randomly selected, where an equal number of trees were selected in managed and unmanaged stands. A transect of 10 × 100 m was centered at each tree (50 m each side to assess Copaifera regeneration. Transects were subdivided into ten plots, of which six were systematically chosen to assess the height, diameter and number of Copaifera seedlings and saplings. The field assessment occurred in 2011 and 2013. To estimate the amount of sunlight transmitted to the forest floor, we computed canopy cover from airborne LiDAR data. According to the results, the abundance of Copaifera seedlings/saplings was higher in managed than unmanaged stands. About 5% of Copaifera regeneration was found between 45–50 m from the Copaifera tree while ~73% of regeneration was concentrated within a 10 m radius of the Copaifera tree. We verified that the diameter distribution of Copaifera regeneration was not a negative exponential distribution, as is typical of most tree species in natural forest. Rather, the Copaifera regeneration had a spatially aggregated distribution. In this short-term analysis, the impact of timber management is not negatively affecting the population structure or dynamics of Copaifera regeneration.

  12. Resin composites

    DEFF Research Database (Denmark)

    Benetti, Ana Raquel; Peutzfeldt, Anne; Lussi, Adrian

    2014-01-01

    OBJECTIVE: To investigate how the modulus of elasticity of resin composites influences marginal quality in restorations submitted to thermocyclic and mechanical loading. METHODS: Charisma, Filtek Supreme XTE and Grandio were selected as they were found to possess different moduli of elasticity...... of resin composite (p=0.81) on the quality of dentine margins was observed, before or after loading. Deterioration of all margins was evident after loading (p....008). CONCLUSIONS: The resin composite with the highest modulus of elasticity resulted in the highest number of gap-free enamel margins but with an increased incidence of paramarginal enamel fractures. CLINICAL SIGNIFICANCE: The results from this study suggest that the marginal quality of restorations can...

  13. Biodegradation testing of solidified low-level waste streams

    International Nuclear Information System (INIS)

    Piciulo, P.L.; Shea, C.E.; Barletta, R.E.

    1985-05-01

    The NRC Technical Position on Waste Form (TP) specifies that waste should be resistant to biodegradation. The methods recommended in the TP for testing resistance to fungi, ASTM G21, and for testing resistance to bacteria, ASTM G22, were carried out on several types of solidified simulated wastes, and the effect of microbial activity on the mechanical strength of the materials tested was examined. The tests are believed to be sufficient for distinguishing between materials that are susceptible to biodegradation and those that are not. It is concluded that failure of these tests should not be regarded of itself as an indication that the waste form will biodegrade to an extent that the form does not meet the stability requirements of 10 CFR Part 61. In the case of failure of ASTM G21 or ASTM G22 or both, it is recommended that additional data be supplied by the waste generator to demonstrate the resistance of the waste form to microbial degradation. To produce a data base on the applicability of the biodegradation tests, the following simulated laboratory-scale waste forms were prepared and tested: boric acid and sodium sulfate evaporator bottoms, mixed-bed bead resins and powdered resins each solidified in asphalt, cement, and vinyl ester-styrene. Cement solidified wastes supported neither fungal nor bacterial growth. Of the asphalt solidified wastes, only the forms of boric acid evaporator bottoms did not support fungal growth. Bacteria grew on all of the asphalt solidified wastes. Cleaning the surface of these waste forms did not affect bacterial growth and had a limited effect on the fungal growth. Only vinyl esterstyrene solidified sodium sulfate evaporator bottoms showed viable fungi cultures, but surface cleaning with solvents eliminated fungal growth in subsequent testing. Some forms of all the waste streams solidified in vinyl ester-styrene showed viable bacteria cultures. 13 refs., 12 tabs

  14. Nuclide content in reactor waste

    International Nuclear Information System (INIS)

    1981-11-01

    Certain corrosion and fission products of importance in reactor waste management cannot be measured by gammaspectrometric techniques. In this study, a method is suggested by which the occurence of such nuclides can be quantitatively related to suitable gamma-emitters of similar origin. The method is tested by statistical analysis on the waste data recorded from two Swedish nuclear power plants. As this method is not applicable for Carbon-14, this nuclide was measured directly in spent ion exchange resins from three Finnish and Swedish power plants. (author)

  15. Evaluation of destructive methods for managing decontamination wastes

    International Nuclear Information System (INIS)

    Piciulo, P.L.; Adams, J.W.

    1986-01-01

    Results are discussed of a laboratory evaluation of destructive methods for processing chemical decontamination wastes. Incineration, acid digestion and wet-air oxidation are capable of degrading decontamination reagents and organic ion-exchange resins. The extent of destruction as a function of operating parameters was waste specific. The reagents used in the testing were: EDTA, oxalic acid, citric acid, picolinic acid and LND-101A

  16. The incineration of radioactive waste

    International Nuclear Information System (INIS)

    Thegerstroem, C.

    1980-03-01

    In this study, made on contract for the Swedish Nuclear Power Inspectorate, different methods for incineration of radioactive wastes are reviewed. Operation experiences and methods under development are also discussed. The aim of incineration of radioactive wastes is to reduce the volume and weight of the wastes. Waste categories most commonly treated by incineration are burnable solid low level wastes like trash wastes consisting of plastic, paper, protective clothing, isolating material etc. Primarily, techniques for the incineration of this type of waste are described but incineration of other types of low level wastes like oil or solvents and medium level wastes like ion-exchange resins is also briefly discussed. The report contains tables with condensed data on incineration plants in different countries. Problems encountered, experiences and new developments are reviewed. The most important problems in incineration of radioactive wastes have been plugging and corrosion of offgas systems, due to incomplete combustion of combustion of materials like rubber and PVC giving rise to corrosive gases, combined with inadequate materials of construction in heat-exchangers, channels and filter housings. (author)

  17. Modelling approach to LILW-SL repository safety evaluation for different waste packing options

    International Nuclear Information System (INIS)

    Perko, Janez; Mallants, Dirk; Volckaert, Geert; Towler, George; Egan, Mike; Virsek, Sandi; Hertl, Bojan

    2007-01-01

    The key objective of the work described here was to support the identification of a preferred disposal concept and packaging option for low and short-lived intermediate level waste (LILW-SL). The emphasis of the assessment, conducted on behalf of the Slovenian radioactive waste management agency (ARAO), was the consideration of several waste treatment and packaging options in an attempt to identify optimised containment characteristics that would result in safe disposal, taking into account the cost-benefit of alternative safety measures. Waste streams for which alternative treatment and packaging solutions were developed and evaluated include decommissioning waste and NPP operational wastes, including drums with unconditioned ion exchange resins in over-packed tube type containers (TTCs). For decommissioning wastes, the disposal options under consideration were either direct disposal of loose pieces grouted into a vault or use of high integrity containers (HIC). In relation to operational wastes, three main options were foreseen. The first is over-packing of resin containing TTCs grouted into high integrity containers, the second option is complete treatment with hydration, neutralization, and cementation of the dry resins into drums grouted into high integrity containers and the third is direct disposal of TTCs into high integrity containers without additional treatment. The long-term safety of radioactive waste repositories is usually demonstrated with the support of a safety assessment. This normally includes modelling of radionuclide release from a multi-barrier near-surface or deep repository to the geosphere and biosphere. For the current work, performance assessment models were developed for each combination of siting option, repository design and waste packaging option. Modelling of releases from the engineered containment system (the 'near-field') was undertaken using the AMBER code. Detailed unsaturated water flow modelling was undertaken using the

  18. Comparison of Mechanical Properties of Resin Composites with Resin Modified Glass Ionomers

    Directory of Open Access Journals (Sweden)

    Taha NA

    2015-06-01

    Full Text Available Statement of Problem: There are controversial reports regarding physical and mechanical properties of resin composites and glass ionomer cements. Some revealed higher strength and hardness for resin composites while others showed a comparable value for glass ionomer cements. Evaluation of mechanical properties of different types of resin composites in comparison with resin modified glass ionomers is not widely studied. Objectives: To measure and compare the flexural strength and Vickers hardness of three resin composites and two resins modified glass ionomer cements before and after ageing. Materials and Methods: Three resin composites, i.e. Filtek Supreme XTE (3M ESPE, Ice (SDI, Gradia (GC, and two resins modified glass ionomers, i.e. Fuji II LC (GC and Riva Light Cure (SDI, were selected. Ten barshaped specimens were prepared for each material and cured using LED curing light. After 24 hours storage in distilled water at 37oC, the specimens were randomly divided into two equal groups (n=5. The first group was tested as a baseline and the second group was restored at 37oC for another 29 days. Flexural strength was performed by four-point bending test using universal testing machine at crosshead speed of 0.5mm/min, and the maximum load at failure was recorded. The specimen’s halves were used for evaluating Vickers hardness, using a Digital Hardness Tester (300 g/15 sec and the Vickers hardness number (VHN was recorded. Data were analyzed using one-way analysis of variance (ANOVA, Tukey’s and student’s t-test. Results: After 24 hours of immersion, the highest hardness number was found for Filtek Supreme and Ice and the highest flexural strength was obtained for Gradia. After 30 days of storage, hardness of Fuji II LC and Gradia showed a significant decrease; flexural strength of Ice and Fuji II LC revealed a significant increase while Gradia and Filtek Supreme showed a significant decrease. Conclusions: Resin modified glass ionomers showed

  19. Technical solutions for waste treatment in the Belene project

    International Nuclear Information System (INIS)

    Büttner, K.; Eichhorn, H.

    2011-01-01

    Outline: In June 2010 NUKEM Technologies GmbH was awarded a contract from ATOMSTROYEXPORT JSC to perform the complete work package related to designing and completion of the equipment for treatment of radioactive waste on the turn-key basis for Belene NPP. Technical Solutions: Waste Streams and Technologies at UKC and UKS; Concentration Plant; Thermal Treatment of Resins Sorting Facility; Biological Waste Water Treatment; Conditioning – Cementation • Sorting of Radwaste; Plasma Facility; Grouting; Filter Press; Monitoring and Tracking

  20. Low and intermediate radioactive waste characterization using MICROSHIELD 5 code

    International Nuclear Information System (INIS)

    Mateescu, Silvia; Pantazi, Doina; Stanciu, Marcela

    2002-01-01

    Low and intermediate radioactive gaseous, liquid and solid waste produced at Cernavoda Nuclear Power Plant must be known from the point of view of contained radionuclide activity, during all steps of their processing, storage and transport, to ensure the nuclear safety of radioactive waste management. As the waste activity changes by radioactive decay and nuclear transmutation, the evolution in time of these sources is necessary to be assess, for the purpose of biological shielding determination at any time. On the other hand, during the transport of waste package at the repository, the external dose rates must meet the national and international requirements concerning radioactive materials transportation on public roads. In this paper, a calculation methodology for waste characterization based on external exposure rate measurement and on sample analysis results is presented. The time evolution of waste activity, as well as the corresponding shielding at different moments of management process, has been performed using MICROSHIELD-5 code. The spent resins proceeded from systems for clean-up and purification of cooling water and moderator, water from spent fuel storage bays, etc. have been analyzed. In this paper an example of spent ionic resins characterization, using the MICROSHIELD 5 code, is presented. (authors)

  1. Final waste classification and waste form technical position papers

    International Nuclear Information System (INIS)

    1983-05-01

    The waste classification technical position paper describes overall procedures acceptable to NRC staff which may be used by licensees to determine the presence and concentrations of the radionuclides listed in section 61.55, and thereby classifying waste for near-surface disposal. This technical position paper also provides guidance on the types of information which should be included in shipment manifests accompanying waste shipments to near-surface disposal facilities. The technical position paper on waste form provides guidance to waste generators on test methods and results acceptable to NRC staff for implementing the 10 CFR Part 61 waste form requirements. It can be used as an acceptable approach for demonstrating compliance with the 10 CFR Part 61 waste structural stability criteria. This technical position paper includes guidance on processing waste into an acceptable stable form, designing acceptable high-integrity containers, packaging cartridge filters, and minimizing radiation effects on organic ion-exchange resins. The guidance in the waste form technical position paper may be used by licensees as the basis for qualifying process control programs to meet the waste form stability requirements, including tests which can be used to demonstrate resistance to degradation arising from the effects of compression, moisture, microbial activity, radiation, and chemical changes. Generic test data (e.g., topical reports prepared by vendors who market solidification technology) may be used for process control program qualification where such generic data is applicable to the particular types of waste generated by a licensee

  2. Programme and french realizations concerning alpha wastes

    International Nuclear Information System (INIS)

    Sousselier, Y.

    1978-01-01

    Water reactors and breeder spent fuels are reprocessed to recover plutonium, minimise wastes and decrease irradiation risks. Alloys formation, glass addition and vitrification or metallic matrix are studied to treat cladding hulls. Plutonium content is controlled by alpha spectrometry or prompt neutrons measurements or neutrons activation. Wastes are calcinated or crushed at low temperature to recover transuranium elements by solvent extraction or precipitation or ionic exchange or ultrafiltration. Wastes are calcinated or crushed at low temperature to recover transuranium elements by solvent extraction or precipitation or ionic exchange or ultrafiltration. Wastes are embedded into bitumen or thermosetting resins and long term storage in geologic formation is studied [fr

  3. Development of polymer concrete radioactive waste management containers

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.; Lee, M. S.; Ahn, D. H.; Won, H. J.; Kang, H. S.; Lee, H. S.; Lim, S.P.; Kim, Y. E.; Lee, B. O.; Lee, K. P.; Min, B. Y.; Lee, J.K.; Jang, W. S.; Sim, W. B.; Lee, J. C.; Park, M. J.; Choi, Y. J.; Shin, H. E.; Park, H. Y.; Kim, C. Y

    1999-11-01

    A high-integrity radioactive waste container has been developed to immobilize the spent resin wastes from nuclear power plants, protect possible future, inadvertent intruders from damaging radiation. The polymer concrete container is designed to ensure safe and reliable disposal of the radioactive waste for a minimum period of 300 years. A built-in vent system for each container will permit the release of gas. An experimental evaluation of the mechanical, chemical, and biological tests of the container was carried out. The tests showed that the polymer concrete container is adequate for safe disposal of the radioactive wastes. (author)

  4. Nuclear heat-load limits for above-grade storage of solid transuranium wastes

    International Nuclear Information System (INIS)

    Clontz, B.G.

    1978-06-01

    Nuclear safety and heat load limits were established for above-grade storage of transuranium (TRU) wastes. Nuclear safety limits were obtained from a study by J.L. Forstner and are summarized. Heat load limits are based on temperature calculations for TRU waste drums stored in concrete containers (hats), and results are summarized. Waste already in storage is within these limits. The limiting factors for individual drum heat load limits were (1) avoidance of temperatures in excess of 190 0 F (decomposition temperature of anion resin) when anion resin is present in a concrete hat, and (2) avoidance of temperatures in excess of 450 0 F (ignition temperature of paper) at any point inside a waste drum. The limiting factor for concrete had heat load limits was avoidance of temperatures in excess of 265 0 F (melt point of high density polyethylene) at the drum liners. A temperature profile for drums and hats filled to recommended limits is shown. Equations and assumptions used were conservative

  5. The mechanism of uranium adsorption on Resin 508 and isoelectric point of the resin

    International Nuclear Information System (INIS)

    Han Qingping; Lu Weichang; Su Huijuan; Hu Jinbo; Zhang Liqin; Chen Banglin

    1990-01-01

    The adsorption process of uranium by Resin 508 at the solid-liquid interface was investigated and the mechanism of uranium adsorption including adsorption dynamics, adsorption thermodynamics and isoelectric point of resin was studied. The results are as follows: The maximum of uranium adsorption is attained at pH5-7; Uranium adsorption isotherm by Resin 508 in experimental conditions agrees with Langmuir's adsorption isotherm, the maximum of uranium adsorbed (Vm) is 716 mg U/g-dried resin; The adsorption of uranium by Resin 508 is an endothermic reaction and ΔH = 16.87 kJ/mol; The exchange-adsorption rate is mainly controlled by liquid film diffusion; The isoelectric points of Resin 508 before and after uranium adsorption are found to be pH7.5 and pH5.7 respectively. It is a specific adsorption for uranium

  6. Waste form development/test

    International Nuclear Information System (INIS)

    Kalb, P.D.; Colombo, P.

    1983-01-01

    The main objective of this study is to investigate new solidification agents relative to their potential application to wastes generated by advanced high volume reduction technologies, e.g., incinerator ash, dry solids, and ion exchange resins. Candidate materials selected for the solidification of these wastes include a modified sulfur cement and low-density polyethylene, neither of which are currently employed commerically for the solidification of low-level waste (LLW). As both the modified sulfur cement and the polyethylene are thermoplastic materials, a heated screw type extruder is utilized in the production of waste form samples for testing and evaluation. In this regard, work is being conducted to determine the range of conditions under which these solidification agents can be satisfactorily applied to the specific LLW streams and to provide information relevant to operating parameters and process control

  7. Ion exchange technology in the remediation of uranium contaminated groundwater at Fernald

    International Nuclear Information System (INIS)

    Sutton, Chris; Glassmeyer, Cathy; Bozich, Steve

    2000-01-01

    Using pump and treat methodology, uranium contaminated groundwater is being removed from the Great Miami Aquifer at the Fernald Environmental Management Project (FEMP) per the FEMP Record of Decision (ROD) that defines groundwater cleanup. Standard extraction wells pump about 3900 gallons-per-minute (gpm) from the aquifer through five ion exchange treatment systems. The largest treatment system k the Advanced Wastewater Treatment (AWWT) Expansion System with a capacity of 1800 gpm, which consists of three trains of two vessels. The trains operate in parallel treating 600 gpm each, The two vessels in each train operate in series, one in lead and one in lag. Treated groundwater is either reinfected back into the aquifer to speed up the aquifer cleanup processor discharged to the Great Miami River. The uranium regulatory ROD limit for discharge to the river is 20 parts per billion (ppb), and the FEMP uranium administrative action level for reinfection is 10 ppb. Spent (i.e., a resin that no longer adsorbs uranium) ion exchange resins must either be replaced or regenerated. The regeneration of spent ion exchange resins is considerably more cost effective than their replacement. Therefore, a project was undertaken to learn how best to regenerate the resins in the groundwater vessels. At the outset of this project, considerable uncertainty existed as to whether a spent resin could be regenerated successfully enough so that it performed as well as new resin relative to achieving very low uranium concentrations in the effluent. A second major uncertain y was whether the operational lifetime of a regenerated resin would be similar to that of a new resin with respect to uranium loading capacity and effluent concentration behavior. The project was successful in that a method for regenerating resins has been developed that is operationally efficient, that results in regenerated resins yielding uranium concentrations much lower than regulatory limits, and that results in

  8. Field lysimeter investigations: Low-level waste data base development program for fiscal year 1995. Volume 8, Annual report, October 1994-- September 1995

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Rogers, R.D.; Larsen, I.L.; Jastrow, J.D.; Sanford, W.E.; Sullivan, T.M.

    1996-06-01

    The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program, funded by the U.S. Nuclear Regulatory Commission, is (a) studying the degradation effects in organic ion-exchange resins caused by radiation, (b) examining the adequacy of test procedures recommended in the Branch Technical Position on Waste Form to meet the requirements of 10 CFR 61 using solidified ion-exchange resins, (c) obtaining performance information on solidified ion- exchange resins in a disposal environment, and (d) determining the condition of liners used to dispose the ion-exchange resins. Compressive test results of 12-year-old cement and vinyl ester- styrene solidified waste form samples are presented, which show effects of aging and self-irradiation. Results of the tenth year of data acquisition from the field testing are presented and discussed. During the continuing field testing, both portland type I-II cement and Dow vinyl ester-styrene waste form samples are being tested in lysimeter arrays located at Argonne National Laboratory-East in Illinois and at Oak Ridge National Laboratory. The study is designed to provide continuous data on nuclide release and movement, as well as environmental conditions, over a 20-year period

  9. Nordic study on reactor waste. Technical part 1 and 2

    International Nuclear Information System (INIS)

    1981-08-01

    An important part of the Nordic studies on system- and safety analysis of the management of low and medium level radioactive waste from nuclear power plants, is the safety analysis of a Reference System. This reference system was established within the study and is described in this Technical Part 1. The reference system covers waste management Schemes that are potential possibilities in either one of the four participating Nordic countries. The reference system is based on: a power reactor system consisting of 6 BWR's of 500 MWe each, operated simultaneously over the same 30 year period, and deep bed granular ion exchange resin wastes from the Reactor Water Clean-Up System (RWCS and powdered ion exchange resin from the Spent Fuel Pool Cleanup System (SFPCS)). Both waste types are supposed to be solidified by mixing with cement and bitumen. Two basic types of containers are considered. Standard 200 liter steel drums and specially made cubicreinforced concrete moulds with a net volume of 1 m 3 . The Nordic study assumes temporary storage of the solidified waste for a maximum of 50 years before the waste is transferred to the disposal site. Transportation of the waste from the storage facilitiy to the disposal site will be by road or sea. Three different disposal facilities are considered: Shallow land burial, near surface concrete bunker, and rock cavern with about 30 m granite cover. (EG)

  10. Decontamination factors of ceramic filter in radioactive waste incineration system

    International Nuclear Information System (INIS)

    Kanbe, Hiromi; Mayuzumi, Masami; Ono, Tetsuo; Yoshiki, Shinya; Kouyama, Hiroaki; Nagae, Madoka; Sekiguchi, Ryosaku; Takaoku, Yoshinobu; Hozumi, Masahiro.

    1987-01-01

    A suspension-firing type radioactive waste incineration system is developed and cold demonstration testing of ceramic filters for the system are carried out. The incineration system, which is useful for a wide variety of waste materials, can serve to simplify the facilities and to reduce the costs for waste disposal. The incineration system can be used for drying-processing of concentrated waste liquids and disposal of flame resistant materials including ion exchange resins and rubber, as well as for ordinary combustible solid materials. An on-line backwash system is adopted to allow the ceramic filters to operate stably for a long period of time. For one-step filtering using the ceramic filter, the decontamination factor is greater than 10 5 for the processing of various wastes. In a practical situation, there exist vapor produced by the spray drier and the cladding in used ion exchange resin, which act to increase the decontamination performance of the ceramic filters to ensure safe operation. For the waste incineration system equipped with a waste gas processing apparatus consisting of a ceramic filter and HEPA filter, the overall decontamination factor is expected to be greater than 10 6 at portions down to the outlet of the ceramic filter and greater than 10 8 at portions down to the outlet of the HEPA filter. (Nogami, K.)

  11. Characteristics of floc formation of anion and cation exchange resin in precoat filter using powdered ion exchange resin

    International Nuclear Information System (INIS)

    Adachi, Tetsurou; Sawa, Toshio; Shindoh, Toshikazu.

    1989-01-01

    The filtration performance of mixed filter aid consisting of powdered anion and cation exchange resins used in the precoat filter is closely related to the characteristics of floc formation. The physical, chemical and electrochemical properties of powdered ion exchange resin were measured and the factors related to floc formation of anion and cation exchange resin were investigated by measuring the specific settle volume of resin floc as an evaluating index. It was found that these factors were mixing ratio, nature of resins and particle size of resins. In addition, it was assumed on the bases of these results that the amount of resin floc was related to sum of the surface electric charges of both resins. The filling ratio of resin floc was related to their product by multiplication and an experimental expression was obtained. The specific settle volume of resin floc could then be simulated by particle size, surface area, ion exchange capacity and degree of ionization of the powdered ion exchange resin. (author)

  12. Characteristics of floc formation of anion and cation exchange resin in precoat filter using powdered ion exchange resin

    Energy Technology Data Exchange (ETDEWEB)

    Adachi, Tetsurou (Nitto Denko Corp., Ibaraki, Osaka (Japan)); Sawa, Toshio; Shindoh, Toshikazu

    1989-09-01

    The filtration performance of mixed filter aid consisting of powdered anion and cation exchange resins used in the precoat filter is closely related to the characteristics of floc formation. The physical, chemical and electrochemical properties of powdered ion exchange resin were measured and the factors related to floc formation of anion and cation exchange resin were investigated by measuring the specific settle volume of resin floc as an evaluating index. It was found that these factors were mixing ratio, nature of resins and particle size of resins. In addition, it was assumed on the bases of these results that the amount of resin floc was related to sum of the surface electric charges of both resins. The filling ratio of resin floc was related to their product by multiplication and an experimental expression was obtained. The specific settle volume of resin floc could then be simulated by particle size, surface area, ion exchange capacity and degree of ionization of the powdered ion exchange resin. (author).

  13. Method of processing radioactive liquid waste

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  14. Seasonal variability in nutrient regeneration by mussel Mytilus edulis rope culture in oligotrophic systems

    NARCIS (Netherlands)

    Jansen, H.M.; Strand, O.; Strohmeier, T.; Krogness, C.; Verdegem, M.C.J.; Smaal, A.C.

    2011-01-01

    Blue mussel Mytilus edulis cultures contribute to nutrient cycling in coastal ecosystems. Mussel populations filter particulate nutrients from the water column and inorganic nutrients are regenerated by excretion of metabolic wastes and decomposition of (pseudo-)faeces. The objective of this study

  15. Enhancement of mass transfer by ultrasound: Application to adsorbent regeneration and food drying/dehydration.

    Science.gov (United States)

    Yao, Ye

    2016-07-01

    The physical mechanisms of heat and mass transfer enhancement by ultrasound have been identified by people. Basically, the effect of 'cavitation' induced by ultrasound is the main reason for the enhancement of heat and mass transfer in a liquid environment, and the acoustic streaming and vibration are the main reasons for that in a gaseous environment. The adsorbent regeneration and food drying/dehydration are typical heat and mass transfer process, and the intensification of the two processes by ultrasound is of complete feasibility. This paper makes an overview on recent studies regarding applications of power ultrasound to adsorbent regeneration and food drying/dehydration. The concerned adsorbents include desiccant materials (typically like silica gel) for air dehumidification and other ones (typically active carbon and polymeric resin) for water treatment. The applications of ultrasound in the regeneration of these adsorbents have been proved to be energy saving. The concerned foods are mostly fruits and vegetables. Although the ultrasonic treatment may cause food degradation or nutrient loss, it can greatly reduce the food processing time and decrease drying temperature. From the literature, it can be seen that the ultrasonic conditions (i.e., acoustic frequency and power levels) are always focused on during the study of ultrasonic applications. The increasing number of relevant studies argues that ultrasound is a very promising technology applied to the adsorbent regeneration and food drying/dehydration. Copyright © 2016 Elsevier B.V. All rights reserved.

  16. An experimental feasibility study on vitrification of Low - and medium-level radioactive waste

    International Nuclear Information System (INIS)

    Park, Jongkil; Song, Myungjae; Choe, Youngson; Cho, Myungyul

    1996-01-01

    Laboratory and pilot tests(all cold tests) were carried out to examine the possibility of vitrification of low-level radioactive waste such as combustible DAW(protection clothes and vinyl seat), ion exchange resins, and evaporator bottoms with three types of vitrification equipment. Pyrolyzed or dried waste material and glass formers were fed into the melting cavity, converted to molten glassy mixture, and poured into a canister. For examination of the optimal ash contents in borosilicate glass waste forms with respect to waste types, compressive strength tests were conducted for several samples of ash contents. In the case of protection clothes, vinyl seat, and spent resin was rapidly reduced up to 5 or 6 times lower than that of neat glass, but hardly changed for dried evaporator bottoms. In order to investigate the possibility of direct vitrification, combustible DAW and spent resin were directly fed into the in-can melter and Pt crucible. Pilot scale joule-heated melter in which plate type electrodes were employed to generate heat and whose melting cavity maintained a near constant molten glass level throughout the vitrification process, was designed and constructed. The total amount of molten glass in the melter was about 125 Kg and the average processing rate was 10 ∼ 15 Kg/h. At least 10 hr of retention time was considered for the best quality of the glassy waste form throughout the long-term tests

  17. EPICOR-II: a field leaching test of solidified radioactively loaded ion exchange resin

    International Nuclear Information System (INIS)

    Davis, E.C.; Marshall, D.S.; Todd, R.A.; Craig, P.M.

    1986-08-01

    As part of an ongoing research program investigating the disposal of radioactive solid wastes in the environment' the Oak Ridge National Laboratory (ORNL) is participating with Argonne National Laboratory, the Idaho National Engineering Laboratory, and the Nuclear Regulatory Commission in a study of the leachability of solidified EPICOR-II ion-exchange resin under simulated disposal conditions. To simulate disposal, a group of five 2-m 3 soil lysimeters has been installed in Solid Waste Storage Area Six at ORNL, with each lysimeter containing a small sample of solidified resin at its center. Two solidification techniques are being investigated: a Portland cement and a vinyl ester-styrene treatment. During construction, soil moisture temperature cells were placed in each lysimeter, along with five porous ceramic tubes for sampling water near the waste source. A meteorological station was set up at the study site to monitor climatic conditions (primarily precipitation and air temperature), and a data acquisition system was installed to keep daily records of these meteorological parameters as well as lysimeter soil moisture and temperature conditions. This report documents the first year of the long-term field study and includes discussions of lysimeter installation, calibration of soil moisture probes, installation of the site meteorological station, and the results of the first-quarter sampling for radionuclides in lysimeter leachate. In addition, the data collection and processing system developed for this study is documented, and the results of the first three months of data collection are summarized in Appendix D

  18. Method of processing radioactive solid wastes

    International Nuclear Information System (INIS)

    Ootaka, Hisashi; Aizu, Tadashi.

    1980-01-01

    Purpose: To improve the volume-reducing effect for the radioactive solids wastes by freezing and then pulverizing them. Method: Miscellaneous radioactive solid wastes produced from a nuclear power plant and packed in vinyl resin bags are filled in a drum can and nitrogen gas at low temperature (lower than 0 0 C) from a cylinder previously prepared by filling liquid nitrogen (at 15kg/cm 2 , -196 0 C) to freeze the radioactive solid wastes. Thereafter, a hydraulic press is inserted into the drum can to compress and pulverize the thus freezed miscellaneous radioactive solid wastes into powder. The powder thus formed does not expand even after removing the hydraulic press from the drum can, whereby the volume reduction of the radioactive solid wastes can be carried out effectively. (Horiuchi, T.)

  19. Mobile concrete solidification systems for power reactor waste

    International Nuclear Information System (INIS)

    Tchemitcheff, E.; Bordas, Y.

    1990-01-01

    In late 1988 SGN received an order from Electricite de France (EDF) for the construction of a mobile concrete solidification system to process secondary system resins generated by the P'4 and N4 series PWR power plants in France. This order was placed in view of SGN's experience with low- and medium-level radioactive waste treatment and conditioning over a period of almost 20 years. In addition to the construction of fixed waste processing facilities using more conventional technologies, SGN has been involved in application of the mobile system concept to the bituminization process in the United States, which led to the construction and commissioning of two transportable systems in collaboration with its American licensee US Ecology. It has also conducted large-scale R ampersand D on LLW/MLW concrete solidification, particularly for ion exchange resins. 5 figs

  20. Radioactive waste disposal

    International Nuclear Information System (INIS)

    Cluchet, J.; Roger, B.

    1975-10-01

    After mentioning the importance of the problem of the disposal of wastes produced in the electro-nuclear industry, a short reminder on a few laws of radioactivity (nature and energy of radiations, half-life) and on some basic dosimetry is given. The conditioning and storage procedures are then indicated for solid wastes. The more active fractions of liquid wastes are incorporated into blocks of glass, whereas the less active are first concentrated by chemical treatments or by evaporation. The concentrates are then embedded into concrete, asphalt or resins. Storage is done according to the nature of each type of wastes: on a hard-surfaced area or inside concrete-lined trenches for the lowest radioactivity, in pits for the others. Transuranium elements with very long half-lives are buried in very deep natural cavities which can shelter them for centuries. From the investigations conducted so far and from the experience already gained, it can be concluded that safe solutions are within our reach [fr